doc_fn: draftord/421/g421x-xpart3.html
DocType: Draft
ID: DOE G 421.X-X Part 3
Title: DOE Good Practices Guide
Summary: Draft DOE G 421.X-X, DOE Good Practices Guide, Criticality Safety Good Practices Program Guide for DOE Nonreactor Nuclear Facilities, is intended to present a comprehensive text of good practices for nuclear criticality safety.
Org: EH
Date_Issue: 02/10/1999
Date_Close:
VdkVgwKey: draftord-6
Directive: 421.1X
Text:
APPENDIX A. PERSONNEL SELECTION, QUALIFICATION, TRAINING,
AND
STAFFING PROGRAM
The purpose of the program is to establish (develop and document) the selection, qualification,
training, and staffing requirements for personnel such that persons are qualified to carry out their
assigned responsibilities, that they have a broad understanding and acceptance of the inherent
risks involved with the operations, and that they maintain a job performance proficiency
consistent with effective control of the hazards and risks associated with the operations. Three
broad categories of the program are considered in this appendix. The categories are (1) the
operations and support personnel associated with fissionable material operations outside of
reactors, (2) the installation nuclear criticality safety staff, and (3) visitors and clerical
employees. The personnel selection criteria and depth and breadth of nuclear criticality safety
training are necessarily variable, depending on the work assignments of personnel. The ensuing
discussion in this appendix is intended to provide guidance for organizations establishing new
programs or improving current programs. This guidance is presented in an a posteriori form,
expressly to emphasize that the specificity of structure and nomenclature for personnel selection,
qualification, training, and staffing is illustrative and suggestive rather than recommendatory.
General requirements of the program are provided in the applicable documents listed in
paragraphs 2.1.9 and 2.3.1.12.
A.1 Program for Operations and Support Personnel. The category of operations and support
personnel includes fissionable material handlers and their supervisors, operations support,
design, maintenance, technical support (including the members of the Nuclear Criticality Safety
Organizations) and emergency response personnel, managers and other administrative personnel,
and persons who enter areas where fissionable material is processed, stored, or handled.
Guidance for the selection, qualification, training, and staffing requirements of these persons is
provided in Chapter IV of DOE Order 5480.20 and ANSI/ANS-8.20-1991. As consistent with
job assignments and personnel acknowledgement of job hazards and risks, the following
elements should be considered for inclusion in the training and qualification program.
A.1.1 Continuing proficiency of personnel. Establish the training and qualification program to
provide continuing proficiency of personnel. Tailor the program to job responsibilities, conduct
of the job, recognition of hazards, and acceptance of risk. Establish requirements of refresher
training. Such training shall be provided at least every two years.
A.1.2 Nuclear fission chain reactions and accident consequences. Discuss the concept of a
nuclear fission chain reaction. Make a distinction among families of chain reactions in which
fission rate decreases with time, those that are sustained with a constant fission rate, and those
that have an exponential increase in the fission rate. Describe the time history of super-critical
excursions for both metal (fast neutron) systems and for moderated (slow neutron) systems.
Include information about the kinetic energy release during the fission burst and compare it to
the equivalent energy measured in familiar events; for example, chemical explosions.
Distinguish between the initial, the delayed, and possible radiation doses from criticality
accidents in light of expected doses at various distances from the source of the criticality as
influenced by the rapidity of evacuation. Discuss health effects of criticality accidents.
A.1.3 Neutron behavior in fissioning systems. Describe neutron induced fission, neutron
capture, and neutron scattering and leakage. Discuss the influence of neutron energy on the
fission probability. Explain neutron moderation as the mechanism that reduces the neutron
energy. Identify several good neutron moderators. Discuss the use of neutron absorbers
(poisons) with emphasis on caveats when relying upon soluble neutron poisons.
A.1.4 Criticality accident history. Review and describe selected criticality accidents. Include
a discussion of the causes of the accidents and their terminations.
A.1.5 Response to criticality accident alarm signals. Train personnel in the recognition of,
and the response to, criticality accident alarms and the relationships of distance, time, and
shielding to the reduction in a received radiation dose.
A.1.6 Nuclear criticality safety parameters. Explain and illustrate the influence of various
nuclear criticality safety parameters on process safety. These include mass, geometry,
interaction/separation, moderation, reflection, concentration, volume, density, neutron poisons,
heterogeneity, and enrichment. Illustrate the concept of contingencies (i.e., the loss of a nuclear
criticality safety parameter control) by examples pertinent to facility operations. Review and
discuss facility single parameter limits.
A.1.7 Policy and procedures. Describe the facility management's nuclear criticality safety
policy and include discussions about the use of operational and facility configuration control
procedures for the control of nuclear criticality safety parameters. Inform employees of their
right to question any operations that they believe may not be safe.
A.1.8 Evaluations. Periodically, perform and document evaluations of the training program and
trained personnel. Retain documentation of the evaluations in accordance with DOE O 200.1,
formerly DOE Order 1324.2A and DOE Order 5480.20.
A.2 Installation Nuclear Criticality Safety Staff. This category includes the manager and
members of the installation Criticality Safety Organization who are responsible for performing
computational or comparative evaluations and safety analyses for fissionable material operations;
for developing procedural, process, and control requirements; and for providing procedural,
process, and equipment/facility reviews and approvals, nuclear criticality safety training program
development, and facility operational reviews, appraisals, audits, and investigations. Broad
personnel selection, qualification, training, and staffing requirements are provided in the
applicable document listed in paragraph 2.1.9. The professional personnel charged with
implementing the programs identified in this Guide are designated nuclear criticality safety
specialists (NCSS).
A.2.1 Qualification. There are currently only general qualification requirements, but ongoing
and future qualification of individuals should consider developing confirmable documentation
that addresses the following:
1. A demonstrated capability to perform installation-specific analyses (and, if
appropriate, facility-specific analyses) of the NCSS job and its tasks for existing
and experienced new-hire NCSS personnel.
2. A qualification checklist, file, card, or other record that identifies each applicable
task and the method(s) by which competence has been demonstrated through
apprenticeship for inexperienced new-hire personnel with performance evaluation
based on actual or representative work products.
3. A baseline education of a baccalaureate degree in engineering or science and a
minimum experience in nuclear criticality safety at the facility of one (1) year to
independently perform NCSS tasks (e.g., be classified as a Specialist), and three
(3) years to provide independent review and quality assurance of NCS tasks (e.g.,
be classified as a Senior Specialist). Equivalencies may be established.
4. Certification of final qualification by line and safety management.
5. Periodic competence confirmations based on practical exercises in one or more of
the four functional specialties (see paragraph A.2.2) consistent with their
responsibility and level of activity in each area.
Qualification documentation should address the following three points:
1. Modes:
a. Formal training (onsite and offsite),
b. Apprenticeship and structured on-the-job training, and
c. Professional development activities.
2. Functional specialization (e.g., code validation, double-contingency analysis,
process support, procedure reviewer, peer reviewer):
a. Analysis,
b. Evaluation,
c. Implementation, and
d. Confirmation.
3. Technical proficiency for each functional specialization (e.g., novice, apprentice,
expert).
For the purposes of this section, nuclear criticality safety specialists (NCSS) are collectively the
professional staff with primary responsibility for implementing the activities and programs
required to support this Guide. As defined below, the job designation, "Specialist," is that of the
baseline individual capable of performing independently (perhaps in a designated functional
area). The more highly qualified Senior Specialist has additional responsibility for providing
independent review and oversight. The titles are functional and, therefore, independent of site-
specific human resource designations that may include engineer, senior engineer, etc., and may
have intermediate classifications.
This section provides guidelines for judging initial qualification and continuing competence of
nuclear criticality safety specialist personnel. DOE Order 5480.20, "Personnel Selection,
Qualification, Training, and Staffing Requirements at DOE Reactor and Non-Reactor Nuclear
Facilities" (2-20-91), does not address qualification for the NCSS position specifically.
However, this Order does address qualification requirements for positions that have similar and
related responsibilities (as described in paragraph A.2.3.5).
Engineers and other personnel assigned a limited number of tasks related to nuclear criticality
safety, and technicians performing routine tasks under direct guidance from nuclear criticality
safety specialists (e.g., audits using a checklist, computer testing, etc.) are excluded from this
Guide. However, all shall be qualified for the tasks they perform and may participate in
appropriate portions of the program described herein.
The qualification processes described in this section require documentation and management
approval for qualifying NCSS personnel. Qualification ultimately is the responsibility of
management, which should also address personal characteristics of maturity, judgment, decision-
making ability, independence, and teamwork.
This qualification process allows for specialization in one or more of four functional specialties
in the nuclear criticality safety discipline and assumes the ability and availability for an
anticipated mode-of-progression in job responsibilities from Entry-Level to Senior Specialist
followed by on-going demonstration of continuing competence.
A.2.2 Functional specialties. As noted in the applicable documents above, all nuclear
criticality safety specialists are considered to be personnel who are
familiar with the physics of nuclear criticality and with associated safety practices
to furnish technical guidance to management appropriate to the scope of
operations;
skilled in the interpretation of data pertinent to nuclear criticality safety and
familiar with operations to serve as advisors to supervision;
involved in technical support functions of surveillance, of analyzing facility data,
planning modifications, reviewing programs, and resolving technical problems
within the area of nuclear criticality safety; and
having the responsibility and authority to comment on, or review and concur with,
nonreactor nuclear facility fissionable material operating processes and
equipment.
Thus, collectively, each requires competence in physics of nuclear criticality, associated safety
practices, and familiarity with facility operations. However, individual specialists may be
qualified to meet all responsibilities or may have specialization consistent with the collective
competence of the entire organization. This section allows for, but does not require, broad
specialization consistent with traditional nuclear criticality safety practices. The following
generic areas are addressed:
1. (Safety) Analysis -- performing design analysis for processes and equipment and
for integrated safety assessment with attention to nuclear criticality safety.
2. Evaluation -- performing the unique subset of (safety) analysis that deals with
computer and other evaluations for system subcriticality (subsequent use of the
term analysis/evaluation reflects the interrelationship between the two functions).
3. Implementation (Administration) -- providing administrative interface for
integrating analysis/evaluation with facility operations and practices (e.g.,
procedures, specifications, postings, and training) and for meeting regulatory and
other requirements.
4. Confirmation -- performing audits and other assessments of compliance with
analysis/evaluation requirements and conditions, regulatory requirements, and
other requirements.
The four basic functional categories (analysis, evaluation, implementation, confirmation) are
described in more detail in the following sections, but they are quite arbitrary, both in name and
content. They represent overlapping functions requiring each nuclear criticality safety specialist
to have common baseline knowledge of all responsibilities. An analysis of the nuclear criticality
safety specialist responsibilities and tasks will be necessary to define common baseline
requirements, name and define each functional specialty (if used), and establish qualification
requirements including education, experience, classroom, and on-the-job training.
Each of the four generic functional specialties and the continuing competence to perform NCSS
tasks (whether divided among the analysis, evaluation, implementation, and confirmation
functions) is described below.
A.2.2.1 Analysis. Analysis, according to this Guide, is a thorough description -- developed by
non-reactor nuclear facility safety management, engineering design personnel, and facility
operations supervision -- that includes sufficient facility, equipment, and fissionable material
process descriptions and controls to permit the identification of normal and abnormal conditions
and means of attaining those conditions. This will enable performance of a safety analysis
specific to nuclear criticality that identifies contingent conditions (potential criticality scenarios)
and the bases for subcriticality (nuclear criticality safety evaluation) and for nuclear criticality
safety (NCS).
Alternatively, (safety) analysis is the documented process to systematically identify the hazards
inherent in an operation; describe and analyze the adequacy of the measures taken to eliminate,
control, or mitigate identified hazards; and analyze and evaluate potential accidents and their
associated risks. Criticality safety analysis is an important aspect of this in non-reactor nuclear
facilities. The NCSS provides unique attention to this aspect.
This analysis includes
(a) modelling facility response to accident conditions and performing related studies
or calculations,
(b) documenting and reviewing calculations used in safety analyses, and
(c) coordinating specialized assistance from safety specialty analysts; these latter
safety specialty analyses are analytical determinations that include
performance and review of special safety analyses and computer code
development and validation related to these calculations, including risk
assessment, radiation dose consequence analysis, and criticality analysis;
preparation and review of safety analysis report chapters, or portions
thereof, that relate to specialty skills; and
specialty project assignments that require specific skills and/or
qualifications that an individual possesses.
A.2.2.2 Evaluation. Evaluation, according to this Guide, is a documented demonstration of the
technical computational basis or comparative evaluation with experimental data that provides the
subcritical operating values in support of the nuclear criticality safety analysis. An evaluation is
a subset of an analysis. It frequently is treated as a separate function based on precedent and due
to the unique need for, and substantial attention directed toward, demonstrating subcriticality for
normal, upset, and accident situations and configurations identified in safety analyses. Also of
importance is maintaining and validating the computer codes and systems used for this purpose.
Where evaluation may be a dominant facet of nuclear criticality safety analysis, it does not by
itself constitute safety analysis.
A.2.2.3 Implementation. Implementation is the process of bringing into existence engineered
safety features, parameter limits, and other controls for criticality safety. It includes most of the
nuclear criticality safety activities other than those directly involved in performing safety
analysis and evaluation (and the specific confirmation activities described next). An important
ingredient is providing advice to, or otherwise assuring appropriate attention of, management and
supervision on their numerous and diverse responsibilities regarding maintaining engineered
safety features, keeping parameters within limits, and enforcing other controls for criticality
safety. Another ingredient is providing needed administrative interface to the other functions,
such as integrating the analysis/evaluation inputs and outputs with facility operations (e.g., work
practices, procedures, specifications, postings, and training) and identifying audit needs and
responding to identified deficiencies. A major responsibility is ensuring that regulatory and
other requirements (e.g., demonstration and documentation of double-contingency) are met.
A.2.2.4 Confirmation. Confirmation includes performing audits and other assessments of
compliance with conditions used in, and specified as a result of, the analysis/evaluation
processes, with regulatory requirements and with other requirements. As deviations and
associated trends are identified, corrective actions may be developed jointly with facility
management as part of the implementation function.
A.2.2.5 Continuing competence. Continuing competence consists of maintaining the ability to
perform the NCSS tasks whether they are divided functionally among analysis, evaluation,
implementation, and confirmation.
A.2.3 Qualification process. This section provides guidelines for judging competency and
determining qualification of nuclear criticality safety specialist personnel. It describes the
overall approach, mode-of-progression, functional breakdown, and requirements.
The qualification process described herein requires documentation and management approval.
Thus, it is a form of certification. However, the term is avoided here because DOE Order
5480.18 defines certification as being the product of an accredited, performance-based training
program. Although not necessarily accredited, the underlying principles of performance-based
training do apply to this training program.
The requirements of this Guide also apply to temporary or consultant personnel who serve as
NCSS or perform major NCSS tasks (e.g., evaluations of subcriticality). They should meet the
qualification requirements consistent with the tasks assigned (i.e., Apprentice if performing
under supervision, Specialist if working independently, and Senior Specialist if providing review
and oversight).
A.2.3.1 Principles. The qualification process applies to both new personnel and those currently
serving as NCSS. In the latter case, there is no "grandfathering" per se, but rather a qualification-
by-documentation process. At the time the Guide takes effect at a facility, incumbents may be
assumed to be qualified to fulfill current job responsibilities. They verify the qualification by
documenting how each major set of task items was met, addressing any deficient areas with
training or practical exercises, and meeting the continuing competence requirements (as outlined
in paragraph A.2.6) on an established schedule. The qualification process does not apply to
consultants and temporary personnel under direct guidance, surveillance, and performance or
task acceptance review from qualified nuclear criticality safety specialists for specifically
directed tasks (e.g., audits, computer testing, etc.).
Qualification is judged based on a combination of education, experience, training, and
professional development. The training consists of formal classroom activities and structured
on-the-job training (OJT) that stresses actual work performance under close guidance from a
Senior Specialist and detailed evaluation of, and feedback on, work products.
This section describes processes for qualification of NCSS personnel that are necessary, but not,
by themselves, sufficient. Qualification is the ultimate responsibility of management who should
address other factors such as maturity, judgment, decision-making, independence, and teamwork.
An anticipated mode-of-progression has all incoming personnel progressing from Entry-Level to
Apprentice, to Specialist, and finally to Senior Specialist. This approach enhances overall
organization capability. From a human resource perspective additional levels may be used to
reflect factors such as experience, maturity, and overall unique value to the organization.
An additional level, Lead Specialist, is supervisory in nature. Because assignment is based on a
variety of factors including availability, it is not part of the mode-of-progression.
The qualification program sets standards and minimum times for promotion to higher levels.
Management also may establish maximum times consistent with reasonable progress in meeting
applicable requirements. Failure to qualify within allotted times may result in reassignment or
other action.
Each facility should perform an analysis of the NCSS job requirements and identify associated
tasks. This is similar to, but less detailed than, the job/task analysis (JTA) used in development
of performance-based training programs (see paragraph 2.1.7). The facility-specific task list is
generated based on job descriptions, tasks identified in this Guide, and other resources. The task
list may be used to define either a single NCSS position or specialty functions (e.g., analysis,
evaluation, implementation, and confirmation). The list is also the basis for developing the
behavioral objectives that define and measure performance for the qualification program. With
functional specialization, minimum qualification levels are established in the specialty, related,
and interface areas.
Qualification requirements are a combination of education, experience, formal training,
structured on-the-job training, professional development, and personal factors or characteristics.
Minimum standards are established in each area. However, equivalencies are appropriate and
should be predetermined and specified in general terms in the qualification procedure. One
example is the type and extent of experience that would be considered in lieu of a technical
degree. Another example is to consider certain advanced degrees or specialized research projects
to be equivalent to an amount of criticality safety experience.
A.2.3.2 Classification levels. The anticipated mode-of-progression begins at the Entry-Level for
personnel newly hired to become nuclear criticality safety specialists. They are designated as
Apprentices after having completed specified requirements and having been judged ready to
perform work under supervision of a Senior Specialist. The designation "Specialist" applies
when the individual is deemed capable of performing independently (perhaps in a designated
functional specialty area). A Senior Specialist is more highly qualified and judged to be
prepared for the additional responsibility of providing independent review and oversight and to
be a Senior Specialist for apprentice-level personnel. A Lead Specialist designation applies to
supervisory/management functions that are outside of the mode-of-progression, with availability
depending on staffing levels, organization structure, vacancies, etc.
The titles are generic. Equivalent designations may be used at a given facility. The titles also
may be separate or independent of site-specific human resources designations that may include
engineer, senior engineer, etc. and may have intermediate (e.g., Specialist I, Specialist II, and
Specialist III) or more advanced classifications (e.g., Principal Senior Specialist).
These classifications are intended to apply to NCSS in general and to functional specialties (e.g.,
analysis), if used. Proposed classification-specific requirements are noted in general in a later
part of this section and in more detail in paragraph A.2.4.
A.2.3.2.1 Entry. The Entry-Level classification is for personnel who meet selection criteria with
a combination of education, experience, and training deemed sufficient to begin in the NCSS
anticipated mode-of-progression. Individuals requiring remedial work may stay in the
classification for an extended period of time. Personnel with experience elsewhere in the same
facility or at another facility may be classified Entry-Level while verifying previous completion
of applicable requirements and satisfying others.
It is most likely that any formal classroom training received by NCSS (including onsite
classroom training, if warranted by candidate numbers and staff size) will be while in the Entry-
Level classification. In addition, facility familiarization (e.g., in-facility assignments) and
introductory projects such as for analysis/evaluation (e.g., a standard problem with a single
computer code) and audit observation/participation are included.
Depending on the results of the analysis of jobs and tasks and on other factors including limited
facility access for security clearance reasons, Entry-Level qualification may be divided along
functional or other lines. In such circumstances, advancement to Apprentice status may be
allowed, contingent on completion of deferred requirements prior to qualification as Specialist.
As the Entry-Level requirements are intended to provide qualification for beginning the
Apprentice phase, this classification provides the basics of criticality safety, and of each of the
functional specialties if such division is made.
A.2.3.2.2 Apprentice. The Apprentice classification is for personnel who have a combination of
education, experience, training, and personal characteristics deemed sufficient to perform NCSS
tasks under close guidance and supervision from, and with additional formal review (i.e., over
and above that required by standard practices) by, a Senior Specialist, and who are able to
participate productively in work activities in the specialty and related areas (defined in paragraph
A.2.3.3).
The classification may be multiple level if Entry-Level requirements have been deferred, e.g.,
due to lack of facility access for security clearance reasons.
New hires may meet experience requirements for Specialist while in this category; experienced
new-hire personnel spend only as much time as necessary to complete specific requirements.
An Apprentice may perform tasks in all or designated functional specialty area(s) under the
guidance of a Senior Specialist. Such tasks include performing part of an analysis, running a
computer code to assess subcriticality, participating in an audit, or preparing a draft regulatory or
administrative document. Completion of Apprenticeship in the specialty and related areas,
Entry-Level requirements in the remaining areas, and approval by management lead to Specialist
status.
A.2.3.2.3 Specialist. The Specialist classification is for personnel with the education,
experience, training, and personal characteristics deemed sufficient to perform NCSS tasks in all
or specialty area(s) independently, subject to normal technical and management review. Time-
in-grade, acceptable work products, specialized training, leadership roles, and management
approval lead to Senior Specialist status.
Personnel may meet experience requirements for Senior Specialist while in this category.
Experienced personnel may spend a lesser amount of time consistent with meeting all other
applicable requirements.
This is the minimum level of qualification allowing independent work. Depending on the size
and extent of the facility, it may apply to specific functional specialties, particular subject area(s)
within the specialty, or specific physical portion(s) or area(s) of the facility. The Specialist also
may specialize further (e.g., risk analysis, code validation, human factors, or auditing) as
appropriate to the collective competence of the organization.
The Specialist may act as a subject matter expert in areas of special competence for qualification
of others (as directed by, and under the cognizance of, the designated Senior Specialist mentor).
The work of the Specialist is subject to the usual reviews and quality assurance practices that are
consistent with local procedures.
The Specialist will continue to have a Senior Specialist mentor. The quality of work products,
specialized training, demonstration of sound judgment, initiative, leadership, and time-in-grade
are measures of suitability for qualification as Senior Specialist. Although the subject areas are
basically the same as those addressed as an Apprentice, attention shifts to increasingly
independent action and to leadership and review related to the work of others.
A Specialist performs tasks in all or designated functional specialty area(s) and learns under the
guidance of a Senior Specialist to provide oversight and quality assurance review of the work of
others. Tasks subject to evaluation include performing and reviewing analyses, calculating and
quality assuring computer calculations that assess subcriticality, leading an audit, or preparing
regulatory and other administrative documents. Completion of Specialist qualification in all
areas, or in a functional specialty, leads to Senior Specialist status.
A.2.3.2.4 Senior. The Senior Specialist classification is for personnel with education,
experience, training, and personal traits deemed sufficient to perform NCSS tasks independently
or in a leadership and oversight role. Personal traits (e.g., initiative, organization skills,
integration ability, and maturity) are especially important to the specialist expected to work
independently, train other specialists, provide review and approval of work products and
documents, and, potentially, make "stop work" decisions (see paragraph 5.1.1.7).
Senior Specialists perform routine final reviews and quality assurance of work originated by
Apprentice, Specialist, and Senior Specialist personnel consistent with local procedures. They
also act in a leadership capacity and, thus, should be experienced enough to teach others how to
do the job and take responsibility for the resulting work products.
The major qualification activities for Senior Specialists relate to demonstrating continuing
competence. High-level technical training develops in-house expertise in specific subject areas.
Management training supports increased leadership and eventual assignment as Lead Specialist,
i.e., supervisor or manager.
A.2.3.2.5 Lead. The Lead Specialist classification is for personnel with education, experience,
training, and personal traits deemed sufficient to supervise or manage the nuclear criticality
safety function in general and NCSS tasks specifically. Significant personal traits include those
desirable for Senior Specialists plus attention to such issues as demonstrated desire and ability
prior to assignment and the ability and willingness to make decisions and be accountable for their
results. It should be recognized that even outstanding technical specialists may not make good
lead specialists and/or supervisors or managers.
Lead Specialist is a classification outside of the mode-of-progression. Its availability depends on
organization staffing and structure and on the availability of organizational positions at given
times.
If the supervisor or manager of a multi-disciplinary safety organization is not highly qualified in
nuclear criticality safety, a subordinate Senior Specialist should be designated as Lead Specialist.
The supervisor or manager should be qualified at least to the Apprentice level so as to be able to
perform all basic NCSS tasks, albeit subject to a Senior Specialist mentor's guidance.
A.2.3.3 Functional specialization. If functional specialization is formalized, it is necessary that
each NCSS qualify to a baseline level in all four functional areas in recognition of the interfaces
described previously (e.g., the interactions among analysis, evaluation, implementation, and
confirmation). Preferably this occurs during the Entry-Level classification, or if necessary,
during Apprenticeship. The Apprentice is intended to be qualified to begin on a function-
specific path in any of the four areas (under direct guidance and supervision of a Senior
Specialist).
The qualification level for personnel specializing in each of the four generic functional areas is
shown in Table A.2.3.3-1. Consistent with the anticipated mode-of-progression, Senior
Specialists qualify in their primary functional area, at least as Specialists in the designated related
area, and at least as Apprentices in the two remaining interface areas. This approach recognizes
the strong need for integration of the substance of the functional areas.
Table A.2.3.3-1. Matrix of Qualification Levels for Each of Four Primary Functional
Specialties.
Primary
Functional
Specialty
Qualification Level in Functional Area:
Analysis
Evaluation
Implementation
Confirmation
Analysis
SENIORSPECIALIST
SPECIALIST
Apprentice
Apprentice
Evaluation
SPECIALIST
SENIORSPECIALIST
Apprentice
Apprentice
Implementation
Apprentice
Apprentice
SENIORSPECIALIST
SPECIALIST
Confirmation
Apprentice
Apprentice
SPECIALIST
SENIORSPECIALIST
The analysis and evaluation functions are closely allied through the unique role of performing a
criticality safety evaluation. Thus, practitioners of each need to be qualified at least as Specialist
(i.e., capable of performing, though not necessarily overseeing, the work) in the related field.
Being qualified as Apprentice in the implementation and confirmation functions (i.e., capable of
performing the tasks under close supervision of a Senior Specialist) recognizes the need to
maintain close contact with the facility and to understand how analysis and evaluation results are
implemented and how it will be verified that resulting requirements are met. The
implementation and confirmation functions also must be called upon in support of developing
analysis assumptions up front and ensuring that limitations identified by an analysis are
implemented in the working environment. Double-contingency analyses, for example, and their
documentation have both technical and administrative aspects.
Similarly, the areas of implementation and confirmation functions are closely allied and, thus, are
related to each other. Implementation is a primary subject of the confirmation activities, with
deviations and other identified weaknesses fed back for corrective action. Both require
Apprentice-level familiarity with analysis and evaluation to assist effectively in specification of
input assumptions; translate output conditions to practical in-facility methods (e.g., procedures,
postings, etc.); and verify compliance/consistency between the evaluation (input assumptions and
results), materials and equipment, and practices.
Appendix B presents further discussion in terms of a graded approach.
A.2.3.4 Continuing competence. Continuing competence of the Senior Specialist is maintained
by performing routine tasks in the functional specialty and ensuring periodic performance of
important tasks in the related and interface areas. The resulting work products are subject to
routine peer evaluation and supervisory/management oversight or, if appropriate, to special
evaluation.
Continuing competence is the long-term qualification program for Senior and Lead Specialists.
Entry-Level, Apprentice, and Specialist personnel maintain qualification through satisfactory
progress in the anticipated mode-of-progression.
A.2.3.5 Requirements. Requirements for NCSS are a combination of education, experience,
formal training, structured on-the-job training, professional development, and personal
characteristics. As with any professional position, each NCSS may be expected to achieve
qualification through a personalized program that addresses specific strengths and weaknesses.
General trade-off equivalencies among the qualification factors identified in this section should
be specified in the qualification procedure. Management should document the bases for each
specific application.
Minimum entry-level standards or selection criteria are identified for new hires. A technical
background is required, consistent with the nature of the NCSS tasks.
Exceptions to requirements (e.g., experience in lieu of degree, credit for advanced degree, or
degree-related criticality safety experience) should be documented.
DOE Order 5480.20, "Personnel Selection, Qualification, Training, and Staffing Requirements at
DOE Reactor and Non-Reactor Nuclear Facilities" (2-20-91) does not address nuclear criticality
safety specialist personnel specifically. However, the NCSS are considered non-reactor nuclear
facility technical support personnel who, according to the Order, have duties that include
involvement in surveillance, analyzing facility data, planning modifications, program review, and
technical problem resolution in their area of expertise (e.g., nuclear criticality safety). The Order
establishes baseline education and experience requirements for general technical support
personnel. It also dictates that when a specific position is equivalent to one defined for category-
A reactor personnel, the requirements for the latter apply. Although the NCSS position is not
directly equivalent, it has similarities to that of the category-A reactor technical support
personnel and, to a lesser extent, to that of the reactor engineer.
Table A.2.3.5-1 compares the relevant education and experience requirements for class-A reactor
and non-reactor nuclear facility personnel with those established for NCSS in this Guide. Each
entry applies to what the Order refers to as "positions with authority to review and concur, and
not to entry-level positions." Thus, the proper NCSS comparison is to the Senior Specialist. All
positions have the same education requirement, while according to DOE Order 5480.20 the
reactor and non-reactor nuclear facility technical support positions have the same experience
requirement, and that for the reactor engineer is greater. The Senior Specialist experience
requirement is greater than that for general technical support personnel both in being three years
and facility-specific (or equivalent as developed in paragraph A.2.3.5.2). Thus, minimum
nuclear and onsite experience requirements, which also may depend on functional specialization
and other qualification factors, are not specified separately.
Table A.2.3.5-1. Education and Experience Requirements from DOE Order 5480.20 for
Reactor and non-Reactor Facility Personnel Compared to Those for NCSS Personnel.
Position
Requirement
Category A Reactor
Technical Support*
Non-Reactor
Nuclear
Facility
Technical
Support*
Nuclear
Criticality
Safety (Senior)
Specialist
General
Reactor
Engineer
Education
Baccalaureate
degree in
engineering or
related science
Baccalaureate
degree in
engineering or
related science
Baccalaureate
degree in
engineering or
related science
Baccalaureate
degree in
engineering or
related science
Experience
Job-Related
Nuclear
Onsite
2 years
1 year
4 years
2 years
6 months
2 years
1 year
3 years**
6 months***
6 months***
* SOURCE: DOE Order 5480.20, "Personnel Selection, Qualification, Training, and
Staffing Requirements at DOE Reactor and non-Reactor Nuclear
Facilities" (2-20-91)
** Equivalent site-specific nuclear criticality safety experience
*** Minimum nuclear and onsite experience not specified, as explained in the text
NOTE: General trade-offs between education and experience are allowed by DOE
Order 5480.20 and as delineated in the body of this Guide.
A.2.3.5.1 Education. The minimum qualification or selection criterion for education is the
baccalaureate degree in an appropriate technical field (e.g., engineering, physical science, human
factors, etc.).
For each functional specialty, some degree disciplines may be more applicable than others, e.g.,
chemical engineering for analysis in a "wet chemistry" facility, nuclear engineering or physics
for evaluation, and human factors engineering for implementation and confirmation. Other
curricula should be evaluated case-by-case, with experience or training requirements increased if
appropriate.
For those lacking a degree in an appropriate technical field, holding an associate (2-year) degree,
or holding a baccalaureate degree in a non-technical field, equivalence may be established on-
the-job through apprenticeship with a Senior Specialist as mentor. Experience and training
requirements should be increased accordingly. Applicable experience should be accumulated at
the level of tasks performed by degreed NCSS personnel and comparable to job requirements in
designated functional specialty and physical areas. Equivalence to the technical baccalaureate
degree may be judged in terms of experience at the Apprentice level and demonstrated ability to
perform tasks required of the NCSS.
An advanced degree or applicable graduate work as an indication of additional analytic ability or
maturity may be judged to reduce the experience requirement. Specific studies or research
related to criticality safety also may be applied to reduce training or professional development
requirements as appropriate.
Specific educational background appropriate to specialization -- e.g., probabilistic risk
assessment for analysis, reactor physics for evaluation, human factors for implementation and
confirmation -- may also be acquired through training and professional development. Overall,
factors such as area of study, level, and other considerations may be applied to adjust experience,
training, and apprenticeship requirements.
A.2.3.5.2 Experience. Direct nuclear criticality safety experience at the given facility sets the
baseline standards -- one year for Specialist and two additional years (total of three years) for
Senior Specialist. The Lead Specialist classification generally requires experience beyond that of
the Senior Specialist, although as described above it is not part of the mode-of-progression. The
minimum experience requirements may be adjusted according to educational background factors.
New-hire personnel accumulate experience while participating in the anticipated mode-of-
progression. Previous experience may reduce the requirements. Facility personnel who have
performed jobs most directly associated with nuclear criticality safety receive the greatest credit.
For those who have worked at one or more other facilities, direct nuclear criticality safety
experience is most directly applicable. Experience that is nuclear related (e.g., reactor fuels,
reactors, safety analysis, health physics, industrial safety, or similar disciplines at the same or
other similar facilities) shall be evaluated with credit given in relationship to applicability to
general and specific NCSS tasks. However, even if all experience requirements are judged to be
met, the individual should still complete all facility familiarization requirements, demonstrate
equivalence to specific training requirements, and complete and have evaluated a designated
number of "projects" (e.g., work products of the type included in the apprenticeship program and
used as the basis for judging continuing qualification).
A.2.3.5.3 Training. Formal training courses may be developed for Entry-Level qualification and
for later activities as appropriate to the size of the facility organization. Such courses should be
performance-based, consistent with the guidance of ANSI/ANS-8.20-1991 (even though training
of NCSS personnel is not addressed explicitly). Subject matter recommended by ANSI/ANS-
8.20-1991 will be addressed, though at greater depth consistent with the needs of the NCSS
audience. Course formats other than lecture, e.g., seminars, workshops, etc., are preferred.
Consistent with ANSI/ANS-8.20-1991 and performance-based training, evaluations of candidate
performance should be conducted. A comprehensive written examination is one alternative,
although realistic problem solving activities that demonstrate both knowledge and ability to
apply it appropriately may be the better choice. Open-book exercises such as applying
ANSI/ANS standards, guides, and other reference materials are appropriate.
Many offsite courses are appropriate for Entry-Level and more advanced qualification. Such
courses should be evaluated for applicability based on characteristics including subject matter,
faculty breadth and expertise, audience makeup (e.g., peers and other contacts at similar
facilities), and instructional approach (lectures, workshop sessions, and practical exercises).
Whether a formal evaluation of participants is provided, it should be verified independently (e.g.,
through the Senior Specialist) that the desired learning has taken place. Value beyond subject
matter is recognized due to interactive activities with faculty and peers from other facilities.
Applicable offsite courses include the general short courses offered by the University of New
Mexico and Los Alamos National Laboratory's critical facilities. Broad or specialized courses on
safety analysis, computer and other computational methods, audits and inspections, human
factors, and other related subjects also deserve consideration.
An appropriate mix of onsite and offsite training courses, refresher seminars, and workshops can
provide the NCSS with knowledge that will support development and maintenance of requisite
facility-specific skills.
A.2.3.5.4 On-the-job training. As with other professionals, the NCSS performs basic recurring
tasks that are similar, but not repetitive in the sense of those performed by production-oriented
operators and technicians (e.g., analyses using the same methods, but each time for a different
situation). Likewise, individual NCSS, even new hires, have differing needs for qualification.
Thus, formal training courses generally are less appropriate than learning-by-doing. A structured
on-the-job training (OJT) approach is indicated for this purpose. The OJT mode of qualification
can be applied from entry level through continuing qualification, with most direct use during the
Apprentice and Specialist classifications. In all cases the training proceeds under the close
supervision and guidance of a Senior Specialist.
On-the-job training, whether standardized or individually orchestrated, is primarily one-to-one
(or one-to-a-few) between the candidate(s) and a Senior Specialist. Work performed by the
candidate (prior to qualification as Specialist) is subject to careful supervision by the Senior
Specialist and to routine peer review, as applicable. Senior Specialists serve as mentors. Subject
matter experts (SME) qualified and experienced in performing a particular task may, on a case-
by-case basis, be assigned by the Senior Specialist to direct, observe, or evaluate performance of
activities. Periodic evaluation of candidate performance is required.
On-the-job training depends heavily on individual initiative of the NCSS candidate and uses
directed self-study -- a training setting without a full-time instructor in which objectives and
conditions are provided by the Senior Specialist, using training materials, or in-facility reviews
and instruction. A qualification checklist, "card" file, or other means (for simplicity, hereafter
referred to as the checklist) may serve as the basis for directing and documenting progress and
completion of designated milestones. Activities that are the means for judging completion of
specific tasks include
Review -- deliberate critical examination of references and training materials,
Observe -- directed careful analytic attention to the performance of another,
Perform -- performance of actual or equivalent tasks using necessary references,
materials, and tools in the normal job environment, and
Simulate -- mimicking task performance at the job site or through task
performance on a mock-up device similar to the actual equipment and work
environment.
For activities such as review of documents and observation of facility evaluations, which do not
automatically generate a work product that is subject to review by a Senior Specialist or subject
matter expert, an appropriate performance evaluation technique is required. This may take the
form of a documented discussion -- explanation or other techniques of evaluation that indicate
proficiency -- with the Senior Specialist, a more formal evaluation, or a written examination.
Workbooks or notebooks that can be reviewed by the Senior Specialist or others also are
appropriate. In all cases it is necessary to document qualification details using the checklist.
The on-the-job training requires performance-based development, i.e., systematic determination
of specific tasks, task analysis for skills and knowledge, and learning objectives that define the
expected content and level of performance.
Exemption from, or reduction in, requirements (e.g., fewer analyses, evaluations, or audits) may
be based on previous experience, but preferably on proficiency testing or work-product review.
Evaluation modes may include
Board evaluations based on oral, walk-around, notebook review, or other
demonstration; these may in turn be divided into
- Mini-boards with the Senior Specialist and a supervisor or subject
matter expert, as appropriate, to judge intermediate milestones, and
- Final board with, at a minimum, the Senior Specialist,
supervisor/manager (chair), a designated SME, and a "facility"
representative; other senior specialists also may be included;
Projects that test technical ability, judgment, etc. reviewed by teams composed
similar to the boards; the process may be accompanied by a final oral "defense"
(or board) evaluation.
Board members should receive training on conduct and participation in the process. The chair
and/or Senior Specialist should receive more detailed training on board setup and conduct.
Where seminars, workshops, and offsite courses (see also paragraph A.2.3.5.3 on Training) are
used as a basis for meeting what is otherwise an OJT task, the content (i.e., the learning by the
candidate) should be evaluated for applicability using the standard OJT processes.
A.2.3.5.5 Professional development. Professional development activities apply to all
classification levels and are a major element in Senior Specialist initial qualification and
continuing competence. They are subject to review with the Senior Specialist, supervision, or
others. Presentation of a seminar may be an appropriate way both to verify the extent of learning
and to share the experience with peers.
Professional development activities include, but are not limited to,
educational activities and technical meetings such as conferences, seminars,
clinics, workshops, tours, forums, or symposia,
college courses (including home study),
professional development courses,
onsite workshops or seminars,
publication of papers, reports, or other peer-reviewed documents,
special onsite and offsite work assignments (e.g., task force membership or a
temporary in-facility assignment),
preparation of a position paper for critical review on a contentious issue of nuclear
criticality safety,
intra-site committees (e.g., safety overview),
inter-site committees (e.g, multi-site corporate, DOE-regional, or DOE-wide), and
regional or national committees (e.g., American Nuclear Society Nuclear
Criticality Safety Division, Institute of Nuclear Materials Management,
ANSI/ANS-8 Standards).
In each case, active participation (e.g., as instructor, chair, or officer) carries more credit than
mere attendance.
A.2.3.5.6 Personal characteristics. The qualification process should include on-going evaluation
and judgment of the readiness of the NCSS candidate to do the whole job and do it
independently. The Senior Specialist should address such issues during the course of the
process. Management has the prerogative on the final judgment based on factors that can include
the candidate's judgment, technical ability, and initiative. As noted previously, each Senior
Specialist may have review, approval, and/or "stop work" authority (e.g., in paragraph 5.1.1.7)
and, thus, needs to be judged capable of implementing them.
A.2.3.6 Overall qualification. The qualification process may be coordinated through the use of a
qualification checklist. This may be a generic form that is readily customized to needs of
individual NCSS candidates. Where functional or other specialization is employed, the
checklists may be tailored appropriately. The checklist should provide guidance on the tasks to
be performed and the means by which the NCSS will be evaluated (i.e., objectives).
By implementing a formal qualification program for the first time, existing personnel use a
qualification-by-documentation approach with the same checklist. They
indicate education, experience, etc.,
indicate how training/task requirements were met and equivalence to Entry-Level
and Specialist programs,
establish a schedule for meeting any serious deficiencies, and
focus primarily on the continuing competence requirements to verify their ability
to perform major work-product tasks.
For new or existing personnel, exceptions to formal qualification requirements may be made
based on judgment and documented alternatives (e.g., experience in lieu of a degree or credit for
an advanced degree, specialized course work, or relevant research activity).
The process should have provisions for final confirmation of competence made by safety
supervision or management (and, if applicable to a specific facility, by cognizant line
management). A comprehensive written examination, oral or facility walk-around examination,
or a combination thereof, should be used to address all major NCSS tasks. Evaluation of realistic
and representative work products should be an important part of the process. Remedial actions
for failures need to be specified.
A.2.4 Classification-specific qualification programs. Subject-matter content for the NCSS
qualification program should be developed from an installation-specific analysis of the job and
its tasks. The analysis may be used to designate functional specialties. Each facility has the
option to use the generic classification levels proposed in this document or use site-specific
classifications consistent with its human resources system.
A.2.5 Functional-specific qualification programs. If functional specialization is employed,
the analysis of the job and its tasks should be used to make an installation-specific check list. As
described in the previous section, these may apply to all of the Entry-Level, Apprentice,
Specialist, and Senior Specialist classifications at progressively more detailed levels. Additional
specialization may be employed with respect to the collective capability of the organization.
A.2.6 Continuing competence. Continuing competence demonstration is required of each
Senior Specialist. It is addressed here rather than with the classifications due to the tie-in to the
four functional specialty areas.
Existing personnel or highly experienced new-hires who have been performing at the Senior
Specialist level employ a qualification-by-documentation approach as described previously
(paragraph A.2.3.6). A Senior Specialist who is a peer should be assigned to validate or verify
basic and continuing competence requirements.
Those designated as Apprentice or Specialist are not specifically subject to these requirements
with normal progress in the qualification mode-of-progression. However, comparable activities
and consistent frequencies should be included on the qualification checklists.
The actual content of the program to ensure continuing competence should be derived from the
analysis of the job and its tasks. General areas and issues are addressed below.
Periodic training in technical and administrative subjects assists in maintaining and improving
job performance and developing broader scope and depth in specific knowledge and skills.
Retraining on subjects included in the Entry-Level, Apprentice, and Specialist portions of the
qualification program is generally not necessary. However, if individual or group performance
problems are identified, they should be addressed. Specific subjects that are appropriate for
continuing training include, but are not limited to,
facility and industry operating experience, audit findings, and deficiency trends,
changes to systems, components, and applicable procedures,
changes to DOE Orders, National Standards, and other guidance, and
major changes to NCSS tasks.
Classroom training, seminars, or "required reading" may be appropriate methods. Examinations,
discussions, or evaluation of work products should be used for confirmation consistent with the
approaches applied in mode-of-progression qualification.
Continuing competence requirements are based on performing activities in each of the four
functional areas (including all required reviews and approvals for the designated level, with
additional review if appropriate). The requirement may be annual or graded based on functional
specialization (e.g., annual for the specialty area, biennial for the related area, and triennial for
general or interface areas). Work products should be provided and evaluated in each of the
following areas:
facility familiarity (e.g., through periodic tours and meeting attendance);
analysis for criticality safety (e.g., double-contingency analyses);
evaluation of subcriticality;
facility audit activities;
implementation activities, e.g.,
- task, job, and procedure audits,
- investigation of criticality safety limit violations, and
- evacuation procedure audits;
onsite professional development activity, e.g.,
- regular task, but elsewhere in the facility or with a different
organization, and
- appropriate activity from the list in paragraph A.2.3.5.5 or
equivalent; and
offsite professional development activity (e.g., see paragraph A.2.3.5.5).
For large facilities where qualification may be based on physical areas or processes, the
requirements apply to each applicable area.
Where functional specialization is used, each work product may be geared according to the
specialization and classification shown in Table A.2.3.3 as follows:
specialty field at the Senior Specialist level -- lead an analysis, evaluation, or audit
effort; review and quality assure analyses, evaluations, or audits; or complete
major administrative responsibilities;
related field at the Specialist level -- perform a new and original analysis,
evaluation, audit, or administrative task; and
general/interface fields at the Apprentice level -- perform (under supervision)
representative analysis/evaluation exercises or participate in audits and
administrative tasks.
A.2.7 Documentation and records. Documentation shall be maintained on the qualification of
each NCSS. The qualification checklist or equivalent documentation may be designed to serve
this purpose. It should show each applicable task, how and when it was accomplished,
examination results if applicable, and who verified completion (of single tasks or groups of
tasks). Final approvals by the Senior Specialist and supervision/management should also appear.
Records shall be developed consistent with facility policies and procedures for training and
qualification. Retention of documentation and records shall be consistent with DOE Orders and
facility procedures.
A.2.8 Evaluation and Documentation. Evaluations of the NCSS qualification program and
personnel should be performed and documented periodically. Documentation of these
evaluations should be retained in accordance with the applicable document listed in paragraph
2.1.2. Additional documentation requirements are provided in the applicable document listed in
paragraph 2.1.9. APPENDIX B. GRADED APPROACH
B.1 Graded Approach to Criticality Safety Analyses and NCSEs. A graded approach to the
performance of criticality safety analyses and the supportive nuclear criticality safety evaluations
(NCSE) should be exercised. A graded approach to the performance of criticality safety analyses
acknowledges that different levels of effort and documentation are appropriate for different
complexities of facility fissionable material operations (i.e., handling, processing, and storing)
and the associated methods and controls applied to maintain subcriticality and safety.
The classification of facility complexity and levels of analyses and evaluations to be performed
should be determined at an organizational level independent of facility operations or production
(e.g., the safety organization reporting to the installation/facility manager). This determination
should be based upon the technical judgment of a nuclear criticality safety specialist.
B.1.1 Levels of analyses and evaluations. Levels of analyses or evaluations range in effort
from simple references -- to common engineering and safety judgment and to national consensus
standard subcritical values (e.g., 450g 239Pu) using a highly reliable control on allowed facility
fissionable material mass -- to a complicated validated computation of neutron interacting arrays
of dissimilar systems involving materials having variable nuclear parameters and numerous
administrative/procedural and physical controls benefitting from probabilistic risk analyses.
Three levels of analysis and evaluation are considered: Levels A, B, and C. Level A analyses
and evaluations may be performed for facilities having fissionable material inventories and
operational conditions that will remain within the envelope of conditions specified for subcritical
values within national consensus standards. Level B analyses and evaluations are performed for
facilities having fissionable material inventories or operational conditions that exceed the
envelope of national consensus standard subcritical values but have fissionable material
inventories and operational conditions that may be analyzed to be safely subcritical by reference
to commonly accepted and used handbook or safety guide values. Where these values are not
based directly on experimental data, such as tables or figures based solely on calculated values,
they should be confirmed from two independent sources. Level C analyses and evaluations are
typically performed for fissionable material inventories and operational conditions that cannot be
addressed with national consensus standards or handbook values. Level C analyses and
evaluations may involve the application of computational techniques requiring computer
program documentation, verification, validation, and user qualification.
In all cases, it shall be shown that all normal and credible abnormal operational conditions and
contingencies remain within the envelope of the specified subcritical process and nuclear
parameters. All physical and administrative controls used for ensuring the subcritical values
shall be clearly identified. The reliabilities of the controls should be described to be acceptable.
This is already covered in paragraph 5.6.1. By order of preference, referable facility historic
data, industrially accepted guidance, and, lastly, experienced engineering judgment about human
and equipment reliability should be used to defend the reliability of nuclear criticality safety
controls.
B.1.1.1 Level A. Level A evaluations are performed by direct reference to national consensus
standard subcritical values. Such references include ANSI/ANS-8.1-1983,R88, Nuclear
Criticality Safety in Operations with Fissionable Materials Outside Reactors, and
ANSI/ANS-8.15-1981,R87, Nuclear Criticality Control of Special Actinide Elements. Though
no additional verification of the subcritical values are required, a clear comparative evaluation of
the operation being evaluated should be given along with the basis of safety.
B.1.1.2 Level B. Level B evaluations are performed with referenced values derived from
published handbooks, safety guide subcritical values, or criticality data. Such well known
references include, but are not limited to, LA-10860-MS, Critical Dimensions of Systems
Containing 235U, 239Pu, and 233U (paragraph 2.3.2.8, 2.3.2.11 of this Guide); NUREG/CR-0095
ORNL/NUREG/CSD-6, Nuclear Safely Guide TID-7016 Revision 2 (paragraph 2.3.2.9 of this
Guide); and ARH-600, Criticality Handbook. The referenced values should be based directly on
experimental data or should be verified to be consistent with independent handbooks or safety
guide subcritical values or validated computational techniques.
Where applicable data are directly available, subcritical values shall be established on bases
derived from experiments, with adequate allowance for uncertainties in the data. In the absence
of directly applicable experimental measurements, the subcritical values may be derived from
calculations made by a method shown to be valid by comparison with experimental data.
Level B analyses should be based on critical data only after appropriate margins of subcriticality
have been applied to the critical values. The use of unpublished experimental logbook data
requires comparison with a Level C evaluation as described in paragraph B.1.1.3. The
identification of and reliability of controls shall be as described in paragraph B.1.1 above.
B.1.1.3 Level C. Level C analyses are performed by the use of a validated computational
technique. Examples include ORNL/NUREG/CSD-2/VI/R2, KENO-Va, An Improved Monte
Carlo Criticality Program with Supergrouping; LA-7396-M, Rev.2, MCNP, A General Monte
Carlo Code for Neutron and Photon Transport; and BNFL SAG/80/P29, Criticality Assessment
Using the Limiting Surface Density (NBn2) Method and Examples of Application. Acceptable
margins of subcriticality and range of applicability for the chosen evaluation technique should
have been determined and documented for use in criticality safety evaluations. No single
computational result should be used for determining the subcriticality and safety of an operation.
Rather, multiple results showing trends and computational reliability will be used. The use of
Level C analyses should be in conformance with ASME NQA-2 requirements. The
identification of and reliability of controls shall be as described in paragraph B.1.1 above.
B.1.2 Complexities of facility fissionable material operations. Complexities of facility
fissionable material operations range from single operations having less than a significant
quantity of fissionable material to multiple operations having large quantities of fissionable
materials processed in multipurpose facilities with many types of interfacing operations and
support activities. Four classes of complexities are defined as follows:
B.1.2.1 Class I. Class I facility operations have less than significant quantities of fissionable
materials presenting no significant risk of criticality within item control areas or material balance
areas. Nuclear criticality safety is applied through facility nuclear material possession and
accountability limits.
B.1.2.2 Class II. Class II facility operations have significant quantities of fissionable materials
and have operations limited to repetitive and routine activities. No significant quantities of
fissionable material wastes are generated in Class II facility operations. Nuclear criticality safety
is applied with physical barriers such as spent or fresh fuel storage racks and single item handling
devices. The fissionable material operations are performed in control areas that effectively
preclude neutron interaction among items. Examples include, but are not limited to, fuel element
examination operations, fissionable material item packaging, and storing.
B.1.2.3 Class III. Class III facility operations have significant quantities of fissionable materials
and perform operations that influence other fissionable material operations within the facility.
Examples include, but are not limited to, analytical laboratories, foundries, machine shops,
dimensional inspection shops, nondestructive testing shops, etc. that exchange materials among
the various operations. Significant quantities of fissionable material wastes in solid and liquid
forms are generated and collected but are not processed to finally recovered forms. The
fissionable material operations are performed in effectively non-neutron interacting item control
areas and material balance areas.
B.1.2.4 Class IV. Class IV facility operations are multipurpose and include all of the
characteristics of a Class III facility but with the addition of complex operations including
solution, waste recovery, waste processing, and decontamination and decommissioning
operations. Additionally, the fissionable material operations may be performed in neutron
interacting item control areas and material balance areas.
B.1.3 Analysis Content. Despite the level of effort and documentation of evaluations and
analyses and the complexity of an operation, the same fundamental elements should be included
and identified in the safety analyses for each discrete operation within the facility. The safety
analyses should be retained in accordance with paragraph 2.1.2. These elements include the
following:
B.1.3.1 Operational description. Using verified as-built sketches, drawings, or flow diagrams of
the equipment, portable containers, and of processes and facilities, the description of the intended
fissionable material operation under analysis should be provided for which the hazard of
criticality exists. Care should be exercised to identify, for additional analysis, ancillary support
equipment or activities that may require independent safety analyses (e.g., vacuum producers,
nonfissionable material feed chemical make-up and supply, compressed gas/air, waste collection,
ventilation, transportation, neutron interaction among other fissionable material systems, etc.)
and that may affect, or be affected by, the operation under consideration. The description should
be of sufficient detail to permit independent evaluations and safety analyses of the operation.
B.1.3.2 Fissionable material forms. Bounding descriptions of the chemical and physical form(s)
of fissionable material in the operation should be provided, including isotopic content, resulting
concentrations, densities, degrees of neutron moderation, degrees of neutron interaction and
reflection considered, and the physicochemical stability of the fissionable material in the
anticipated normal or abnormal operating environment.
B.1.3.3 Credible operating condition changes. This includes the description of the normal and
abnormal credible changes in operating conditions that could alter a nuclear parameter (i.e.,
geometry/volume, spacing/interaction, neutron absorption, concentration/density, mass,
moderation, reflection, and enrichment) beyond intended operating conditions. The description
should include a characterization of any resultant conditions, masses, forms, materials, etc.
adversely affecting subcriticality and safety.
B.1.3.4 Analysis of accident scenarios. This includes the identification of event sequences
leading to credible nuclear criticality accident scenarios (a single scenario probability exceeding
a frequency of 1 x 10-6 per year) and associated consequences to workers, the public, and
facilities. Bases should be specified if no credible accident scenarios can be determined.
B.1.3.5 Need for CAS or CDS. A review for the need and placement of a nuclear criticality
accident alarm or detection system should be provided. Alarm and detector coverage shall be
provided as necessary, or a reference supplied that indicates fulfillment of the alarm or detector
need and placement (see Section 5.4).
B.1.3.6 Safety controls description. The description of the passive and active safety controls
that are part of the operation should be identified and should include the intended
administratively or physically controlled value(s) for each of the nuclear parameters. If a specific
nuclear parameter does not affect the operation, a short justification for excluding the nuclear
parameter from the analysis should be provided. Technical practices and measurement control
programs used for ensuring the reliability of safety controls should be provided.
B.1.3.7 NCSE summary. The summary description of the validated technical nuclear criticality
safety evaluations (computational or comparative) showing the subcriticality of the operation
under normal and abnormal conditions should be provided. The safety evaluation should
identify and consider interactions with any other fissionable material operations within the
facility.
B.1.4 Performance of nuclear criticality safety analyses and NCSEs. As indicated above, a
"Graded Approach" acknowledges that different levels of effort and documentation are
appropriate for different complexities of facility fissionable material operations. The gradation
of levels of effort and of documentation and the complexities of operations may be seen as a two-
dimensional matrix, as shown in Table B.1.4-1, which is used for grading the approach and
resources required for performing the nuclear criticality safety analysis and NCSE. The table
footnotes provide explanations about the resources that are numbered 1 through 4. A peer review
is to be conducted for any NCS analysis and associated evaluation.
B.1.5 Results of the Graded Approach. As indicated by the combination of complexities of
operations with levels of effort required for analyses or evaluations, as described in paragraphs
B.1.1 and B.1.2 and shown in Table B.1.4 above, an NCS analysis or evaluation may result in a
seemingly minor safety document for Class I - Level A type analyses or evaluations, whereas a
Class IV - Level C analysis or evaluation may result in a rather prodigious report. Additionally,
the required resources can be quite variable. In all cases, the safety analysis shall contain all of
the elements described in paragraph B.1.3 that are relevant to the operation, or appropriate NCS
analyses or evaluations that supply these elements may be referenced. More in-depth
descriptions and examples of such analyses and evaluations are provided in Sections 5.7, 5.8, and
5.9.
Table B.1.4-1. Resources required for performance of the NCSE.
Facility Complexity
Analysis Effort
Class I
Class II
Class III
Class IV
Level A
1
2
2
2
Level B
2
2 or 3
3
3
Level C
--
3 or 4
4
4
Legend: Level of Effort and Personnel Qualifications
1. Operations Supervision.
2. Qualified Nuclear Criticality Safety Specialist having experience interpreting safety
guides and critical data references, in conjunction with an experienced
process/operations engineer who is familiar with operational process, equipment,
and facility normal and abnormal conditions.
3. Qualified Nuclear Criticality Safety Specialist having operational and process
knowledge and experience interpreting safety guides and critical data references, in
conjunction with an experienced process/operations engineer who is familiar with
operational process, equipment, and facility normal and abnormal conditions.
4. Qualified Nuclear Criticality Safety Specialist having operational and process
knowledge, experience interpreting safety guides and critical data references, and
computational validation and analysis experience, in conjunction with an
experienced process/operations engineer who is familiar with operational process,
equipment, and facility normal and abnormal conditions.
APPENDIX C. ESTIMATING THE WAITING TIME UNTIL THE
SIMULTANEOUS COLLAPSE OF TWO CONTINGENCIES
(Adapted from Author's Text)
C.1 Introduction. This appendix provides an interface between criticality safety and safety
analysis. Recent emphasis calls for probabilistic safety assessments in addition to the traditional
qualitative and quantitative, but deterministic, assessments. That emphasis supplies the motive for
this appendix, which is narrowly focused on the Double-Contingency Principle (DCP) as applied
in criticality safety practice.
C.1.1 DCP Review. The definition of the DCP is stated as, "Process designs shall, in general,
incorporate sufficient factors of safety to require at least two unlikely, independent, and concurrent
changes in process conditions before a criticality accident is possible." For example, given a
fissile material workstation in a glovebox, the "two unlikely" events are inadvertent double-
batching and inadvertent flooding with water. In this example, work begins at the workstation at
time zero. The purpose of this probabilistic model is to make a probabilistic statement about the
waiting time until the workstation is simultaneously flooded and double-batched.
C.1.2 Markov Model. A Markov model is convenient and tractable. In such a model, (1) the
time span from recovery from a flooded condition to onset of the next flooded condition is an
exponentially distributed random variable, (2) the time span from the onset of a flooded condition
to recovery from that flooded condition is an exponentially distributed random variable, and (3)
those two random variables are independent. A similar set of statements applies to the double-
batching situation.
C.1.3 Probabilistic Description. Given estimates of mean failure and mean recovery times of
the two independent contingencies, the model can be used to generate a probabilistic description of
the waiting time to the first simultaneous collapse; or, if estimates of mean failure and mean
recovery times of the two independent contingencies are unavailable, the model can be used to
construct parameter surveys to bound estimates that could satisfy a criterion for mean time to
simultaneous collapse.
C.2 General Markov Model. The construction of a Markov model for the general situation
follows. For k = 1, 2, let k(t) = 1 if contingency k is in its desired state; let k(t) = 0 if
contingency k is in its undesired state. Suppose that at time zero both contingencies are in the
desired states:
1(0) = 1 and 2(0) = 1. For k = 1, 2, let 1/ k be the mean time between transitions from desirable
to undesirable states for the kth contingency. Similarly, let 1/k be the mean time between
transition from undesirable to desirable states. If the Markov model is invoked, then the sojourns
between transitions are independent, exponentially distributed random variables. The process is
assumed to begin in state (1,1) (i.e., 1(t) = 1 and 2(t) = 1). The waiting time until the first visit
to state (0,0) (i.e., 1(t) = 0 and 2(t) = 0) is to be determined. That waiting time is also a random
variable to be determined as follows.
The (0,0) state is that in which both contingencies are in undesired states, and in practice, is a state
from which exit is possible. However, it is convenient for modeling purposes to make (0,0) an
absorbing state, one from which exit is not possible. If state (0,0) is an absorbing state and T, a
random variable, is the waiting time until the first visit to (0,0), given that the process begins in
state (1,1); then for any t > 0, the events [T t] and [ 1(t) = 0 and 2(t) = 0] are equivalent. That
equivalence simplifies the following mathematical demonstration.
Figure 1 displays a state transition diagram for the two-state Markov process. For i = 0,1 and
j = 0,1; let Pij(t) P[ 1(t) = i and 2(t) = j]. The incantation that corresponds to the right hand side
of the last definition is "probability that 1 at time t equals i and 2 at time t equals j." Then from
the figure, the system of first order differential equations that the Pij satisfy is
(1)
Since the process begins in state (1,1), the initial conditions for system (1) are: P11(0) = 1,
P10(0) = 0, P01(0) = 0, P00(0) = 0.
The technique used for constructing a solution to system (1) is the Laplace transform. Let L
represent the Laplace transform operator, and for i = 0, 1 and j = 0, 1 let fij = L Pij. The application
of L to system (1) yields
(2)
Solving system (2) for f00 yields:
f00(s) = N(s)/D(s) (3)
where
N(s) = (2 1 2)s + ( 1 1 2 + 1 2 2 + 1 2 1 + 1 2 2)(4)
and
D(s) = s[s3 + (2 1 + 2 2 + 1 + 2)s2 + ( 1 1 + 2 2 + 3 1 2 + 1 1 + 1 2
+ 2 1 + 2 2 + 2 2)s + ( 1 1 2 + 1 2 2 + 1 2 1 + 1 2 2)](5)
Equations (4) and (5) are unnecessarily expanded to highlight the symmetry of the subscripts.
The Laplace transform of P[T t] is f00, and P[T t] is the cumulative distribution function (CDF)
that describes T, the waiting time until the simultaneous occurrence of the two contingencies.
Hence, a fundamental property of Laplace transforms and the fact that P[T 0] = 0 imply that s
f00(s) is the Laplace transform of d/dt P[T t]. But d/dt P[T t] is the probability density function
(PDF) that describes T; let fT represent that PDF, and let gT LfT. Then from (3):
(6)
where the last equation in (6) defines D*.
To invert gT requires finding roots of the cubic D*; the coefficients of D* appear in (5). In
application where the i and the i are assigned numerical values, computer-based root finding
routines could be used; and fT could be found by inverting gT.
Although inversion of gT is unproductive in the general case, useful information can be extracted
from gT without inversion. That is;
(7)
where represents expectation. Hence gT is a moment generating function for T. In particular, if
exp(-st) is expanded in a Taylor series about 0, it is found that T = -g T (0) where the prime
represents differentiation with respect to s. Differentiation and algebraic manipulation applied to
(4),(5), and (6) yields:
(8)
Equation (8) is presented in the expanded form to highlight the symmetry of the relationship.
A special case of (8) is enlightening. In practical cases, if application of the double contingency
principle is to yield significant safety advantage, the mean times of transition from desirable to
undesirable states should be much longer than the mean times of transition from undesirable to
desirable states. In the context of the model, this translates into the assertion that for every i = 1, 2
and j = 1, 2, i j. In this special case (8) becomes
(9)
The advantage to be gained by using two contingencies instead of one contingency is
demonstrated in the following quantitative estimate of examining the mean time to the first
simultaneous occurrence of two contingencies. Suppose 1/ 1 = 5 years, 1/ 2 = 10 years, 1/1 = 5
days, and 1/2 = 2 days. Then equation (9) applies, and T 2600 years; the advantage is
substantial in this case.
Equation 8 is provided for the general case and equation 9 is provided for the special (and usually
applicable) case.
C.3 Symmetric Case. The "symmetric case" is for circumstances in which both contingencies
are described by identical probabilistic models, i.e., 1 = 2 and 1 = 2 . The symmetric
case can be treated as above by starting with a state transition diagram and writing down the
corresponding first-order linear system of differential equations. The system is 3 x 3 matrix
instead of 4 x 4 matrix because the states (0,1) and (1,0) are indistinguishable.
The symmetric case is logically equivalent to what reliability theorists call the two-unit-active-
redundant case, and it has been completely solved, and is provided as follows.
As before, let T be the waiting time until the first visit to state (0,0). Then for time t 0,
(10)
where
(11)
and
.
(12)
Equation (10) is a complete probabilistic description of T. To obtain a corresponding result for the
asymmetric case requires finding the roots of the cubic D* defined in equation (6).
Although there is no simple equivalent of (10) for the asymmetric case, equation (10) may be
conservatively used in the asymmetric case by setting = max( 1, 2) and = min(1, 2). Such
an application may be useful to gain quick insight and might even suffice without further analysis
if the result satisfies the preset criterion.
REFERENCES
APPENDIX D. EXAMPLES OF DESIGN OF NUCLEAR
CRITICALITY SAFETY CONTROLS
D.1 Double-contingency analyses. The purpose of this section is to provide an example of a
double-contingency analysis of a potential criticality scenario to evaluate compliance with the
Double-Contingency Principle (paragraph 5.7.7). The main points of the illustration are
(a) identifying the potential criticality scenario (paragraph 5.7.5),
(b) evaluating the scenario for compliance with the Double-Contingency Principle
(paragraph 5.7.7), and
(c) identifying the associated means of control (paragraph 5.7.4).
Example # 1 below provides an involved scenario analysis with control reliability/failure
evaluations for acceptability.
D.1.1 Example # 1. Assume that the quantity of fissile nuclide required for a particular operation
is 2 kg of 239Pu in oxide form that is greater than the minimum critical mass. On this basis,
criticality protection is solely dependent upon excluding moderation from the area since
geometry/volume is not controlled. Since nuclear criticality safety depends on the control of a
single nuclear parameter, moderation, two separate and independent barriers need to be provided
to prevent loss of moderation control. Thus, as shown in Figure D.1.1 (upper left-hand corner),
nuclear criticality safety considerations require that moderating liquids be excluded from the dry
processing location containing fissile material. Reviews of the design identified two credible
sources of liquid to the dry location under operating conditions: (1) liquid backflow from an
associated off-gas scrubber system, and (2) the unauthorized manual addition of liquids by
operating personnel. Before proceeding, a brief description of the scrubber system is given below.
The off-gas scrubber system is provided to cool and scrub the off-gas coming from the dry
location that contains fissile material in powder form (upper left-hand corner of Figure D.1.1). A
vacuum is pulled on the system using a vacuum air jet located above the separator tank that is
supplied by the high-pressure facility air system (90 psig). The off-gas first passes through the
scrubber tank, where it mixes with liquid in the scrubber and forms a two-phase flow in the line to
the separator tank. From the separator tank the off-gas goes to the vessel vent system. The liquid
in the separator tank is circulated back (pumped) to the scrubber tank.
The design incorporates a jet bypass line leading to the vessel vent system (see Figure D.1.1).
This bypass line contains an automatic valve (normally closed during operation of the jet) that is
electrically interlocked to a high pressure switch. Also shown is a rupture disk located just off the
separator tank. Note that for simplicity, Figure D.1.1 shows only those instrumentation and
control features in the system that are referred to below.
D.1.1.1 Identifying potential criticality scenarios - logic diagram. In accordance with paragraph
5.7.5, "Identifying Potential Criticality Scenarios," a logic diagram is constructed (see Figure
D.1.2) as an aid to systematically identify the various scenarios that could lead to the accidental
addition of liquid to the dry location, which is the mechanism for a potential criticality accident in
this case. The logic diagram shows two credible liquid sources: Source 1 is liquid coming from
the scrubber system; and Source 2 is liquid from manual addition to the cabinet (operator error).
Pursuing Source 1 (Figure D.1.2), three basic phenomena are identified: (1) back siphonage, (2)
backflow resulting from a pumping action, and (3) backflow resulting from high pressure in the
scrubber system. For the high-pressure case, two initiating events are identified: (1) eructation,
and (2) pluggage of the air jet at the exit resulting in high pressure facility air (90 psig) applied to
the scrubber system (Figure D.1.1). As shown in Figure D.1.2, back siphonage and eructation are
judged to be incredible for this particular design and associated operating conditions. The
pumping action case is identified in Figure D.1.2 as worthy of study, but it is not developed here
(for simplicity). The potential criticality scenario designated for study below deals with pluggage
of the air jet. This scenario is highlighted in Figure D.1.2 and may be summarized as follows:
Potential criticality scenario - Mechanism: liquid addition to the dry location - Source:
scrubber system liquid - Phenomenon: backflow due to high pressure in the scrubber
system - Initiating event: pluggage of air jet at the exit.
D.1.1.2 Evaluation against the Double-Contingency Principle.
D.1.1.2.1 Identifying the two barriers for double-contingency. Simply stated, the Double-
Contingency Principle says that two independent, controlled barriers should exist to prevent
occurrence of a potential criticality accident scenario. The application of this principle is shown
symbolically in Figure D.1.3, which is a duplicate of Figure D.1.2, with the two barriers added.
For this example, it is assumed that the two barriers chosen are (1) pressure relief via the jet
bypass pressure/interlock system, and (2) pressure relief via the rupture disk. As illustrated in
Figure D.1.4, with these barriers in place, this potential criticality scenario requires the occurrence
of all of the following: (1) the initiating event - jet plugged at exit, (2) the failure of Barrier 1 -
failure to relieve pressure via the jet bypass pressure/interlock system), and (3) the failure of
Barrier 2 - failure to relieve pressure via the rupture disk.
D.1.1.2.2 Qualification of the barriers for double-contingency
As discussed in paragraph 5.7.7, it is important that the failure of a barrier for double-contingency
be an unlikely event. The determination of whether a failure of a barrier for double-contingency is
unlikely may be made on the basis of engineering judgment or failure rate data, if available. For
this example, assume that failure rate data are available. In accordance with paragraph 5.7.7.3, the
guidelines for acceptability when quantitative data are available are: (1) Guideline 1 - the
estimated probability that the barrier will fail is no greater than once in 100 demands or
0.01/demand, and (2) Guideline 2 - the product of {the estimated frequency of the initiating event}
times {the estimated probability of failure of the barrier - as applied in Rule 1} is not greater than
once in 10 years or 0.1/year.
Figure D.1.1. Schematic of dry location and scrubber system.
Figure D.1.2. Logic diagram for potential criticality via liquid addition to dry location.
Figure D.1.3. Logic diagram for potential criticality via liquid addition to dry location - two
barriers added.
Figure D.1.4. Contingency analysis diagram. Guideline 1 - As shown in Figure D.1.4 and Table D.1.1, the probability that Barrier 1 will fail
upon demand is estimated at 0.005/demand, thus meeting the 0.01/demand guideline.
Correspondingly, the failure probability of Barrier 2 is estimated at 0.003/demand, which is better
than the 0.01/demand guideline.
Both barriers are judged to meet Guideline 2. The frequency of the initiating event -- pluggage of
the air jet at the exit during operation -- is estimated (conservatively) to be in the vicinity of once
every twenty months (based on previous experience with similar equipment and operating
conditions). Therefore, the frequency is shown as 0.6/year (Figure D.1.4 and Table D.1).
For Barrier 1: (estimated frequency of the initiating event) times (estimated probability of failure
of Barrier 1) = 0.6/year 0.005/demand = 0.003/year, thus meeting the 0.1/year guideline.
For Barrier 2: (estimated frequency of the initiating event) times (estimated probability of failure
of Barrier 2) = 0.6/year 0.003/demand = 0.0018/year, thus meeting the 0.1/year guideline.
Note: As a point of interest, in this example the estimated frequency for this potential criticality
scenario (based solely on the three factors discussed above) is:
0.6/year 0.005 0.003 = 9 10-6/year,
that is a recurrence interval of approximately 111,000 years.
Independency of barriers. The two barriers in this example are judged to be sufficiently
independent. On the negative side, both barriers involve the sensing of a common process
parameter, high pressure, and both have the same basic function, which is to relieve pressure.
However, on the positive side the two barriers do not share components, and they operate quite
differently -- not likely to be subject to common-cause errors during facility operations such that
both systems would be inadvertently taken out of service, or in maintenance operations such that
common calibration or set-point errors might occur.
D.1.1.3 Identifying the means of control for each contingency barrier. As discussed in paragraph
5.7.7.4, the prominent identification of the means of control associated with a barrier for double-
contingency is important. Special care should be exercised to maintain these controls during
facility operation, maintenance activities, and subsequent design changes. As shown in Figure
D.1.4 and Table D.1.1, five controls are associated with Barrier 1 (see bottom of Figure D.1.4).
The failure of any one of these could defeat the barrier. Three of the five are hardware items.
They are the sensor, the electrical interlock, and the automatic valve. All three will require
administrative controls in the form of functional testing and preventive maintenance to maintain
high reliability. The other two controls (of these five) will require special procedural controls
(such as verification that the manual valve in the bypass line is OPEN prior to operating the air
jet). Only one means of control is associated with Barrier 2, that is, the rupture disk itself).
D.1.1.4 Review, relative to the other nuclear criticality safety objectives. The last step in the
double-contingency analysis is to reflect back on the design relative to all six of the basic design
objectives discussed in paragraph 5.7.3, particularly the following two objectives.
Objective 3: Is there a feasible design alternative that will completely eliminate this
potential criticality scenario? In this example the possibilities may include design
alternatives to (1) eliminate the use of liquids in the auxiliary systems to the dry location
(probably not practical here), or (2) eliminate the 90-psig motive force (in favor of an
alternative). Table D.1.1. Contingency Analysis - Summary Sheet
STATEMENT OF CRITICALITY SCENARIO
Specific Location: Dry Location
Mechanism: Liquid addition to dry location
Source: Scrubber system liquid
Phenomenon: Backflow due to high pressure in scrubber system
Initiating Event: Pluggage of air jet at exit.
INITIATING EVENT Pluggage of air jet at exit - estimated frequency, approx. 0.6/year
BARRIER 1
DESCRIPTION: Relieve (high, abnormal) pressure via jet bypass pressure/interlock system.
QUALIFICATION OF BARRIER 1:
Guideline 1: Estimated Probability of barrier failure - 0.005/demand.
Guideline 2: Product of (est. freq. of initiating event) times barrier failure prob. = 0.6/year
0.005 = 0.003/year.
LIST OF ASSOCIATED MEANS OF CONTROL:
1. High-pressure switch - separator tank (open jet bypass valve at >4 psig).
2. Electrical interlock - interlocks pressure switch to automatic valve in jet bypass line
to OPEN on demand.
3. Jet bypass valve (automatic) in jet bypass line.
4. Manual valve in jet bypass line - requires administrative controls to ensure valve open.
5. Vent line to vessel vent system - requires administrative control to ensure/verify that
line is free. (Note: liquid overflow line to serve as backup.)
BARRIER 2
DESCRIPTION: Relieve (high, abnormal) pressure via the rupture disk on separator tank.
QUALIFICATION OF BARRIER 2:
Guideline 1: Estimated Probability of barrier failure - 0.003/demand
Guideline 2: Product of (est. freq. of initiating event) times barrier failure prob. = 0.6/year
0.003 = 0.0018/year.
LIST OF ASSOCIATED MEANS OF CONTROL
1. Rupture disk on separator tank (rupture pressure >6 psig)
Objective 1: If feasible, have the preferred methods been incorporated? For example, the
use of geometry control in the dry location (if feasible) could eliminate the necessity of
precluding liquids from the dry location for reasons of nuclear criticality safety.
D.2 Examples of eliminating unnecessary criticality scenarios. Rather than accepting an element
of risk, it is preferred that the risk be removed entirely, if feasible. As discussed in paragraph
5.7.7, an effort should be made to explore the feasibility of design changes aimed at eliminating
potential criticality scenarios. The three examples below are intended to illustrate the intent and
lines of inquiry.
D.2.1 Example # 1 - Removing a potential water source to a dry area. A design concept
incorporates a water-cooled heat exchanger to cool the off-gas from a process. Evaluations reveal
a potential criticality scenario that begins with cooling water leaking across the tubes of the heat
exchanger (the initiating event), followed by the loss of detection and protective measures, and
ending with water reaching a location that must remain dry for nuclear criticality safety.
Before accepting this risk, consideration should be given to the feasibility of alternative cooling
means that will completely eliminate this scenario. For example, it may be feasible to provide the
off-gas cooling function using a design that does not involve water, such as with an air-cooled or
freon-cooled design. Using an alternative cooling method, the potential source of water to the dry
location is entirely eliminated.
D.2.2 Example # 2 - Eliminating the motive force. A design concept incorporates an air jet
connected to a process vessel to be used for the vacuum transfer of liquids into a vessel. The air
jet is supplied by a high-pressure facility air system. Evaluations show a potential criticality
scenario starting with pluggage of the exit to the jet with trash or other material, which produces a
high positive pressure in the process vessel. In turn, the high pressure provides a motive force
causing liquid in the vessel (containing fissile nuclides) to accidentally backflow through
interconnecting piping and reach locations that are unsafe for criticality, such as instrument air
systems, cold feed tanks, and ventilation systems.
In such a case, the feasibility of alternative design concepts, such as an electrically driven pump or
alternative system, should be explored that, while retaining the solution transfer capability, have
no potential for producing large positive pressures on the vessel contents.
D.2.3 Example # 3 - Eliminating the potential for over-concentration. A design concept
incorporates an evaporator for concentrating aqueous solutions containing fissionable material
product. Nuclear criticality safety of the evaporator is based on limiting the concentration of the
fissionable material product to a safe value. An automatic control system is used to regulate the
specific gravity of the concentrate. (The specific gravity can be directly correlated to product
concentration levels.) Backup protection against product over-concentration is achieved using
active protective devices (sensors and interlocks) that shut off the steam supply to the evaporator
when the specific gravity of the concentrate approaches the limit for nuclear criticality safety. A
potential criticality scenario is identified that begins with the loss of specific gravity control,
followed by failure of the active protective devices to shut off the steam supply, and resulting in
high product concentration levels exceeding the nuclear criticality safety limits.
In this case, design considerations should be given to identifying a feasible means to eliminate the
possibility of product over-concentration. For example, the circumstances may permit using a
value for the steam supply pressure to the evaporator that is high enough to achieve the normal
product concentration level but low enough to thermodynamically preclude the evaporator system
from being capable of attaining the higher product concentration levels associated with nuclear
criticality safety concerns. With this approach, a criticality accident due to product over-
concentration is not possible, regardless of the proper performance of the control and protective
devices.
D.3 Examples of passive-engineered features and devices. The purpose of this section is to
provide examples of the group of controls called passive-engineered features and devices, that are
discussed in paragraph 5.7.4.1.1. This group consists of fixed, passive design features and devices
with no moving parts. No electrical, mechanical, or hydraulic action is required. In many cases,
these features and devices are employed to protect against the unwanted transport of liquids from
favorable to unfavorable locations.
D.3.1 Air break. An air break is a simple, highly reliable means for backflow or back siphonage
prevention with virtually no failure mechanisms. With this device, an air gap is created by
interrupting a piping system. This device is illustrated in Figure D.3.1 and is applicable to
situations where line pressure may be broken. Note that such a device would rank very high as a
preferred control for nuclear criticality safety considering reliability, range of coverage, and
operational support requirements. Regarding range of coverage, this device provides direct,
positive protection against backflow to the feed tank in Figure D.3.1 -- independent of the reason
for the backflow. For these reasons, the air break should be employed as standard practice,
whenever applicable.
D.3.2 Barometric seal leg. Figure D.3.2 illustrates the use of barometric seal leg connections, or
gooseneck connections, when there are multiple-source line connections to a main header. Here, a
gooseneck connection is used for each source connected to the header. The arrangement shown in
Figure D.3.2 includes overflow capability from the header and acts to prevent liquid that has
arrived to the header (from one line source) from back-flowing through other line source
connections. Of course, undetected pluggage of the overflow line could defeat the safety function.
Because of its simplicity and effectiveness, this arrangement should be incorporated whenever
backflow from a header through a source line could introduce nuclear criticality safety concerns.
D.3.3 Criticality drain. A criticality drain is a device that normally serves both radiological and
criticality safety functions while preventing liquid buildup in moderation controlled enclosures
such as gloveboxes. Figure D.3.3 illustrates the use of a J-trap type criticality drain. The portion
of the drain inside the glovebox is raised slightly above the bottom and has a baffle to prevent
clogging (some types use screen mesh stand-offs). Thus, the maximum credible depth of liquid in
the glovebox is a fraction of minimum critical thickness. The portion of the device below the
glovebox is partially filled with an oil selected for its low evaporation rate and resistance to
combustion. This oil forms a radiological seal, and this region of the device may be transparent or
have a level indicator and fill port. The end of the J-trap may be open or connected to vented drain
piping based upon radiological considerations. In the event of a spill or leak exceeding the inside
lip height, liquids pass through the trap. The J-trap and any connecting piping are large enough in
diameter to accommodate the maximum credible flow rate into the glovebox. If the drain(s) are
piped to receiver vessel(s), they shall be criticality-safe and equipped with overflow lines to avoid
backups.
Figure D.3.1. Schematic of air break. Figure D.3.2. Illustration of barometric seal.
Figure D.3.3. Criticality Drain.
D.3.4 Nuclear safety blank. A nuclear safety blank is a simple, positive means for preventing the
accidental transfer of liquid through a line to an unsafe location. This blank typically consists of a
flat, solid metal disk inserted in a pipe flange to block the flow of liquid in special circumstances,
such as special processing campaigns, where the accidental transfer of liquid through the line to
another location could lead to criticality concerns. The device should be designed to make
unplanned removal mechanically difficult and labeled for easy identification in the field. A
spectacle flange is a nuclear safety blank combined with a second disk with flow hole(s) and
resembles a pair of eyeglasses. This design provides flexibility while having the advantage of
providing positive proof that flow is blocked if the disk with the hole(s) is visible. However, all
nuclear safety blanks should be leak tested and surveyed for wear and corrosion at start-up and at
appropriate intervals. With suitable administrative controls to guard against unplanned removal,
these devices would likely qualify as a double-contingency control, whereas administrative
controls to keep a block valve in the closed position would not qualify.
D.3.5 Large line sizes. Under certain conditions, the pluggage of a line can cause the unplanned
redirection of liquid to an unsafe location. By selecting a large, but safe, line-size larger than
would otherwise be employed, it may be possible to make pluggage of the line considerably less
likely to occur than would otherwise be the case.
D.3.6 Restricting orifices. Under certain conditions, the occurrence of an abnormally high flow
rate in a line can lead to a criticality concern. In such a case, a restricting orifice in the line can
provide a simple, reliable means of protection.
D.3.7 Relative elevation. The relative elevations of various equipment items and piping in a
facility can be an important consideration in determining the potential for the unplanned transport
of liquid from safe to unsafe locations. For example, simple leakage past a block valve can result
in the unplanned flow of liquid (by gravity) from a source tank to a receiving tank located at a
lower elevation. This mode of unplanned transport (by gravity) can be eliminated in the design
concept by reversing the respective elevations of the two tanks.
These examples serve to illustrate the importance of clear identification of those design features
and controls important to nuclear criticality safety. Many of the design features and devices in this
group, such as a restricting orifice or size of a line, are not normally associated with nuclear
criticality safety, and in the absence of clear identification, their importance to nuclear criticality
safety may be overlooked.
D.4 Examples of active protective devices. The group of controls identified as active protective
devices is discussed in paragraph 5.7.4.1.2. These devices are characterized as add-on devices
involving moving parts, are designed to act upon demand, or are sensing devices. Many such
devices are electrical/mechanical. The first two examples below illustrate devices in this group
that are strictly mechanical (preferred to complex electro-mechanical systems unless there is a
demonstrable benefit from additional complexity).
D.4.1 Rupture disk. Phenomena causing abnormally high pressure in a vessel can cause the
unwanted flow of liquid in the vessel to unsafe locations, as illustrated in the example in section
D.1.1. The normal engineering function of a rupture disk is to protect the vessel itself from over-
pressurization. However, it may be feasible in a given situation to select a lower pressure rating
for the rupture disk (than would otherwise be needed) to limit maximum pressures in a vessel
below the values required to transfer the liquid to an unsafe location. Assuming that adequate
reliability of this device can be established, the rupture disk would serve a valuable nuclear safety
function in addition to its vessel protective function.
D.4.2 Backflow prevention devices. As discussed in section D.3, an air break provides very
effective protection against backflow and back siphonage. However, there are situations where an
air break device is not suitable, since line pressure would be lost. When it is necessary to maintain
line pressure, an in-line device may be considered. A review of the various backflow and back
siphonage prevention designs could include: (1) single check-valve design, (2) double check-
valve design, (3) double check-valve design with vent, (4) reduced-pressure device, and (5)
reduced-pressure device with internal air gap. This spectrum of design types serves to illustrate
the general notion involved in selecting a double-contingency means of control. Due to questions
of seal integrity, it is likely that most of these backflow prevention devices would be determined to
have insufficient reliability to qualify as a double-contingency means of control. On the other
hand, one or more of these designs may so qualify, depending on unique design features and the
service conditions involved.
D.4.3 Radiation monitoring systems. A radiation detector, readout, alarm, and associated motor-
or air-operated valve(s) that close(s) on a dose rate set point is a relatively simple active protection
system for either radiological safety, criticality safety, or both. Such systems may be portable or
fixed. They can be conservatively used for criticality safety of geometrically unfavorable tanks by
assuming that all radioactivity is due to the presence of fissionable nuclides. More sophisticated
applications use gamma spectrum analyzers to more accurately estimate fissionable material
content. In-line detectors should be close to the lower side of piping at strategic points to
maximize detecting solids-buildup that sampling may miss, and detectors on tanks should be
located near places where solids-buildup is most likely. A variant of this system is a soluble
neutron poison monitoring system where increasing neutron flux means poison concentration is
decreasing. However, these detectors should not be located under pipes or tanks because poison
can precipitate and skew results. Regardless of the specific application, it is important to design,
operate, and maintain such systems to avoid frequent false alarms and thus create a distrust of
instrument readings and alarms. APPENDIX E. SOFTWARE CONFIGURATION CONTROL PROCEDURE
This appendix provides an acceptable approach to ensure that the nuclear criticality safety software
system used in support of contractor installation nuclear criticality safety organization(s) will
provide accurate and reliable results,
provide rigorous structure to implement software changes, and
prevent unauthorized changes to the software.
Included in this appendix are the actions and responsibilities for maintaining the quality and
integrity of the nuclear criticality safety software system used in support of the contractor
installation nuclear criticality safety organization(s). Except when specifically included in a
Software Catalog, vendor-supplied systems software, such as operating systems, linkers,
compilers, and data base management systems used by the contractor installation, are excluded
here and covered by separate configuration control for which the contractor is responsible.
E.1 Specific responsibilities.
E.1.1 Contractor safety organization manager. The contractor safety organization manager
acts as or appoints a Software System Team Chairperson;
assumes overall responsibility for the configuration control of the Contractor
Nuclear Criticality Safety Software System;
maintains membership and charter on the Software System Team through
coordination with the Contractor Nuclear Criticality Safety Committee (See Form
E.7 for charter);
schedules and coordinates annual surveillance of the software configuration control
program;
requests a surveillance/audit of configuration control for software utilized for
nuclear criticality safety computations once every five years;
maintains a current listing of authorized users as notified by System Administrator;
distributes pertinent information on the software changes, Software Catalog,
validations, and other sources to authorized users as appropriate;
participates, coordinates, and manages the handling and resolution of Software
Revision Reports and Software Nonconformance Reports as prescribed in this plan;
and
maintains hard copy documentation for a retention period consistent with paragraph
2.1.2 for
- Software Configuration Control Plans,
- Software Catalogs (Form E.3),
- Software Revision Reports (Form E.1),
- Software Nonconformance Reports (Form E.2),
- Request for User Access (Form E.4),
- Audit and Surveillance Reports, and
- Software System Team Charter and Membership.
E.1.2 Software System Team. The Software System Team
by majority, determines those development, verification, testing, and record
keeping operations to be covered by the Configuration Control Plan and the access
controls to be required,
when a new software system is believed to be ready for use, reviews and approves
the Software Catalog for completeness and correct access control,
develops the requirements for software Verification and Configuration Control
Tests, coordinates the performance of the required tests, and approves all new or
revised software before production use,
ensures that documentation has been updated (e.g., Configuration Control Plan,
Software Catalog, Access Control, records of Verification and Configuration
Control Tests, etc.),
upon request, assists the "quality organization" and other organizations in
performing software appraisals, audits, and surveillances,
when a change to the software is requested, reviews the Software Change Request,
Software Revision Report, Part A (Form E.1), to decide if and when the change
should be made and completes Parts B and C, as appropriate,
reviews Software Nonconformance Reports (Form E.2) and determines and
documents resolution by a majority in agreement, and
develops, implements, and maintains a Software Disaster Plan, as appropriate
(Form E.5).
E.1.3 Functional System Manager. The Functional System Manager
serves as the principal Nuclear Criticality Safety Organization contact to the
software user with regard to the content of the software,
provides notification to installation software users of changes to the software
systems, nonconformance reports, specialized machine dependent job control
language (JCL) requirements, and current Software Catalog, and, if serving as the
lead contact for more than one installation, maintains communication with each
installation represented,
participates in the handling and resolution of Software Revision Reports and
Software Nonconformance Reports as prescribed in this plan,
ensures that a Software Catalog is prepared for each mainframe computer software
application, is kept current, and a copy provided to each user,
verifies correct version of software is transferred into Migration Storage Area from
the Development Storage Area and performs or coordinates the Software
Verification Tests, and
upon approval of the change for production use, ensures that the version
identification of any departmental procedure or plans that reference the software by
version are updated.
E.1.4 System Administrator. The System Administrator
ensures that master copies of the previous versions of machine executable modules
and source code are maintained in the Archive Storage Area, and that a hard copy
listing and documentation of the latest version are maintained,
retains a copy of all Software Revision Report (Form E.1) forms,
prepares the Software Catalog and sends a copy of each updated catalog to all
members of the Software System Team,
notifies the Software System Team that programming of a requested revision is
complete and has been transferred to the Migration Storage Area for Verification
Testing,
checks the Software Revision Reports and supporting documentation for
completeness and forwards the report to the Software Developer,
performs the transfer of software to the Production Storage Area and Archive
Storage Area when all proper tests and approvals authorize the transfer,
verifies and ensures the proper version of the executable code is in the Production
Storage Area and the most recently superseded version of the source and executable
code is stored in the Archive Storage Area,
develops, implements, and maintains the Configuration Control testing of the
software production version and maintains appropriate documentation of testing,
develops, implements, and maintains a NCS Software Programmer's Manual to
document the procedure used in transferring, compiling, and otherwise using the
software, and
subject to the Software System Team Chairperson's approval, procures and
maintains computer equipment to perform archiving and testing responsibilities.
E.1.5 Installation nuclear criticality safety organization. The installation nuclear criticality safety
organization
ensures all users of the NCS Software System utilize software that is covered by
this Configuration Control Plan for mainframe computations,
ensures the computer software contained in the Software Catalog (Form E.3) is
properly validated for the intended use,
assists in the performance of Verifications and Configuration Control tests, as
necessary,
authorizes access to the software covered under this plan for users in the installation
Criticality Safety Organization and other contractors per the User Access form
(Form E.4), forwards completed User Access forms to the Software System
Administrator, and provides notification to the Software System Administrator
when user access needs to be removed,
develops and implements Disaster Plans where appropriate and forwards a copy of
these plans to the Software System Team, Form E.5,
ensures that each user granted access to the software is provided with training in the
proper use of the software,
develops and implements the appropriate Quality Assurance and Quality Control
Programs to ensure the correctness of calculational results and use of the software,
assists the Software System Team in implementing software changes, testing new
software, user access control, and any other areas where appropriate,
may request changes by initiating the Software Revision Report (Form E.1) in order
to define modification requirements, and
reports problems encountered to the proper Functional System Manager using the
Software Nonconformance Report (Form E.2).
E.1.6 Software developer. The software developer
makes ONLY those software changes that have been approved by the Software
System Team on a Software Revision Report (Form E.1),
may propose software changes on a Software Revision Report,
updates software version identification in a program when changes are made,
assists the Software System Team in conducting the Verification Test of the
software modification,
supplies information to the System Administrator on software version identification
and software changes, as appropriate, and
works with Software System Team to update the supporting documentation.
E.2 Software identification. Initial system configuration consists of a catalog of application
specific software. This Software Catalog defines the baseline system configuration. Access
control is established by the Software System Team and is maintained by the System
Administrator. Unambiguous labeling should provide traceability from source modules to
executable modules (Form E.6).
Versions should be uniquely identified in such a way that the update sequence may be readily
determined. The version number and revision number shall be listed at least once on all output.
E.3 Software control. Users of software are responsible for ensuring that any software used is the
currently approved version and that the use and application is validated.
All modifications to the nuclear criticality safety software system require the approval of the
Software System Team using the procedure in section E.4 of this plan.
The software residing in the Production Storage Area will be audited by the Quality Division to
ensure that the correct version is in use and that no changes have been made.
Hard copy computer printouts should have, printed on a header, the version and date of revision of
the principal software unit generating the printout.
All modifications of software will be acceptance tested as specified on the Software Revision
Report.
E.4 Software change procedure. A software change is initiated by any user by completing Part A
of the Software Revision Report.
The request is sent to a member of the Software System Team.
The Software System Team Chair/Functional System Manager/Software Administrator transmits
the report to the other members of the Software System Team, as needed, to determine if and when
the change should be made.
Approval or rejection is documented by completing Part B of the Software Revision Report. If the
modification is to be made, the Verification Test Plan should be developed and documented on the
Software Revision Report, Part B. NOTE: The level of detail in the Verification Test is
determined by the Software System Team based on the extent of the software change and the
consequences of unintended or unanticipated changes. The Software Revision and associated
Verification Test are approved by the Software System Team by signing the appropriate spaces on
the form. If the Software Revision Report is rejected, the Software System Team Chairperson
provides an explanation for rejection and provides a copy to the requestor.
A copy of the approved Software Revision Report (Parts A and B) is sent to the System
Administrator. The System Administrator provides the Software Developer a copy of the current
source code.
The software modifications are made in the Development Storage Area. Once the software
modifications have been made to the satisfaction of the Software Developer and the System
Administrator, the software is transferred to the Migration Storage Area by the Functional System
Manager. Part C of the Software Revision Report documents the completion of this step.
The Verification Test is performed in the Migration Storage Area by the Functional System
Manager with assistance, where appropriate, from the installation NCS Organizations.
The performance of the software in the Verification Test is evaluated by the Software System
Team. Part D of the Software Revision Report documents the Verification Test results and the
acceptance/rejection of the results by the Software System Team.
Software System Team approval of the Software Revision Report, Part D, provides notification to
the System Administrator to transfer the new version of the software into the Production Storage
Area and a copy of the current version (source and executable code) to the Archive Storage Area.
Completion of Part E of the Software Revision Report documents the software transfers, bit-by-bit
comparison of the new Production version, and completion of the software revision procedure.
E.5 Nonconformance Report procedure. A Nonconformance Report is initiated by completing
Part A of the Nonconformance Report (Form E.2).
The request is sent to a member of the Software System Team.
The Software System Team Chair/Functional System Manager/Software Administrator transmit
the report to the other members of the Software System Team, as needed, to determine the actions
to be taken to prevent recurrence of the nonconformance.
The Software System Team Chair provides nonconformance notification to the Quality Assurance
Division and the Occurrence Reporting System, as appropriate.
In extraordinary cases, the System Administrator or the Software System Team Chairman may
authorize shutting down a program that presents immediate and major danger to safety or the
environment. In such cases, the Software System Team should authorize the use of the corrected
software, full details of the incident shall be provided in the documentation for the change, and a
Nonconformance Report shall be initiated. The changed software shall have a new version
identification.
E.6 Software testing. Configuration Control Test: Testing procedure, requirements, and plan are
determined by the Software System Team. At a minimum, the Configuration Control Test should
include (a) a periodic (every quarter) bit-by-bit comparison of the production version against an
archived production version stored at the time the production version was installed, and (b)
quarterly testing by each installation using installation specific validation cases. Documented
records of these tests shall be maintained by the System Administrator.
Verification Test: Testing procedures, requirements, and plans are determined by the Software
System Team pursuant to section E.4 of this plan. The level of detail found in the test plan will be
commensurate with the complexity of the software change. As part of a Software Change Request
implementation, transfer tests will be performed to verify the copying and transferring of software
from one computing platform to another computing platform as listed in the software catalogs. Form E.1 Software Revision Report.
SAMPLE
Part A - Request for Software Change
(to be completed by Software User/Developer)
Report No.
SRR-
Reason for the requested change and Software Nonconformance Report No.(SNR- ):
Description of requested change:
Modules affected:
Describe anticipated or known effects the change will have on:
A. Sample problem results
B. Calculational time/efficiency
C. Existing documentation
Name of requestor and signature:
Date:
Part B - Software System Team Approval/Rejection
(to be completed by Software System Team)
(approval requires four affirmative signatures from Software System Team)
Approval
Rejection
Functional System Managers
System Administrator
Software System Team Chairperson
Reason for rejection:
Software Verification Test Plan attached?_____ Page 1 of 2
Form E.1 (cont.)
Part C - Software Change Documentation
(to be completed by Software Developer and System Administrator)
Describe the change and components affected
File names for new source or data:
Describe the results of the Software Developer testing performed:
Does the change affect existing documentation? If so, update and attach new documentation.
Software change completed
Software Developer _____________________ Date ___________
System Administrator _____________________ Date ___________
Software transfer: Development Storage Area to Migration Storage Area by Functional System Manager
SYS01 _____________________ Date __________ SYS03 _____________________ Date __________
SYS02 _____________________ Date __________ SYS04 _____________________ Date __________
Part D - Software Verification Test Evaluation (verification results attached)
(to be completed by System Software Team)
Verification tests results accepted and permission granted to transfer software from Migration Storage Area to
Production Area
Functional System Manager _____________________ Date ___________
Functional System Manager _____________________ Date ___________
Functional System Manager _____________________ Date ___________
System Administrator _____________________ Date ___________
Software System Team Chairperson _____________________ Date ___________
Part E - Software Change Implementation in Production Storage Area
(to be completed by System Administrator)
Page 2 of 2
Form E.2 Software Nonconformance Report.
SAMPLE
Part A - Report of Software Nonconformance or Error:
(to be completed by Software user)
Report No.
SNR-
Software user name and address:
Software title/version/date:
Description of software nonconformance or error:
Cause of nonconformance or error:
Effect on previous calculations:
Recommended corrective action:
Part B - Software Nonconformance Assessment and Action Plan
(to be completed by Software System Team)
Cause of nonconformance and effect on previous software users:
Immediate action is required to stop use of software? _____
Reportable event per Occurrence Reporting System? _____
Recommended corrective action:
Software System Team approval of recommended corrective actions:
Functional System Manager _____________________ Date ___________
Functional System Manager _____________________ Date ___________
Functional System Manager _____________________ Date ___________
System Administrator _____________________ Date ___________
Software System Team Chairperson _____________________ Date ___________
Corrective actions completed
Software System Team Chairperson _____________________ Date ___________
Form E.3 NCS Software System Version No. 1 Catalog
Computer node
Updated:
SAMPLE
Module
Date
Revision
Data set Name
Form E.4 Request for User Access.
User access is requested for the following Contractor Nuclear Criticality Safety software:
The proposed user and their supervisor have been informed and understand that validation,
(establishment of correctness or bias in calculated results) is a user responsibility and that the contractor
makes no claim of correctness for the computer software or for computer calculations performed by
others.
Type Proposed User's Name and UID:_________________________________________
Proposed User (Signature): _____________________________________ Date: ____________________
User's Address: ________________________________________ User's Phone #: __________
User's Supervisor (Signature): __________________________________ Date:____________________
Organization: ____________________________________________
Installation Nuclear Criticality
Safety Organization Head (Signature): ____________________________ Date:____________
SEND COMPLETED FORM TO:
____________________________________________________________________________________
(TO BE COMPLETED BY SOFTWARE SYSTEM ADMINISTRATOR)
User access was activated on this date: ___________________________________
System Administrator Signature: ________________________________________________________
Copy: Software System Team Chair Form E.5 NCS Software Disaster Plan.
A disaster plan is not necessary for the NCS software because of the redundancy provided by multiple
computing systems. The NCS software will be provided on the following systems, for example:
1. Computing System #1 I.D.
2. Computing System #2 I.D.
3. Computing System #3 I.D.
Therefore, it is judged to be incredible that all NCS software versions could be simultaneously
destroyed. Form E.6 Software Labeling Protocol Examples
Source
NCSS.ZAZ39461.Module.V#R###.FORT (.ASM)
Production Subroutine library
NCSS.ZAZ39461.Sublib.V#R###.LOAD
Archive Subroutine library
NCSS.ZAZ39461.Sublib.V#R###.ARCHIVE
Production Load Modules
NCSS.ZAZ39461.module.V#R###.S###.LOAD
Migration Load Modules
NCSS.YCR39461.module.V#R###.S###.LOAD
Archive Load Modules
NCSS.ZAZ39461.module.V#R###.S###.ARCHIVE
Data Libraries
NCSS.ZAZ39461.identification.V#R###.DATA
MODULE = program name (such as KENOVA, CSAS25, and SUBLIB)
V# is the nuclear criticality safety software version number.
R### is the module revision number.
S### is the subroutine library revision number.
Form E.7 NCS Software System Team (NCSSST) Sample Charter.
Objective: The nuclear criticality safety software system team (NCSSST) acts as the change control
board for the company's Nuclear Criticality Safety Software. The team should:
maintain the company's Nuclear Criticality Safety Software Configuration Control
Plan,
determine and implement necessary changes to the NCS software pursuant to the
configuration control plan,
address NCS software nonconformance reports as appropriate, and
provide assistance to other organizations in the area of software configuration
control.
Mtng Freq: At the discretion of the team (minimum - once per year)
Team
Membership: Chairperson
Contractor Central Safety and Health organization or designee
System Administrator
Computer, hardware or software maintenance/operations organization
Installation Functional System Manager(s)
Installation representative(s)
Reporting: The NCSSST is directly accountable to the Contractor Central Safety and Health
organization. APPENDIX F. EXAMPLE COMPUTATIONAL TECHNIQUE VALIDATIONS
This appendix provides more detailed descriptions and example implementation of the required
elements for computational technique validation as described in paragraphs 5.8.4 and 5.8.5, i.e.,
(a) the selection and description of the critical experiments used in the validation, or an
appropriate reference that describes the experiments in adequate detail to permit
reconstruction of computational input,
(b) the selection and description of the computational method that is to be validated along with
any necessary data for performing calculations or comparisons (e.g., neutron cross
sections, material bucklings, limiting surface densities, or other similar data),
(c) the selection and description of the computer/calculator platform and associated operating
system used in the validation,
(d) the nuclear properties, such as cross sections, that should be consistent with experimental
measurements of these properties,
(e) a description of similarities and differences between the critical experiments and the
calculational models used for the validation,
(f) all geometric, material, and nuclear physics related input variables used for the validation of
the calculational or comparative method, with sketches provided,
(g) the basis for the calculational or comparative bias and the determination of an acceptance
criterion for calculated subcritical results, and
(h) establishing the areas of applicability of the calculational or comparative bias and the
acceptance criterion developed from the validation effort.
F.1 Selection and description of critical experiments. The selection and description of critical
experiments used as benchmarks for the calculational method validation should be similar to, and
representative of, the problems that are to be evaluated. The benchmarks' physical compositions,
geometric configurations, and other nuclear characteristics should be reviewed to ensure applicability
(similarity) to the future problems for which the validation is intended. Unfortunately, critical
experiments available for benchmarking tend to emulate single units. A particular problem evaluation
may require calculations for a single unit, as well as arrays of units (such as in fissionable material
package or storage array evaluations). Such a problem poses a difficulty in benchmark selection
because there is a paucity of critical experiments of large arrays. Because of these concerns, it may
be necessary to model a wide variety of benchmark experiments to adequately assess the validity of
the calculational method used in the evaluation. Sufficient numbers and quality of experiments
should be selected to provide a statistically justifiable basis for subcritical acceptance criteria.
F.2 Selection and description of the computational method. The selection of the computational
method should be related to the particular expertise and experience of the criticality safety specialist
performing the validation and should be related to the difficulty of the eventual problem evaluation
and the relevance of the benchmark data to the computational technique. Examples of calculational
methods are the three-dimensional multi-group or the pointwise cross section Monte Carlo codes,
KENO-V.a, or MCNP and VIM, respectively; the one-dimensional multi-group Sn discrete-ordinates
transport theory code, ANISN; the diffusion theory code, GAMTEC II - HFN; and hand calculation
methods such as the limited surface density, density analog, one- or two-group restricted three-
dimensional diffusion theory, or solid angle methods. Associated computational data (e.g., cross
section libraries, scattering quadrature sets, material bucklings, diffusion lengths, etc.) shall be
identified. The computational method and associated computational data shall be described or
referenced in the validation documentation. When computer neutronics calculations are used, the
type of computing platform should be stated along with relevant code configuration control
information. This information may be provided by reference.
An example partial listing of computer codes, models, and hand calculational methods that
historically have been successfully used for nuclear criticality safety evaluations is provided in Table
F.2.
Table F.2. Partial Listing of Computer Codes, Models, and Hand Calculational Methods
Computer Codes
- R. N. Blomquist, "VIM Continuous Energy Monte Carlo Transport Code," Proceedings of the
International Conference on Mathematics and Computations, Reactor Physics, and
Environmental Analysis, Portland, OR, (April 30 - May 4, 1995). (On-line user's guide for VIM is
shipped with the code, available at RSICC).
- J. F. Briesmeister, Ed., "MCNP, A General Monte Carlo Code for Neutron and Photon Transport,"
LA-7396-M, Rev. 2, Los Alamos National Lab. (Sept. 1986).
- L. L. Carter, C. R. Richey and C. E. Hughey, "GAMTEC-II: A Code for Generating Consistent
Multigroup Constants Utilized in Diffusion and Transport Theory Calculations," BNWL-35, Pacific
Northwest Laboratory, (March 1965).
- N. M. Greene, L. M. Petrie, "XSDRNPM-S: A One-Dimensional Discrete-Ordinates Code for
Transport Analysis," ORNL/NUREG/CSD-2/V2/R1 (June 1983).
- J. R. Lilley, "Computer Code HFN; Multi-Group, Multi-Region Neutron Diffusion Theory in One
Space Dimension," HW-71545, General Electric Company, Richland, Washington, (July 1962).
- R. D. O'Dell, F. W. Brinkley, Jr., D. R. Marr, and R. E. Alcouffe, "Revised User's Manual for
ONEDANT: A Code Package for One-Dimensional, Diffusion-Accelerated, Neutral-Particle
Transport," LA-9184-M Rev. (December 1989). (On-line user's manuals for TWODANT are
shipped with the program source.)
- L. Petrie, N. Landers, "KENO-Va, An Improved Monte Carlo Criticality Program with Supergrouping,"
ORNL/NUREG/CSD-2/VI/R2 (Dec. 1984).
- W. A. Rhoades and R. L. Childs, "An Updated Version of the DOT 4 One-and Two-Dimensional
Neutron/Photon Transport Code," ORNL-5851 (July 1982).
- W. A. Rhoades and R. L. Childs, "The DORT Two-Dimensional Discrete Ordinates Transport Code,"
Nucl. Sci. Eng. 99, 1, 88-89 (May 1988).
- W. A. Rhoades and R. L. Childs, "The TORT Three-Dimensional Discrete Ordinates Neutron/Photon
Transport Code," ORNL-6268 (November 1987).
- C. R. Richey, "EGGNIT: A Multigroup Cross Section Code," BNWL-1203, Battelle Memorial Institute
Pacific Northwest Laboratories, Richland, Washington (November 1969).
- V. S. W. Sherriffs, "MONK, A General Purpose Monte Carlo Neutronics Program," SRD-R-86, United
Kingdom Atomic Energy Authority Safety and Reliability Directorate, Culcheth Warrington,
January 1978.
- T. P. Wilcox, E. M. Lent, "COG: A Particle Transport Code Designed to Solve the Boltzmann
Equation for Deep-Penetrating (Shielding) Problems." Draft Report, Lawrence Livermore
National Lab. (Oct. 1986).
Models and Hand Calculational Methods
- J.T. Thomas, "Solid Angle and Surface Density as Criticality Parameters," NUREG/CR-1615 and
ORNL/NUREG/CSD/TM-15, U.S. Nuclear Regulatory Commission (1980).
- J.T. Thomas, "Surface Density and Density Analog Models for Criticality in Arrays of Fissile
Materials," Nucl. Sci. Eng., 62, 424 (1977).
- M.C. Evans, "Criticality Assessment Using the Limiting Surface Density (NB2n) Method and
Examples of Application," BNFL SAG/80/P29, British Nuclear Fuels plc (1980).
- H.F. Henry, C.E. Newlon and J.R. Knight, "Extensions of Neutron Interaction Criteria," K-1478 (July
1961).
- F.G. Welfare, "A Comparison of the Solid Angle Technique with KENO IV Calculations," Trans. Am.
Nucl. Soc., 43, 410 (1982).
- C.E. Newlon, "Solid Angle-Interaction Potential Method: Illustrative Problems," K-L-6328 (Sept.
1973).
- D.R. Oden, J.K. Thompson, M.A. Lewallen, "Critique of the Solid Angle Method," NUREG/CR-005,
U.S. Nuclear Regulatory Commission (1978).
- D. C. Hunt, "Comparative Calculational Evaluation of Array Criticality Models," Nucl. Technol., 30,
190 (1976).
- S. J. Altschuler and C. L. Schuske, "A Model for the Safe Storage of Fissile Solutions," Nucl.
Technol., 17, 110 (1973).
- C. L. Schuske and S. J. Altschuler, "Models for the Safe Storage of Dry and Wet Fissile Oxides,"
Nucl. Technol., 19, 84 (1973).
- S. J. Altschuler and C. L. Schuske, "Models for the Safe Storage of Fissile Metal," Nucl. Technol.,
13, 131 (1972).
F.3 Description of similarities and differences. Nearly every computational model of a benchmark
experiment requires some modeling approximations. The computational model approximations of the
benchmarks should be described and include discussions on the similarities and bases of the
differences.
F.4 Input variables. The geometric, material, and nuclear physics related input variables used for the
validation of the calculational or comparative method should be provided along with sketches that
relate the benchmark to the computational model.
F.5 Acceptance criteria. The acceptance criteria are developed from the bias of calculated results
and the uncertainties of the experimental data, the calculational technique, and the calculational
models.
The basis for the calculational or comparative bias and the determination of acceptance criteria for
calculated subcritical results shall be provided. For nuclear criticality safety calculational method
validation purposes, the bias is defined as a measure of the systematic disagreement between the
results calculated by a method and experimental data. The usual method of determining the
calculational bias is to correlate the results of the benchmark critical experiments with the calculated
results of the code being validated. With a value of unity, keff = 1.0, for each benchmark critical
experiment, the bias is the deviation of the calculated values of keff from unity. The average bias is
usually determined by one of two methods: (1) taking the difference between a simple average of
the pooled calculated results and unity, that may be adequate for a specific validation, or (2) taking
the difference between a linear regression of the calculated results (as a function of some
independent variable, e.g., average energy group (AEG) of neutrons causing fission) and unity, that is
usually necessary for a global validation. The first method produces a single value for the bias, while
the second method produces a variable bias that is a function of the independent variable due to
trends. Generally, neither the bias nor its uncertainty is constant; both are typically a function of one
or more physical or nuclear variables. Physical variables include, for example, material composition,
density, and enrichment. Nuclear variables include AEG causing fission, ratio of thermal absorption
to total absorptions, ratio of total fissions to thermal fissions, fractional neutron leakage, and others.
To appropriately validate cases where the calculated flux is relatively large in the intermediate energy
range and small in the fast and thermal regions, the code user needs to use benchmark quality critical
experiments with similar AEG values and flux-energy group distributions.
Uncertainties in the validation calculations come from three general sources. The first source arises
from limitations associated with the critical experiment and inadequacies of determinations and
documentation. These can include uncertainties in the material and fabrication tolerance of the
experimental hardware and fuel (compositions, assays, masses, densities, dimensions), the
experimenter's manipulation or adjustments, or both, to obtain the reported data, and an inadequate
description of the experimental layout and surroundings. The second source is from the
computational method, that may include uncertainties in the mathematical equations solved, the
calculational approximations utilized in solving the mathematical equations, the convergence criteria,
the cross-section data evaluation process and the manipulation of cross-section data, and limitations
of the computer hardware. The third source is from the calculational models developed to emulate
the experiment. These include uncertainties because of material and dimensional modeling
approximations, the selection of various code options, individual modeling/coding techniques, and
interpretation of the calculated results.
For computational method validation purposes, it is usually not practical or necessary to quantify and
qualify all the individual uncertainties. The total uncertainty can be estimated through the application
of any valid statistical treatment of the data. The total uncertainty determined usually appears as the
bias and a variability in the bias, depending upon the statistical analysis applied. The combination of
the bias and uncertainty in the bias is deduced from the mean value being calculated to establish a
subcritical value, e.g., acceptance criteria. This subcritical value and any other values considered to
be less subcritical are taken to be critical within the confidence limits applied to the statistical
technique to determine the uncertainty. A margin of subcriticality should be deduced from the
previously described subcritical value to ensure subcriticality.
Where calculational methods of evaluation are used to predict neutron multiplication factors, the
calculated multiplication factor, ks, shall be equal to or less than an established allowable neutron
multiplication factor, upper subcritical limit; i.e.,
ks kc - ks - kc - km
where
ks = the calculated allowable maximum multiplication factor, keff, upper subcritical limit (USL),
of the system being evaluated for normal or credible abnormal conditions or events.
kc = the mean keff, that results from the calculation of the benchmark criticality experiments
using a particular calculational method. If the calculated values of keff for the criticality
experiments exhibit a trend with a physical or nuclear variable, then kc shall be determined
by extrapolation on the basis of a best fit to the calculated values. The criticality
experiments used as benchmarks in computing kc should have material compositions
(neutron poisons and moderators), geometric configurations, neutron energy spectra, and
nuclear characteristics (including reflectors) similar to those of the system being evaluated.
Generally neither the bias nor its uncertainty is constant; both should be expected to be
functions of composition and other variables.
ks = an allowance for
(a) statistical or convergence uncertainties, or both, in the computation of ks,
(b) material and fabrication tolerances, and
(c) uncertainties due to limitations in the geometric or material representations used in the
computational method.
kc = a margin for uncertainty in kc that includes allowance for
(a) uncertainties in the critical experiments,
(b) statistical or convergence uncertainties, or both, in the computation of kc,
(c) uncertainties due to extrapolation of kc outside the range of experimental data, and
(d) uncertainties due to limitations in the geometrical or material representations used in the
computational method.
km = an arbitrary margin to ensure the subcriticality of ks. The margin in the correlating variable,
that may be a function of composition and other variables, shall include allowances for the
uncertainty in the bias and for uncertainties due to any extensions of the areas of
applicability. A value for km should be described and documented.
F.6 Areas of applicability. An integral part of a code validation effort is to define the areas of
applicability for the validation. There are three conditions that must be satisfied to ensure that
calculations done to evaluate or support a real situation fall within the areas of applicability for the
validation of the calculational method being used. These are materials (and associated nuclear
properties), geometry, and neutron energy spectrum. Frequently, the correlating variable of AEG of a
neutron causing fission is used to define an area of applicability for the validation and related
computational bias. A discussion of the bases or judgments as to what constitutes the validation
areas of applicability should be provided.
The areas of applicability should identify the important variables and characteristics for which the
code was (or was not) validated. For example, the areas of applicability may include specific types
of fissionable materials (HEU, LEU, plutonium of low 240Pu content, or others), material form (solution
or metal, water-moderated or carbon-moderated, and others), geometric configurations (single units
or arrays, heterogeneous or homogeneous, dissimilar units, or other conditions), and reflector
materials (water, concrete, steel, lead, or others). The areas of applicability are intended to identify
specific limits (upper and lower) of the variable or characteristic used to correlate the bias and
uncertainties. For example, the areas of applicability may be defined in terms of the moderating ratio
like H:X = 10 to 500, or in terms of the average energy group causing fission such as an
AEG = 6.5 to 21.5, or in terms of the ratio of total fissions to thermal fissions like F:Fth = 1.0 to
5.0. For subsequent use of a validated code, the user should show that the variables and
characteristics of the problem being calculated fall within the areas of applicability defined during the
validation.
The areas of applicability of a calculational method may be extended beyond the range of
experimental conditions over which the bias is established by making use of correlated trends in the
bias. Where the extension is large, the method should be
(a) validated with a stepwise approach in developing a repertoire of benchmarks for the
purpose of identifying individual potentially compensating biases associated with individual
changes in materials, geometries, or neutron spectra, and
(b) supplemented by other calculational methods to provide a better estimate of the bias(es) in
the extended areas of applicability.
F.7 Example validation. This example describes a statistical technique used to establish the
maximum allowable calculated keff, acceptance criterion (also called the upper subcritical limit)
resulting from a computational method validation effort. Various elements of the technique are
derived from different references. This example provides more detail than what is provided in the
footnoted references, and the equations may be in a different, but algebraically identical, form. The
equations in this example are usually in a basic form, while other algebraically identical forms (not
presented here) are more convenient for computational purposes.
One method used to validate the KENO criticality code and associated cross sections for
establishing an acceptance criterion is to determine the single sided, uniform width, closed interval,
lower tolerance band, (LTB) for calculated keff values of critical systems. For application, this LTB
becomes the upper subcritical limit (USL) acceptance criterion. A system is considered acceptably
subcritical if a calculated keff plus two standard deviations lies below the USL, i.e.,
keff + 2 < USL:
For a set of n keff calculations of critical experiments with a corresponding independent variable x,
determine the linear least-squares fit, k(x), of the data as a function of x.
k(x) = b0 + b1 x, where
and
In these equations, and others to follow, a summation, , means the sum of all values from i=1 to
i=n, where n is the sample size, that is the number of critical experiments upon which the validation
is based. The independent variable, x, is used to specify the areas of applicability, as described in
section F.6.
The next step is to determine the "pooled" variance,
(the variance of the fit, or mean square error) =
(the within variance of the data) = and
i is the standard deviation associated with each calculated keff.
The pooled standard deviation is then the square root of the variance, .
The within-variance, , represents the contribution of the variance from KENO or other Monte
Carlo codes that have a standard deviation associated with the calculated keff values. For
deterministic codes that do not have a standard deviation associated with the keff values, the within-
variance is zero. It should be noted that the within-variance is not a part of the statistical method
presented in footnotes 110 and 111, but was included here because of the inherent uncertainty from
a Monte Carlo type code.
The next step is to determine a multiplier, C, of the pooled standard deviation such that there is at
least confidence that a proportion P of the population (of future calculations of critical systems)
will lie above the line defined by k(x) minus Csp. This is the LTB as determined by the technique,
and
LTB = k(x) Csp .
The confidence, that is selected by the validator, is defined by
(1 1) = the one-sided confidence band about the linear regression, and
(1 2) = the confidence on the variance of the fit.
Since the expression for presents one equation and two unknowns, either 1 or 2 must be selected
such that the other can be determined. In practice, the (1 1) confidence is selected to be the
same value as the confidence, typically 0.95. With = 0.95 and (1 1) = 0.95, then
(1 2) = 0.975. The proportion P is usually chosen to be 0.999.
The multiplier C is determined from
zp is the standard normal variable of the proportion P for a normal distribution,
is from the Chi-square distribution for (n 2) degrees of freedom at the (1 2)
confidence,
and n is number of calculated critical experiments used in the validation.
NOTE: The author of footnote 111 is not consistent with the subscript notation used during the
development of the technique. The technique is based upon the "upper tail" of the
Chi-square distribution, such that
Most Chi-square tables typically denote the "lower tail," such that
then, the upper tail of the distribution is
Thus, to obtain the value of (author's notation), enter the table to find
(typical notation), as the author does in the example problem for footnote
111.
C* is evaluated over the range of the independent variable, a < x < b, where a and b are,
respectively, the lower and upper limits of the areas of applicability. C* is determined by calculating
values for g, h, , and A, where
and
A = g/h .
The values of , A, and (n 2) are used to determine a value D from Table F.7.1, at the (1 - 1)
confidence. Table F.7.1 covers the range of 0.5 A 1.5; then
C* = Dg .
If A is outside the range of 0.5 to 1.5, then use 1/A, , and (n 2) to determine D; then
C* = Dh .
In Table F.7.1, the values for D have been derived by evaluating the double integral given in footnote
110 and are essentially identical to the D values given in Table 3 of footnote 110. Table F.7.1
covers the same range for (n-2), , A as the footnote, and also includes a 0.99 confidence (not
included in the footnoted table). Table F.7.1 is provided for users who may not have access to
footnote 110 and who may wish to impose a more restrictive confidence criteria.
Once these values have been determined, the linear regression, k(x), and the LTB should be
graphically depicted for future reference. The LTB is the USL for the maximum allowable keff, ks, as a
function of the independent variable, as shown in Figure F.7.1. For application, a calculated keff plus
two standard deviations shall lie below the USL line, keff, ks, + 2 < USL.
This statistical method for code validation allows the USL to be established such that there is a high
degree of confidence that a calculated result that satisfies the acceptance criteria is indeed
subcritical. Although a margin of subcriticality is not determined by the technique, a margin can be
defined as the difference between the (1 1) confidence on the linear regression for a single future
calculation and the USL. The (1 1) confidence on a single future calculation is determined by
is from the student-t distribution at the (1 1) confidence and (n 2) degrees of
freedom, and
sp is the pooled standard deviation previously determined.
Since w(x) is a curvilinear function, and it is desirable to have a constant width margin, the
expression is evaluated at xi = a and xi = b (the lowest and highest values, respectively, of the
independent variable). The larger of the two is the constant W, to be deducted from the linear
regression, k(x), to provide a uniform width confidence band for a single future calculation. The
margin of subcriticality is the difference between the uniform width confidence band for a single
future calculation and the USL, or
margin of subcriticality = [k(x) W] [k(x) Csp]
= Csp W .
Figure F.7.1 graphically depicts typical results of the single-sided, uniform-width, closed-interval, LTB
technique for code validation. Since the calculational bias has been accounted for in the linear
regression, it is not uniquely determined. Numerically, the average calculational bias at any point
within the areas of applicability is the difference between the linear regression k(x) and unity. Table F.7.1. Calculated D Values
(1 - 1) = 0.99
A
(n-2)
| |
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
4
.0
9.44
7.89
6.91
6.34
5.78
5.50
5.22
5.08
4.94
4.87
4.80
4
.1
9.44
7.89
6.91
6.20
5.78
5.50
5.22
5.08
4.94
4.80
4.80
4
.3
9.44
7.89
6.91
6.20
5.78
5.50
5.22
5.08
4.94
4.80
4.80
4
.5
9.16
7.75
6.91
6.20
5.64
5.36
5.08
4.94
4.87
4.80
4.80
4
.7
9.16
7.75
6.77
6.06
5.50
5.22
5.08
4.94
4.80
4.80
4.66
4
.9
9.16
7.75
6.63
5.92
5.36
5.08
4.80
4.73
4.66
4.66
4.66
6
.0
7.47
6.20
5.50
4.94
4.52
4.23
4.09
3.95
3.88
3.81
3.81
6
.1
7.47
6.20
5.50
4.94
4.52
4.23
4.09
3.95
3.88
3.81
3.81
6
.3
7.47
6.20
5.50
4.94
4.52
4.23
4.09
3.95
3.88
3.81
3.81
6
.5
7.47
6.20
5.43
4.87
4.52
4.23
4.02
3.95
3.81
3.81
3.74
6
.7
7.47
6.20
5.36
4.80
4.38
4.16
3.95
3.88
3.81
3.74
3.74
6
.9
7.47
6.20
5.36
4.66
4.23
3.95
3.81
3.81
3.74
3.74
3.74
8
.0
6.77
5.64
4.94
4.38
4.02
3.81
3.67
3.53
3.46
3.46
3.39
8
.1
6.77
5.64
4.94
4.38
4.02
3.81
3.67
3.53
3.46
3.46
3.39
8
.3
6.77
5.64
4.87
4.38
4.02
3.81
3.67
3.53
3.46
3.46
3.39
8
.5
6.77
5.64
4.87
4.38
4.02
3.74
3.60
3.53
3.46
3.39
3.39
8
.7
6.77
5.64
4.80
4.30
3.95
3.74
3.53
3.46
3.43
3.39
3.39
8
.9
6.77
5.64
4.80
4.23
3.81
3.60
3.46
3.39
3.39
3.39
3.39
10
.0
6.34
5.36
4.59
4.09
3.81
3.57
3.43
3.32
3.25
3.25
3.18
10
.1
6.34
5.36
4.59
4.09
3.81
3.57
3.43
3.32
3.25
3.25
3.18
10
.3
6.34
5.29
4.59
4.09
3.78
3.53
3.39
3.32
3.25
3.25
3.18
10
.5
6.34
5.29
4.59
4.09
3.74
3.53
3.39
3.32
3.25
3.21
3.18
10
.7
6.34
5.29
4.59
4.02
3.71
3.46
3.32
3.25
3.25
3.18
3.18
10
.9
6.34
5.29
4.52
4.02
3.60
3.39
3.25
3.18
3.18
3.18
3.18
12
.0
6.06
5.08
4.45
3.95
3.64
3.43
3.29
3.18
3.14
3.11
3.11
12
.1
6.06
5.08
4.45
3.95
3.64
3.43
3.29
3.18
3.14
3.11
3.11
12
.3
6.06
5.08
4.45
3.95
3.60
3.39
3.25
3.18
3.11
3.11
3.07
12
.5
6.06
5.08
4.38
3.95
3.60
3.39
3.25
3.18
3.11
3.11
3.07
12
.7
6.06
5.08
4.38
3.88
3.57
3.32
3.21
3.14
3.11
3.07
3.07
12
.9
6.06
5.08
4.38
3.81
3.46
3.25
3.14
3.11
3.04
3.04
3.04
14
.0
5.92
4.94
4.30
3.81
3.53
3.32
3.18
3.11
3.04
3.04
3.00
14
.1
5.92
4.94
4.30
3.81
3.53
3.32
3.18
3.11
3.04
3.04
3.00
14
.3
5.92
4.94
4.30
3.81
3.53
3.32
3.18
3.11
3.04
3.04
3.00
14
.5
5.92
4.94
4.30
3.81
3.50
3.29
3.18
3.07
3.04
3.00
3.00
14
.7
5.92
4.94
4.30
3.81
3.46
3.25
3.11
3.04
3.04
3.00
2.97
14
.9
5.92
4.94
4.23
3.74
3.39
3.18
3.04
3.00
2.97
2.97
2.97
Table F.7.1 (cont.)
(1 - 1) = 0.99 (cont.)
A
(n-2)
| |
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
16
.0
5.85
4.87
4.23
3.74
3.46
3.25
3.11
3.04
2.97
2.97
2.93
16
.1
5.85
4.87
4.23
3.74
3.46
3.25
3.11
3.04
2.97
2.97
2.93
16
.3
5.85
4.87
4.23
3.74
3.46
3.25
3.11
3.04
2.97
2.97
2.93
16
.5
5.85
4.87
4.23
3.74
3.43
3.21
3.11
3.04
2.97
2.97
2.93
16
.7
5.85
4.87
4.16
3.74
3.39
3.18
3.04
3.00
2.97
2.93
2.93
16
.9
5.85
4.87
4.16
3.67
3.32
3.11
3.00
2.97
2.93
2.93
2.90
20
.0
5.71
4.73
4.09
3.67
3.36
3.14
3.04
2.93
2.90
2.86
2.86
20
.1
5.71
4.73
4.09
3.67
3.36
3.14
3.04
2.93
2.90
2.86
2.86
20
.3
5.71
4.73
4.09
3.64
3.32
3.14
3.00
2.93
2.90
2.86
2.86
20
.5
5.71
4.73
4.09
3.64
3.32
3.11
3.00
2.93
2.90
2.86
2.86
20
.7
5.71
4.73
4.09
3.60
3.29
3.11
2.97
2.90
2.86
2.86
2.86
20
.9
5.71
4.73
4.09
3.57
3.25
3.04
2.90
2.86
2.86
2.83
2.83
24
.0
5.57
4.66
4.02
3.60
3.29
3.07
2.97
2.90
2.83
2.83
2.83
24
.1
5.57
4.66
4.02
3.60
3.29
3.07
2.97
2.90
2.83
2.83
2.83
24
.3
5.57
4.66
4.02
3.60
3.29
3.07
2.97
2.90
2.83
2.83
2.83
24
.5
5.57
4.66
4.02
3.57
3.25
3.07
2.93
2.86
2.83
2.83
2.79
24
.7
5.57
4.66
4.02
3.53
3.25
3.04
2.93
2.86
2.83
2.83
2.79
24
.9
5.57
4.66
4.02
3.53
3.18
2.97
2.86
2.83
2.79
2.79
2.79
30
.0
5.50
4.59
3.95
3.53
3.21
3.04
2.90
2.83
2.79
2.76
2.76
30
.1
5.50
4.59
3.95
3.53
3.21
3.04
2.90
2.83
2.79
2.76
2.76
30
.3
5.50
4.59
3.95
3.53
3.21
3.04
2.90
2.83
2.79
2.76
2.76
30
.5
5.50
4.59
3.95
3.50
3.21
3.00
2.90
2.83
2.79
2.76
2.76
30
.7
5.50
4.59
3.95
3.46
3.18
2.97
2.86
2.79
2.76
2.76
2.76
30
.9
5.50
4.59
3.95
3.46
3.11
2.90
2.83
2.76
2.76
2.76
2.76
40
.0
5.43
4.52
3.88
3.46
3.14
2.97
2.86
2.79
2.76
2.72
2.72
40
.1
5.43
4.52
3.88
3.46
3.14
2.97
2.86
2.79
2.76
2.72
2.72
40
.3
5.43
4.52
3.88
3.46
3.14
2.97
2.86
2.79
2.76
2.72
2.72
40
.5
5.43
4.52
3.88
3.46
3.14
2.97
2.83
2.76
2.72
2.72
2.72
40
.7
5.43
4.52
3.88
3.43
3.11
2.93
2.83
2.76
2.72
2.72
2.72
40
.9
5.43
4.52
3.88
3.39
3.04
2.86
2.76
2.72
2.72
2.69
2.69
50
.0
5.36
4.45
3.85
3.39
3.11
2.93
2.83
2.76
2.72
2.69
2.69
50
.1
5.36
4.45
3.85
3.39
3.11
2.93
2.83
2.76
2.72
2.69
2.69
50
.3
5.36
4.45
3.85
3.39
3.11
2.93
2.83
2.76
2.72
2.69
2.69
50
.5
5.36
4.45
3.85
3.39
3.11
2.93
2.79
2.76
2.72
2.69
2.69
50
.7
5.36
4.45
3.81
3.39
3.07
2.90
2.79
2.72
2.69
2.69
2.69
50
.9
5.36
4.45
3.81
3.36
3.04
2.83
2.72
2.69
2.69
2.69
2.69
Table F.7.1 (cont.)
(1 - 1) = 0.95
A
(n-2)
| |
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
4
.0
5.68
4.81
4.25
3.85
3.58
3.39
3.23
3.13
3.04
2.99
2.93
4
.1
5.68
4.80
4.23
3.85
3.58
3.37
3.23
3.13
3.04
2.99
2.93
4
.3
5.64
4.80
4.23
3.83
3.57
3.36
3.21
3.11
3.04
2.97
2.91
4
.5
5.64
4.76
4.18
3.78
3.51
3.32
3.16
3.07
2.99
2.93
2.90
4
.7
5.61
4.71
4.11
3.71
3.43
3.23
3.09
3.00
2.93
2.88
2.86
4
.9
5.57
4.66
4.01
3.57
3.25
3.07
2.95
2.88
2.83
2.81
2.79
6
.0
4.94
4.18
3.67
3.34
3.09
2.92
2.79
2.71
2.63
2.58
2.55
6
.1
4.94
4.18
3.67
3.34
3.09
2.92
2.79
2.71
2.63
2.58
2.55
6
.3
4.94
4.16
3.67
3.32
3.07
2.90
2.78
2.69
2.62
2.57
2.55
6
.5
4.94
4.15
3.64
3.29
3.04
2.86
2.74
2.65
2.60
2.55
2.53
6
.7
4.90
4.13
3.58
3.21
2.97
2.79
2.69
2.60
2.55
2.51
2.49
6
.9
4.90
4.09
3.51
3.13
2.85
2.67
2.57
2.51
2.48
2.46
2.46
8
.0
4.64
3.92
3.44
3.11
2.88
2.72
2.60
2.52
2.46
2.42
2.39
8
.1
4.64
3.92
3.44
3.11
2.88
2.72
2.60
2.52
2.46
2.42
2.39
8
.3
4.64
3.92
3.43
3.09
2.86
2.71
2.59
2.51
2.45
2.41
2.38
8
.5
4.62
3.90
3.41
3.07
2.83
2.67
2.56
2.48
2.42
2.39
2.36
8
.7
4.62
3.87
3.36
3.02
2.78
2.62
2.51
2.44
2.39
2.36
2.34
8
.9
4.62
3.85
3.30
2.93
2.67
2.51
2.42
2.36
2.34
2.32
2.31
10
.0
4.48
3.78
3.30
2.99
2.77
2.61
2.49
2.42
2.36
2.32
2.29
10
.1
4.48
3.78
3.30
2.99
2.76
2.61
2.49
2.42
2.36
2.32
2.29
10
.3
4.48
3.76
3.29
2.97
2.75
2.60
2.49
2.41
2.35
2.32
2.28
10
.5
4.46
3.76
3.27
2.95
2.72
2.56
2.46
2.39
2.34
2.30
2.27
10
.7
4.46
3.74
3.23
2.90
2.67
2.52
2.42
2.34
2.30
2.27
2.26
10
.9
4.45
3.71
3.20
2.83
2.58
2.42
2.33
2.27
2.25
2.24
2.23
12
.0
4.38
3.69
3.21
2.91
2.69
2.54
2.43
2.35
2.30
2.27
2.24
12
.1
4.38
3.68
3.21
2.91
2.69
2.54
2.43
2.35
2.30
2.27
2.24
12
.3
4.38
3.67
3.21
2.90
2.68
2.53
2.42
2.34
2.29
2.26
2.23
12
.5
4.38
3.67
3.20
2.87
2.65
2.50
2.40
2.33
2.27
2.24
2.22
12
.7
4.36
3.65
3.16
2.83
2.61
2.46
2.35
2.29
2.25
2.22
2.20
12
.9
4.36
3.64
3.13
2.76
2.51
2.36
2.27
2.22
2.20
2.19
2.19
14
.0
4.30
3.62
3.16
2.85
2.64
2.49
2.39
2.31
2.26
2.22
2.20
14
.1
4.30
3.62
3.16
2.85
2.64
2.49
2.39
2.31
2.26
2.22
2.20
14
.3
4.30
3.62
3.15
2.85
2.63
2.48
2.38
2.30
2.25
2.21
2.20
14
.5
4.30
3.60
3.14
2.82
2.61
2.46
2.35
2.28
2.23
2.20
2.18
14
.7
4.29
3.58
3.11
2.78
2.56
2.42
2.32
2.25
2.20
2.18
2.17
14
.9
4.29
3.58
3.07
2.71
2.48
2.32
2.23
2.19
2.16
2.15
2.15
Table F.7.1 (cont.)
(1 - 1) = 0.95 (cont.)
A
(n-2)
| |
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
16
.0
4.25
3.57
3.13
2.81
2.61
2.46
2.35
2.28
2.23
2.20
2.17
16
.1
4.25
3.57
3.12
2.81
2.60
2.46
2.35
2.28
2.23
2.20
2.17
16
.3
4.25
3.57
3.11
2.80
2.59
2.45
2.34
2.27
2.22
2.19
2.16
16
.5
4.25
3.56
3.09
2.78
2.57
2.42
2.32
2.26
2.20
2.18
2.15
16
.7
4.23
3.55
3.07
2.75
2.53
2.38
2.28
2.22
2.18
2.15
2.14
16
.9
4.23
3.53
3.04
2.69
2.44
2.29
2.20
2.16
2.13
2.13
2.13
20
.0
4.18
3.51
3.07
2.76
2.56
2.42
2.31
2.24
2.19
2.15
2.13
20
.1
4.18
3.51
3.06
2.76
2.56
2.41
2.31
2.24
2.19
2.15
2.13
20
.3
4.18
3.50
3.06
2.75
2.55
2.40
2.30
2.23
2.18
2.14
2.13
20
.5
4.18
3.50
3.04
2.73
2.52
2.38
2.28
2.21
2.17
2.13
2.12
20
.7
4.16
3.49
3.02
2.70
2.49
2.34
2.24
2.18
2.14
2.12
2.11
20
.9
4.16
3.48
2.99
2.63
2.41
2.26
2.17
2.13
2.10
2.09
2.09
24
.0
4.13
3.47
3.02
2.72
2.52
2.38
2.28
2.21
2.16
2.13
2.11
24
.1
4.13
3.47
3.02
2.72
2.52
2.38
2.28
2.20
2.16
2.13
2.10
24
.3
4.13
3.46
3.02
2.72
2.51
2.37
2.27
2.20
2.15
2.12
2.10
24
.5
4.13
3.46
3.00
2.71
2.49
2.35
2.25
2.19
2.14
2.11
2.09
24
.7
4.13
3.44
2.99
2.67
2.45
2.31
2.21
2.15
2.12
2.09
2.08
24
.9
4.13
3.44
2.95
2.61
2.37
2.23
2.14
2.10
2.08
2.07
2.06
30
.0
4.09
3.43
2.99
2.70
2.49
2.35
2.25
2.18
2.13
2.10
2.08
30
.1
4.09
3.43
2.99
2.70
2.49
2.35
2.25
2.18
2.13
2.10
2.08
30
.3
4.09
3.43
2.99
2.69
2.49
2.34
2.24
2.18
2.13
2.10
2.07
30
.5
4.09
3.43
2.97
2.67
2.46
2.32
2.22
2.16
2.12
2.09
2.06
30
.7
4.09
3.41
2.95
2.63
2.42
2.28
2.19
2.13
2.09
2.07
2.05
30
.9
4.08
3.41
2.92
2.58
2.34
2.20
2.12
2.07
2.05
2.05
2.05
40
.0
4.04
3.39
2.95
2.66
2.46
2.32
2.22
2.15
2.11
2.07
2.05
40
.1
4.04
3.39
2.95
2.66
2.46
2.32
2.22
2.15
2.11
2.07
2.05
40
.3
4.04
3.39
2.95
2.65
2.45
2.31
2.21
2.15
2.10
2.07
2.05
40
.5
4.04
3.38
2.93
2.63
2.43
2.29
2.20
2.13
2.09
2.06
2.05
40
.7
4.04
3.37
2.92
2.61
2.40
2.26
2.16
2.11
2.07
2.05
2.04
40
.9
4.04
3.37
2.89
2.55
2.32
2.18
2.10
2.05
2.04
2.03
2.02
50
.0
4.02
3.37
2.93
2.64
2.44
2.30
2.20
2.14
2.09
2.06
2.04
50
.1
4.02
3.37
2.93
2.64
2.44
2.30
2.20
2.14
2.09
2.06
2.04
50
.3
4.02
3.36
2.92
2.63
2.43
2.29
2.20
2.13
2.09
2.05
2.04
50
.5
4.02
3.36
2.92
2.62
2.42
2.27
2.18
2.12
2.07
2.05
2.03
50
.7
4.02
3.36
2.90
2.59
2.38
2.24
2.15
2.09
2.05
2.03
2.02
50
.9
4.02
3.35
2.87
2.54
2.31
2.16
2.08
2.04
2.02
2.01
2.01
Table F.7.1 (cont.)
(1 - 1) = 0.90
A
(n-2)
| |
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
4
.0
4.38
3.74
3.32
3.03
2.82
2.66
2.55
2.46
2.39
2.34
2.30
4
.1
4.38
3.74
3.32
3.02
2.81
2.66
2.55
2.46
2.39
2.34
2.30
4
.3
4.38
3.72
3.30
3.00
2.79
2.64
2.53
2.44
2.37
2.32
2.28
4
.5
4.34
3.69
3.26
2.96
2.75
2.60
2.49
2.41
2.34
2.29
2.26
4
.7
4.30
3.64
3.20
2.89
2.67
2.52
2.42
2.34
2.28
2.25
2.21
4
.9
4.27
3.58
3.09
2.76
2.53
2.38
2.28
2.23
2.19
2.17
2.16
6
.0
3.95
3.37
2.99
2.71
2.52
2.38
2.28
2.20
2.14
2.10
2.06
6
.1
3.95
3.37
2.98
2.71
2.52
2.38
2.28
2.20
2.14
2.10
2.06
6
.3
3.94
3.36
2.97
2.70
2.50
2.37
2.27
2.19
2.13
2.09
2.05
6
.5
3.94
3.33
2.93
2.66
2.47
2.33
2.23
2.16
2.10
2.06
2.03
6
.7
3.91
3.29
2.88
2.60
2.41
2.27
2.18
2.11
2.06
2.02
2.00
6
.9
3.88
3.25
2.80
2.49
2.29
2.15
2.07
2.02
1.98
1.97
1.96
8
.0
3.77
3.21
2.83
2.57
2.40
2.26
2.16
2.09
2.04
1.99
1.96
8
.1
3.77
3.21
2.83
2.57
2.39
2.26
2.16
2.09
2.03
1.99
1.96
8
.3
3.76
3.19
2.82
2.56
2.38
2.25
2.15
2.08
2.02
1.98
1.95
8
.5
3.75
3.17
2.79
2.53
2.34
2.21
2.12
2.05
2.00
1.96
1.93
8
.7
3.73
3.14
2.74
2.48
2.29
2.16
2.07
2.00
1.96
1.93
1.91
8
.9
3.72
3.11
2.68
2.38
2.18
2.05
1.97
1.92
1.90
1.88
1.87
10
.0
3.66
3.11
2.75
2.50
2.32
2.20
2.10
2.03
1.98
1.93
1.90
10
.1
3.66
3.11
2.75
2.49
2.32
2.19
2.10
2.02
1.97
1.93
1.90
10
.3
3.65
3.10
2.73
2.49
2.31
2.18
2.09
2.02
1.96
1.92
1.89
10
.5
3.65
3.08
2.71
2.45
2.27
2.15
2.05
1.99
1.94
1.91
1.88
10
.7
3.64
3.06
2.67
2.41
2.22
2.10
2.01
1.95
1.91
1.87
1.85
10
.9
3.63
3.02
2.61
2.32
2.13
2.00
1.92
1.87
1.84
1.83
1.82
12
.0
3.60
3.05
2.69
2.45
2.27
2.15
2.05
1.98
1.93
1.90
1.87
12
.1
3.60
3.05
2.69
2.45
2.27
2.15
2.05
1.98
1.93
1.89
1.86
12
.3
3.59
3.04
2.68
2.43
2.26
2.13
2.04
1.98
1.92
1.88
1.86
12
.5
3.58
3.02
2.66
2.41
2.23
2.11
2.02
1.95
1.90
1.87
1.84
12
.7
3.58
3.00
2.62
2.36
2.18
2.06
1.97
1.91
1.87
1.84
1.82
12
.9
3.57
2.98
2.56
2.27
2.09
1.96
1.88
1.83
1.81
1.80
1.79
14
.0
3.55
3.01
2.65
2.42
2.24
2.12
2.03
1.96
1.91
1.87
1.84
14
.1
3.55
3.01
2.65
2.41
2.24
2.12
2.02
1.96
1.91
1.87
1.84
14
.3
3.55
3.00
2.64
2.40
2.23
2.10
2.02
1.94
1.90
1.86
1.83
14
.5
3.54
2.99
2.62
2.37
2.20
2.08
1.99
1.92
1.87
1.84
1.82
14
.7
3.53
2.96
2.58
2.33
2.15
2.03
1.94
1.88
1.84
1.81
1.80
14
.9
3.52
2.94
2.53
2.25
2.06
1.94
1.86
1.81
1.79
1.77
1.77
Table F.7.1 (cont.)
(1 - 1) = 0.90 (cont.)
A
(n-2)
| |
0.5
0.6
0.7
0.8
0.9
1.0
1.1
1.2
1.3
1.4
1.5
16
.0
3.51
2.98
2.63
2.39
2.22
2.09
2.01
1.94
1.89
1.85
1.82
16
.1
3.51
2.98
2.63
2.39
2.22
2.09
2.00
1.94
1.88
1.85
1.82
16
.3
3.51
2.97
2.62
2.38
2.20
2.08
1.99
1.93
1.87
1.84
1.81
16
.5
3.50
2.96
2.59
2.35
2.18
2.05
1.97
1.91
1.86
1.82
1.80
16
.7
3.50
2.93
2.56
2.31
2.13
2.01
1.92
1.87
1.83
1.80
1.78
16
.9
3.50
2.92
2.51
2.23
2.04
1.92
1.84
1.80
1.77
1.76
1.75
20
.0
3.47
2.94
2.59
2.35
2.19
2.06
1.98
1.91
1.86
1.82
1.80
20
.1
3.47
2.93
2.59
2.35
2.19
2.06
1.98
1.91
1.86
1.82
1.79
20
.3
3.47
2.93
2.58
2.34
2.17
2.05
1.96
1.90
1.85
1.81
1.79
20
.5
3.46
2.92
2.56
2.31
2.15
2.03
1.94
1.88
1.83
1.80
1.77
20
.7
3.45
2.90
2.52
2.27
2.10
1.98
1.90
1.84
1.80
1.77
1.76
20
.9
3.45
2.88
2.48
2.20
2.01
1.89
1.82
1.77
1.75
1.73
1.71
24
.0
3.44
2.91
2.56
2.33
2.16
2.05
1.96
1.89
1.84
1.80
1.78
24
.1
3.44
2.91
2.56
2.33
2.16
2.05
1.95
1.89
1.84
1.80
1.78
24
.3
3.43
2.90
2.56
2.32
2.15
2.03
1.94
1.88
1.83
1.80
1.77
24
.5
3.43
2.89
2.53
2.29
2.13
2.01
1.92
1.86
1.81
1.78
1.76
24
.7
3.43
2.87
2.50
2.25
2.08
1.96
1.88
1.82
1.78
1.76
1.74
24
.9
3.43
2.85
2.46
2.18
1.99
1.87
1.80
1.76
1.73
1.72
1.72
30
.0
3.41
2.88
2.54
2.31
2.14
2.02
1.94
1.87
1.82
1.79
1.76
30
.1
3.41
2.88
2.54
2.31
2.14
2.02
1.94
1.87
1.82
1.79
1.76
30
.3
3.41
2.88
2.53
2.30
2.13
2.01
1.93
1.86
1.81
1.78
1.75
30
.5
3.40
2.86
2.51
2.27
2.11
1.99
1.91
1.84
1.80
1.76
1.74
30
.7
3.40
2.85
2.48
2.23
2.06
1.94
1.87
1.81
1.77
1.74
1.73
30
.9
3.39
2.83
2.43
2.16
1.98
1.86
1.79
1.74
1.72
1.71
1.70
40
.0
3.38
2.85
2.52
2.28
2.13
2.01
1.92
1.85
1.80
1.77
1.74
40
.1
3.38
2.85
2.52
2.28
2.12
2.01
1.92
1.85
1.80
1.77
1.74
40
.3
3.38
2.85
2.51
2.27
2.11
1.99
1.91
1.84
1.80
1.76
1.74
40
.5
3.37
2.84
2.49
2.25
2.09
1.97
1.89
1.83
1.78
1.75
1.73
40
.7
3.37
2.82
2.46
2.21
2.05
1.93
1.85
1.79
1.75
1.73
1.71
40
.9
3.36
2.81
2.42
2.14
1.96
1.84
1.77
1.73
1.70
1.69
1.69
50
.0
3.36
2.85
2.50
2.27
2.11
1.99
1.91
1.84
1.80
1.76
1.73
50
.1
3.36
2.84
2.50
2.27
2.11
1.99
1.91
1.84
1.80
1.76
1.73
50
.3
3.36
2.84
2.49
2.26
2.10
1.98
1.90
1.83
1.79
1.75
1.73
50
.5
3.36
2.83
2.48
2.24
2.08
1.96
1.87
1.81
1.77
1.74
1.72
50
.7
3.36
2.81
2.45
2.20
2.03
1.92
1.84
1.78
1.74
1.72
1.70
50
.9
3.36
2.79
2.41
2.13
1.95
1.83
1.76
1.72
1.69
1.69
1.68
Figure F.7.1. Typical results for the single-sided, uniform-width, closed-interval, LTB technique. As an example of this technique, assume that 29 critical experiments have been modeled and
calculated. The calculated keff, standard deviation, and average energy group causing fission (the
independent variable, x) are shown in Table F.7.2.
Table F.7.2. Input Data for Example Problem
keff
AEG
0.99647
0.99776
1.00764
0.99587
0.99744
1.00337
0.99609
1.00108
0.99737
0.98408
0.98871
0.99527
0.98804
1.01363
1.01660
1.00874
1.01190
1.00980
1.00565
1.01929
1.00860
0.99487
0.99257
1.00132
0.99154
1.00028
0.99565
0.98574
0.98733
0.00337
0.00326
0.00311
0.00365
0.00327
0.00335
0.00395
0.00378
0.00325
0.00342
0.00361
0.00292
0.00273
0.00401
0.00445
0.00485
0.00479
0.00498
0.00397
0.00407
0.00411
0.00462
0.00382
0.00450
0.00420
0.00374
0.00413
0.00415
0.00416
14.82
14.81
14.83
14.44
14.28
14.84
14.73
15.08
15.20
15.31
15.26
15.50
15.49
14.36
14.36
14.36
14.38
14.35
14.10
14.12
14.10
15.04
14.90
14.90
14.90
15.43
15.44
15.43
15.43
Table F.7.3 summarizes the various terms calculated to establish the USL and margin of
subcriticality, with = 0.999 and = 0.95. From these results, the USL is defined by the straight
line USL = 1.1900 0.0153 x, and is statistically valid only between the range of AEG from 14.10
to 15.50. Any calculated keff + 2 that is below the USL is adequately subcritical, with a margin of
subcriticality of at least 0.02.
Table F.7.3. Calculated Terms for Example Problem
n
linear regression, k(x)
minimum value of x, a
maximum value of x, b
average x (AEG),
average keff,
variance of fit,
within variance,
pooled variance,
pooled standard deviation, sp
Zp @ P = 0.999
x2 @ (n 2), (1 2)
g
h
A
D
C*
C
Csp
LTB = USL
student-t @ (n 2) (1 1)
W (max. at x=a and x=b)
minimum margin of
subcriticality, Csp-w
= 29
= 1.2266 0.015295 x
= 14.10
= 15.50
= 14.8341
= 0.99975
= 3.8260-05a
= 1.5304-05
= 5.3564-05
= 7.3187-03
= 3.090
= 14.57
= 0.3497
= 0.3266
= 0.3951
= 1.0705
= 2.274 (interpolated from Table C-1)
= 0.79525
= 4.9973
= 0.0366
= 1.1900 0.015295 x
= 1.703
= 0.0132
= 0.0234
aRead as 3.8260 x 10-5.
Figure F.7.2 provides a plot of the resultant single-sided, uniform-width, closed-interval, lower
tolerance band technique developed from the example.
There may be valid reasons to reduce the USL. There are many factors that may both change the
AEG and affect other parameters as well in a multiplying manner. For example, suppose very low
temperatures cannot be ruled out for the application in question. Low temperatures will increase the
AEG and can also increase density. Therefore cold increases reactivity by increasing density, may
change the AEG to be outside the range of applicability of critical experiments, and decreases the
margin of safety.
Figure F.7.2. Example results for the single-sided, uniform-width, closed-interval, LTB technique. APPENDIX G. BIBLIOGRAPHY OF JOURNAL ARTICLES AND MEETING AND
CONFERENCE PROCEEDINGS
In the bibliography that follows, JA designates a journal article, PA designates a proceedings article,
and P designates complete proceedings for which individual article listings have been omitted in this
bibliography. The JA, PA, and P groupings follow in succession. Items are listed chronologically
within each JA, PA, and P grouping. Except for G-JA001, G-JA002, and G-JA170, all journal
articles can be found in either Nuclear Science and Engineering or Nuclear Technology (formerly
Nuclear Applications or Nuclear Applications Technology).
G-JA001. Gerald Goertzel, "Minimum Critical Mass and Flat Flux," J. Nucl. Energy, 2, 193 (1956).
G-JA002. Dixon Callihan, "Homogeneous Critical Assemblies," Progress in Nuclear Energy, Series I,
Physics and Mathematics (ed. R.A. Charpie, J. Horowitz, D.J. Hughes, and D.J. Littler), p. 227,
McGraw-Hill, New York (1956).
G-JA003. Roger H. White, "Topsy, A Remotely Controlled Critical Assembly Machine," Nucl. Sci.
Eng., 1, 53 (Mar. 1956).
G-JA004. R.E. Peterson and G.A. Newby, "An Unreflected U-235 Critical Assembly," Nucl. Sci.
Eng., 1, 112 (May 1956).
G-JA005. Richard N. Olcott, "Homogeneous Heavy Water Moderated Critical Assemblies. Part 1.
Experimental," Nucl. Sci. Eng., 1, 327 (Aug. 1956).
G-JA006. John D. Orndoff, "Prompt Neutron Periods of Metal Critical Assemblies," Nucl. Sci. Eng.,
2, 450 (Jul. 1957).
G-JA007. J.K. Fox, L.W. Gilley, and J.H. Marable, "Critical Parameters of a Proton-Moderated and
Proton-Reflected Slab of U235," Nucl. Sci. Eng., 3, 694 (Jun. 1958).
G-JA008. A. Sauer, "Optimum Control and Flat Flux," Nucl. Sci. Eng., 5, 71 (Jan. 1959).
G-JA009. J. Devooght, "Restricted Minimum Critical Mass," Nucl. Sci. Eng., 5, 190 (Mar. 1959).
G-JA010. W. Goad and R. Johnston, "A Monte Carlo Method for Criticality Problems," Nucl. Sci.
Eng., 5, 371 (Jun. 1959).
G-JA011. Norman Ketzlach, "Nuclear Safety of Iron-Encased Fuel Elements," Nucl. Sci. Eng., 6, 187
(Sep. 1959).
G-JA012. J. Ernest Wilkins, Jr., "Minimum Total Mass," Nucl. Sci. Eng., 6, 229 (Sep. 1959).
G-JA013. G.A. Linenberger, J.D. Orndoff, and H.C. Paxton, "Enriched-Uranium Hydride Critical
Assemblies," Nucl. Sci. Eng., 7, 44 (Jan. 1960).
G-JA014. G.A. Jarvis, G.A. Linenberger, J.D. Orndoff, and H.C. Paxton, "Two Plutonium-Metal
Critical Assemblies," Nucl. Sci. Eng., 8, 525 (Dec. 1960).
G-JA015. L.B. Engle, G.E. Hansen, and H.C. Paxton, "Reactivity Contributions of Various Materials
in Topsy," Nucl. Sci. Eng., 8, 543 (Dec. 1960).
G-JA016. G.E. Hansen, H.C. Paxton, and D.P. Wood, "Critical Plutonium and Enriched-Uranium-
Metal Cylinders of Extreme Shape," Nucl. Sci. Eng., 8, 570 (Dec. 1960).
G-JA017. D.P. Wood, C.C. Byers, and L.C. Osborn, "Critical Masses of Cylinders of Plutonium
Diluted with Other Metals," Nucl. Sci. Eng., 8, 578 (Dec. 1960).
G-JA018. G.E. Hansen, D.P. Wood, and W.U. Geer, "Critical Masses of Enriched- Uranium Cylinders
with Multiple Reflectors of Medium-Z Elements," Nucl. Sci. Eng., 8, 588 (Dec. 1960).
G-JA019. Leona Stewart, "Leakage Neutron Spectrum from a Bare Pu239 Critical Assembly," Nucl.
Sci. Eng., 8, 595 (Dec. 1960).
G-JA020. E.A. Plassmann and D.P. Wood, "Critical Reflector Thicknesses of Spherical U233 and Pu239Systems," Nucl. Sci. Eng., 8, 615 (Dec. 1960).
G-JA021. C.B. Mills, "Minimum Critical Dimensions for Water Solutions," Nucl. Sci. Eng., 9, 377
(Mar. 1961).
G-JA022. Mathew M. Shapiro, "Minimum Critical Mass in Variable Density and Epithermal
Reactors," Nucl. Sci. Eng., 10, 159 (Jun. 1961).
G-JA023. John Mihalczo and C.B. Mills, "Criticality of Low Enrichment U235 in Hydrogen," Nucl. Sci.
Eng., 11, 95 (Sep. 1961).
G-JA024. J.R. Knight, "Calculation of the Neutron Leakage Spectra and Neutron Doses from Bare
Critical Systems," Nucl. Sci. Eng., 11, 239 (Nov. 1961).
G-JA025. C.B. Mills and G.I. Bell, "Criticality of Low Enrichment Uranium in Hydrogen," Nucl. Sci.
Eng., 12, 469 (Apr. 1962).
G-JA026. Hugh K. Clark, "Interaction of Fissionable Units," Nucl. Sci. Eng., 15, 20 (Jan. 1963).
G-JA027. W.B. Rogers, Jr. and F.E. Kinard, "Material Buckling and Critical Masses of Uranium Rods
Containing 3 wt %U235 in H2O," Nucl. Sci. Eng., 20, 266 (Nov. 1964).
G-JA028. C.R. Richey, R.C. Lloyd, and E.D. Clayton, "Criticality of Slightly Enriched Uranium in
Water-Moderated Lattices," Nucl. Sci. Eng., 21, 217 (Feb. 1965).
G-JA029. C.R. Richey, J.D. White, E.D. Clayton, and R.C. Lloyd, "Criticality of Homogeneous
Plutonium Oxide-Plastic Compacts at H:Pu=15," Nucl. Sci. Eng., 23, 150 (Oct. 1965).
G-JA030. H.K. Clark, "Effect of Distribution of Fissile Material on Critical Mass," Nucl. Sci. Eng., 24,
133 (Feb. 1966).
G-JA031. R.C. Lloyd, C.R. Richey, E.D. Clayton, and D.R. Skeen, "Criticality Studies with Plutonium
Solutions," Nucl. Sci. Eng., 25, 165 (Jun. 1966).
G-JA032. Dale E. Hankins, "Effect of Reactivity Addition Rate and of Weak Neutron Source on the
Fission Yield of Uranium Solutions," Nucl. Sci. Eng., 26, 110 (Sep. 1966).
G-JA033. L.E. Hansen and E.D. Clayton, "Criticality of Plutonium Compounds in the
Undermoderated Range, H:Pu 20," Nucl. Appl., 3, 481 (Aug. 1967).
G-JA034. Carroll B. Mills, "Low Critical Mass," Nucl. Appl., 4, 17 (Jan. 1968).
G-JA035. C.R. Richey, "Theoretical Analyses of Homogeneous Plutonium Critical Experiments,"
Nucl. Sci. Eng., 31, 32 (Jan. 1968).
G-JA036. C.R. Richey, "Criticality of Heterogeneous Arrays Undergoing Dissolution," Nucl. Sci.
Eng., 31, 40 (Jan. 1968).
G-JA037. Thomas Gutman, "A Surface Density Evaluation of Critical Array Data," Nucl. Appl., 4,
121 (Feb. 1968).
G-JA038. R.C. Lloyd, E.D. Clayton, and J.H. Chalmers, "Criticality of Arrays of 233U Solution," Nucl.
Appl., 4, 136 (Mar. 1968).
G-JA039. J. Wallace Webster, "Minimum Thickness of a Water-Reflected Infinite Slab of an
Aqueous Solution of 235UO2F2 at Optimum Concentration," Nucl. Sci. Eng., 32, 133 (Apr. 1968).
G-JA040. S.R. Bierman and G.M. Hess, "Minimum Critical 235U Enrichment of Homogeneous,
Hydrogenous Uranyl Nitrate Systems," Nucl. Sci. Eng., 32, 135 (Apr. 1968).
G-JA041. C.G. Chezem and E.J. Lozito, "Investigation of the Criticality of Low-Enrichment Uranium
Cylinders," Nucl. Sci. Eng., 33, 139 (Jul. 1968).
G-JA042. R.C. Kispert, et al., "Crystallization Characteristics of Acidic Uranyl Nitrate Solutions,"
Nucl. Appl., 5, 224 (Oct. 1968).
G-JA043. S.R. Bierman, L.E. Hansen, R.C. Lloyd, and E.D. Clayton, "Critical Experiments with
Homogeneous PuO2-Polystyrene at 5H/Pu," Nucl. Appl., 6, 23 (Jan. 1969).
G-JA044. Robert E. Rothe, "Critical Measurements on an Enriched Uranium Solution System," Nucl.
Sci. Eng., 35, 267 (Feb. 1969).
G-JA045. L.E. Hansen, E.D. Clayton, R.C. Lloyd, and S.R. Bierman, "Critical Parameters of
Plutonium Systems. Part I: Analysis of Experiments," Nucl. Appl., 6, 371 (Apr. 1969).
G-JA046. L.E. Hansen and E.D. Clayton, "Critical Parameters of Plutonium Systems. Part II:
Interpretation," Nucl. Appl., 6, 381 (Apr. 1969).
G-JA047. David R. Smith and William U. Geer, "Critical Mass of a Water-Reflected Plutonium
Sphere," Nucl. Appl. Technol., 7, 405 (Nov. 1969).
G-JA048. Robert E. Rothe, C.L. Schuske, and E.E. Hicks, "The Criticality of a Uranium-Solution Slab
Under Various Reflector Conditions," Nucl. Appl. Technol., 7, 505 (Dec. 1969).
G-JA049. William Bradley Lewis, "A Practical Approach to Nuclear Critical Safety," Nucl. Appl.
Technol., 7, 523 (Dec. 1969).
G-JA050. Donald C. Coonfield, Grover Tuck, Harold E. Clark, and Bruce B. Ernst, "Critical Mass
Irregularity of Steel-Moderated Enriched Uranium Metal Assemblies with Composite Steel-Oil
Reflectors," Nucl. Sci. Eng., 39, 320 (Mar. 1970).
G-JA051. Grover Tuck and Harold E. Clark, "Critical Parameters of a Uranium Solution Slab-Cylinder
System," Nucl. Sci. Eng., 40, 407 (Jun. 1970).
G-JA052. Richard A. Wolfe, "The Nuclear Criticality Safety Aspects of Plutonium-238," Nucl. Appl.
Technol., 9, 218 (Aug. 1970).
G-JA053. Harold E. Clark and Grover Tuck, "An Empirical Formula which Predicts the Critical
Parameters of a Planar Array of Uranium-Solution-Filled Cylinders," Nucl. Appl. Technol., 9, 814
(Dec. 1970).
G-JA054. Deanne Dickinson and C.L. Schuske, "An Empirical Model for Safe Pipe Intersections for
Containing Fissile Solution," Nucl. Technol., 10, 179 (Feb. 1971).
G-JA055. S.R. Bierman and E.D. Clayton, "Critical Experiments with Unmoderated Plutonium
Oxide," Nucl. Technol., 11, 185 (Jun. 1971).
G-JA056. D.C. Hunt and Robert E. Rothe, "Criticality Measurements on Uranium Metal Spheres
Immersed in Uranium Solution," Nucl. Sci. Eng., 46, 76 (Oct. 1971).
G-JA057. Thomas Gutman, "Anomalies in the Analysis of Air-Spaced Arrays," Nucl. Technol., 12,
162 (Oct. 1971).
G-JA058. William Bradley Lewis, "A Practical Approach to Nuclear Critical Safety II -- Critique of a
Model," Nucl. Technol., 12, 276 (Nov. 1971).
G-JA059. R.L. Currie, P.B. Parks, and J.L. Jarriel, "Static and Pulsed Reactivity Measurements on
Large Uranium-235 Fuel Forms in Water," Nucl. Technol., 12, 356 (Dec. 1971).
G-JA060. S.J. Altschuler and C.L. Schuske, "Models for the Safe Storage of Fissile Metal," Nucl.
Technol., 13, 131 (Feb. 1972).
G-JA061. Harold E. Clark and Grover Tuck, "An Empirical Formula that Predicts the Critical
Parameters of a Uranium Solution Slab-Cylinder System," Nucl. Technol., 13, 257 (Mar. 1972).
G-JA062. G.R. Handley, "Some Effects of Water Sprinklers on Array Criticality Safety Analyses,"
Nucl. Technol., 13, 71 (Apr. 1972).
G-JA063. R.C. Lloyd, E.D. Clayton, and L.E. Hansen, "Criticality of Plutonium Nitrate Solution
Containing Soluble Gadolinium," Nucl. Sci. Eng., 48, 300 (Jul. 1972).
G-JA064. S.R. Bierman and E.D. Clayton, "Critical Experiments with Homogeneous PuO2-Polystyrene at 50 H/Pu," Nucl. Technol., 15, 5 (Jul. 1972).
G-JA065. S.J. Raffety, "Homogeneous Critical Assemblies of 2 and 3% Uranium-235-Enriched
Uranium in Paraffin," Nucl. Sci. Eng., 48, 433 (Aug. 1972).
G-JA066. J.T. Mihalczo, "Criticality of Graphite- and Polyethylene-Reflected Uranium (93.2%)-Metal
Cylinders and Annuli," Nucl. Sci. Eng., 49, 489 (Dec. 1972).
G-JA067. C.L. Schuske and D.C. Hunt, "A Storage Container for Fissile Materials," Nucl. Technol.,
16, 562 (Dec. 1972).
G-JA068. S.R. Bierman, E.D. Clayton, and L.E. Hansen, "Critical Experiments with Homogeneous
Mixtures of Plutonium and Uranium Oxides Containing 8, 15, and 30 wt% Plutonium," Nucl. Sci.
Eng., 50, 115 (Feb. 1973).
G-JA069. R.C. Lloyd, S.R. Bierman and E.D. Clayton, "Criticality of Plutonium Nitrate Solutions
Containing Borated Raschig Rings," Nucl. Sci. Eng., 50, 127 (Feb. 1973).
G-JA070. S.J. Altschuler and C.L. Schuske, "A Model for the Safe Storage of Fissile Solutions,"
Nucl. Technol., 17, 110 (Feb. 1973).
G-JA071. S.J. Altschuler and C.L. Schuske, "A Study of Criticality Parameters Affecting the
Handling and Storage of Fissile Metal," Nucl. Technol., 18, 55 (Apr. 1973).
G-JA072. C.L. Brown, L.C. Davenport, and D.P. Oden, "Plutonium Fuel Technology Part III: Nuclear
Criticality Safety Considerations in LWR (Pu,U)O2 Fuel Fabrication," Nucl. Technol., 18, 109 (May
1973).
G-JA073. Grover Tuck, Harold E. Clark, and Donald L. Alvarez, "Enriched Uranium Metal-Solution
Systems Separated by Neutron Poisons," Nucl. Technol., 18, 216 (Jun. 1973).
G-JA074. R.C. Lloyd, E.D. Clayton, L.E. Hansen, and S.R. Bierman, "Criticality of Plutonium Nitrate
Solutions in Slab Geometry," Nucl. Technol., 18, 225 (Jun. 1973).
G-JA075. C.L. Schuske and S.J. Altschuler, "A Storage Vessel for Fissile Solutions," Nucl. Technol.,
18, 305 (Jun. 1973).
G-JA076. Ricardo Artigas, "Criticality Safety Evaluation of a Shipping Container for Moderated Low-
Enriched Uranium Compounds," Nucl. Technol., 19, 16 (Jul. 1973).
G-JA077. C.L. Schuske and S.J. Altschuler, "Models for the Safe Storage of Dry and Wet Fissile
Oxides," Nucl. Technol., 19, 84 (Aug. 1973).
G-JA078. R.C. Lloyd and E.D. Clayton, "The Criticality of High Burnup Plutonium," Nucl. Sci. Eng.,
52, 73 (Sep. 1973).
G-JA079. J.T. Thomas, "Critical Three-Dimensional Arrays of U(93.2)-Metal Cylinders," Nucl. Sci.
Eng., 52, 350 (Nov. 1973).
G-JA080. E.D. Clayton, "Fissionability and Criticality: From Protactinium to Californium and
Beyond," Nucl. Sci. Eng., 52, 417 (Nov. 1973).
G-JA081. C.L. Schuske and S.J. Altschuler, "Storage Capacity for Fissile Material as a Function of
Facility Size," Nucl. Technol., 20, 179 (Dec. 1973).
G-JA082. D.C. Hunt and Robert E. Rothe, "A Criticality Study of Fissile-Metal and Fissile-Solution
Combinations," Nucl. Sci. Eng., 53, 79 (Jan. 1974).
G-JA083. D.C. Hunt and C.L. Schuske, "Minimum Critical Masses of Arrays of Fissile Oxide and
Metal Elements in Water," Nucl. Technol., 22, 263 (May 1974).
G-JA084. Robert E. Rothe and D.C. Hunt, "The Effect of Position on the Criticality of Uranium
Spheres in Uranium Solution Cylinders," Nucl. Sci. Eng., 54, 360 (Jul. 1974).
G-JA085. D. Dickinson, "Calculations on Raschig-Ring-Poisoned Plutonium Solution Systems Using
Hansen-Roach Cross Sections," Nucl. Sci. Eng., 54, 367 (Jul. 1974).
G-JA086. E.D. Clayton, "Anomalies of Criticality," Nucl. Technol., 23, 14 (Jul. 1974).
G-JA087. C.L. Schuske, D. Dickinson, and S.J. Altschuler, "Surface Density Method Employing Unit
Shape Factor (s/v) for the Storage of Fissile Materials," Nucl. Technol., 23, 157 (Aug. 1974).
G-JA088. Grover Tuck, "Simplified Methods of Estimating the Results of Accidental Solution
Excursions," Nucl. Technol., 23, 177 (Aug. 1974).
G-JA089. R.C. Lloyd, S.R. Bierman, and E.D. Clayton, "Criticality of Plutonium-Uranium Mixtures
Containing 5 to 8 wt% Plutonium," Nucl. Sci. Eng., 55, 51 (Sep. 1974).
G-JA090. S.R. Bierman and E.D. Clayton, "Critical Experiments to Measure the Neutron Poisoning
Effects of Copper and Copper-Cadmium Plates," Nucl. Sci. Eng., 55, 58 (Sep. 1974).
G-JA091. C.R. Richey, "Reexamination of the Value for the Minimum Critical Concentration of
Plutonuim-239 (sic) in Water," Nucl. Sci. Eng., 55, 244 (Oct. 1974).
G-JA092. Robert E. Rothe, "The Influence of Neutron Poisons on High-Concentration Plutonium
Solutions," Nucl. Sci. Eng., 55, 482 (Dec. 1974).
G-JA093. C.L. Schuske and D. Dickinson, "Criticality Design of a Large-Capacity Plutonium Melting
Crucible," Nucl. Technol., 25, 72 (Jan. 1975).
G-JA094. Robert E. Rothe, Donald L. Alvarez, and Harold E. Clark, "The Criticality of Periodically
Boron-Poisoned Enriched Uranium Solution Systems,"
Nucl. Technol., 25, 502 (Mar. 1975).
G-JA095. S.J. Altschuler and C.L. Schuske, "Fissile-Solution Storage in Pipes Isolated by Concrete
or Water," Nucl. Technol., 26, 23 (May 1975).
G-JA096. C.L. Schuske and J.D. McCarthy, "Rocky Flats Experience with Borosilicate Glass Rings
for Criticality Control," Nucl. Technol., 26, 254 (Jul. 1975).
G-JA097. Deanne Dickinson, "Nominally Reflected Pipe Intersections Containing Fissile Solution,"
Nucl. Technol., 26, 265 (Jul. 1975).
G-JA098. S.R. Bierman, "Critical Experiments-Benchmarks (Pu-U Systems)," Nucl. Technol., 26,
352 (Jul. 1975).
G-JA099. V. Drke, et al., "Experimental and Theoretical Studies of Criticality Safety by Ingress of
Water in Systems with Pebble-Bed High-Temperature Gas-Cooled Reactor Fuel," Nucl. Sci. Eng., 57,
328 (Aug. 1975).
G-JA100. R.C. Lloyd and E.D. Clayton, "Criticality Safety Data Applicable to Processing Liquid-Metal
Fast Breeder Reactor Fuel," Nucl. Sci. Eng., 59, 21 (Jan. 1976).
G-JA101. O.C. Kolar, H.F. Finn, and N.L. Pruvost, "Livermore Plutonium Array Program: Experiments
and Calculations," Nucl. Technol., 29, 57 (Apr. 1976).
G-JA102. R.C. Lloyd and E.D. Clayton, "Criticality of Plutonium-Uranium Nitrate Solutions," Nucl.
Sci. Eng., 60, 143 (Jun. 1976).
G-JA103. C.L. Schuske and Hugh C. Paxton, "History of Fissile Array Measurements in the United
States," Nucl. Technol., 30, 101 (Aug. 1976).
G-JA104. Douglas C. Hunt, "A Review of Criticality Safety Models Used in Evaluating Arrays of
Fissile Materials," Nucl. Technol., 30, 138 (Aug. 1976).
G-JA105. Deanne Dickinson and G. Elliott Whitesides, "The Monte Carlo Method for Array Criticality
Calculations," Nucl. Technol., 30, 166 (Aug. 1976).
G-JA106. Douglas C. Hunt and Deanne Dickinson, "Comparative Calculational Evaluation of Array
Criticality Models," Nucl. Technol., 30, 190 (Aug. 1976).
G-JA107. S.R. Bierman and E.D. Clayton, "Critical Experiments with Low-Moderated Homogeneous
Mixtures of Plutonium and Uranium Oxides Containing 8, 15, and 30 wt% Plutonium," Nucl. Sci.
Eng., 61, 370 (Nov. 1976).
G-JA108. S.R. Bierman, "Reactivity Measurements under Conditions Typical to Fuel Element
Dissolution," Nucl. Technol., 31, 339 (Dec. 1976).
G-JA109. J.T. Thomas, "Surface Density and Density Analog Models for Criticality in Arrays of
Fissile Materials," Nucl. Sci. Eng., 62, 424 (Mar. 1977).
G-JA110. R.C. Lloyd and E.D. Clayton, "Effect of Fixed and Soluble Neutron Absorbers on the
Criticality of Uranium-Plutonium Systems," Nucl. Sci. Eng., 62, 726 (Apr. 1977).
G-JA111. E.D. Clayton and B.M. Durst, "Comment on the Interpretation and Application of Limiting
Critical Concentrations of Fissile Nuclides in Water," Nucl. Technol., 33, 110 (Apr. 1977).
G-JA112. Jack K. Thompson, "Minimum Critical Mass of Plutonium-Polyethylene System Found To
be Significantly Lower than Plutonium-Water System," Nucl. Technol., 33, 235 (mid-Apr. 1977).
G-JA113. B.L. Koponen, "Reactivity Enhancement in Transportation and Storage Arrays Due to
Fissile Material Density Reductions," Nucl. Technol., 34, 242 (Jul. 1977).
G-JA114. E.D. Clayton, H.K. Clark, D.W. Magnuson, J.H. Chalmers, Gordon Walker, N. Ketzlach,
Ryohei Kiyose, C.L. Brown, D.R. Smith, and R. Artigas, "Basis for Subcritical Limits in Proposed
Criticality Safety Standard for Mixed Oxides," Nucl. Technol., 35, 97 (mid-Aug. 1977).
G-JA115. C. Devillers and P. Blum, "Neutron Multiplication and Shielding Problems in Pressurized
Water Reactor Spent Fuel Shipping Casks," Nucl. Technol., 35, 112 (mid-Aug. 1977).
G-JA116. J.J. Koelling, G.E. Hansen, and C.C. Byers, "Fission and Explosive Energy Releases of
PuO2, PuO2-UO2, UO2, and UO3 Assemblies," Nucl. Technol., 35, 611 (Oct. 1977).
G-JA117. G. Schulze and H. Wrz, "In-Line Determination of Fissile Material in High-Activity
Solutions," Nucl. Technol., 35, 663 (Oct. 1977).
G-JA118. S.R. McNeany and J.D. Jenkins, "An Evaluation of ENDF/B-IV and Hansen-Roach
Uranium-233 Cross Sections for Use in Criticality Calculations," Nucl. Sci. Eng., 65, 441 (Mar.
1978).
G-JA119. Charles R. Marotta, "A Simple Relationship of Maximum /k Due to Compaction of
Unmoderated Fissile Materials," Nucl. Technol., 39, 323 (Aug. 1978).
G-JA120. J.T. Thomas, "Reflectors, Infinite Cylinders, Intersecting Cylinders, and Nuclear
Criticality," Nucl. Sci. Eng., 67, 279 (Sep. 1978).
G-JA121. Francis H. Lewis and Patrick D. Soran, "Calculations of UH3 Critical Masses for Various
Uranium Enrichments," Nucl. Sci. Eng., 68, 116 (Oct. 1978).
G-JA122. J.S. Philbin and S.A. Dupree, "Subcriticality Guidelines for Liquid-Metal Fast Breeder
Reactor Spent Fuel Shipping Cask Designs," Nucl. Technol., 40, 284 (Oct. 1978).
G-JA123. George C. Wu and Lawrence Ruby, "Possible Criticality of Protactinium-231," Nucl. Sci.
Eng., 68, 349 (Dec. 1978).
G-JA124. Robert E. Rothe and Inki Oh, "Benchmark Critical Experiments on High-Enriched Uranyl
Nitrate Solution Systems," Nucl. Technol., 41, 207 (Dec. 1978).
G-JA125. Inki Oh and Robert E. Rothe, "A Calculational Study of Benchmark Critical Experiments on
High-Enriched Uranyl Nitrate Solution Systems," Nucl. Technol., 41, 226 (Dec. 1978).
G-JA126. William D. Bromley and James S. Olszewski, "Safety Calculations and Benchmarking of
Babcock and Wilcox Designed Close Spaced Fuel Storage Racks," Nucl. Technol., 41, 341 (mid-Dec.
1978).
G-JA127. R.C. Lloyd, S.W. Heaberlin, E.D. Clayton, and R.D. Carter, "Assessment of Criticality
Safety," Nucl. Technol., 42, 13 (Jan. 1979).
G-JA128. Shi-Ping and Duaine G. Lindstrom, "A Random Geometry Model in Criticality Calculations
of Solutions Containing Raschig Rings," Nucl. Sci. Eng., 69, 363 (Mar. 1979).
G-JA129. S.R. Bierman, B.M. Durst, and E.D. Clayton, "Critical Separation Between Subcritical
Clusters of Low Enriched UO2 Rods in Water with Fixed Neutron Poisons," Nucl. Technol., 42, 237
(Mar. 1979).
G-JA130. S.R. Bierman, B.M. Durst, E.D. Clayton, R.I. Scherpelz, and Howard T. Kerr, "Critical
Experiments with Fast Test Reactor Fuel Pins in Water," Nucl. Technol., 44, 141 (Jun. 1979).
G-JA131. Anil Kumar and M. Srinivasan, "Comments on `A Simple Relationship of Maximum /k
Due to Compaction of Unmoderated Fissile Materials' and the Use of the Trombay Criticality Formula
for the Same," Nucl. Technol., 44, 322 (Jul. 1979).
G-JA132. J.C. Manaranche, et al., "Critical Experiments with Lattices of 4.75-wt%-235U-Enriched
UO2 Rods in Water," Nucl. Sci. Eng., 71, 154 (Aug. 1979).
G-JA133. R.C. Lloyd, B.M. Durst, and E.D. Clayton, "Effect of Soluble Neutron Absorbers on the
Criticality of Low-Uranium-235-Enriched UO2 Lattices," Nucl. Sci. Eng., 71, 154 (Aug. 1979).
G-JA134. G.E. Hansen and H.C. Paxton, "Thor, A Thorium-Reflected Plutonium-Metal Critical
Assembly," Nucl. Sci. Eng., 71, 287 (Sep. 1979).
G-JA135. G.E. Hansen and H.C. Paxton, "A Critical Assembly of Uranium Enriched to 10% in
Uranium-235," Nucl. Sci. Eng., 72, 230 (Nov. 1979).
G-JA136. W. Seifritz and P. Wydler, "Criticality of Neptunium-237 and Its Possible Utilization in
Nuclear Reactors," Nucl. Sci. Eng., 72, 272 (Nov. 1979).
G-JA137. S.R. Bierman, B.M. Durst, and E.D. Clayton, "Criticality Experiments with Subcritical
Clusters of Low Enriched UO2 Rods in Water with Uranium or Lead Reflecting Walls," Nucl. Technol.,
47, 51 (Jan. 1980).
G-JA138. B.M. Durst, S.R. Bierman, E.D. Clayton, and J.F. Mincey, "Critical Experiments with Solid
Neutron Absorbers and Water-Moderated Fast Test Reactor Fuel Pins," Nucl. Technol., 48, 128 (mid-
Apr. 1980).
G-JA139. H.K. Clark, "Subcritical Limits for Special Fissile Actinides," Nucl. Technol., 48, 164 (mid-
Apr. 1980).
G-JA140. Allen S. Benjamin and David J. McCloskey, "Spent Fuel Heatup Following Loss of Water
During Storage," Nucl. Technol., 49, 274 (Jul. 1980).
G-JA141. K. Subba Rao and M. Srinivasan, "BeH2 as a Moderator in Minimum Critical Mass
Systems," Nucl. Technol., 49, 315 (Jul. 1980).
G-JA142. J.C. Manaranche, et al., "Dissolution and Storage Experiment with 4.75-wt%-235U-
Enriched UO2 Rods," Nucl. Technol., 50, 148 (Sep. 1980).
G-JA143. P.K. Job, M. Srinivasan, and V.R. Nargundkar, "Analysis of the BeO-Reflected Uranium-
233 Nitrate Solution Subcritical Multiplication Experiments Conducted at Purnima Laboratories,"
Nucl. Technol., 51, 87 (Nov. 1980).
G-JA144. Gary S. Hoovler, et al., "Critical Experiments Supporting Close Proximity Water Storage of
Power Reactor Fuel," Nucl. Technol., 51, 217 (Dec. 1980).
G-JA145. S.R. Bierman and E.D. Clayton, "Criticality Experiments with Subcritical Clusters of Light
Water Reactor Type Fuel Separated by a Flux Trap," Nucl. Technol., 51, 342 (Mar. 1981).
G-JA146. R.C. Lloyd, S.R. Bierman, E.D. Clayton, and B.M. Durst, "Criticality of 4.3 wt% Uranium-
235 Enriched UO2 Rods in Uranyl Nitrate Solution Containing Gadolinium," Nucl. Sci. Eng., 78, 121
(Jun. 1981).
G-JA147. S.R. Bierman and E.D. Clayton, "Criticality Experiments with Subcritical Clusters of 2.35
and 4.31 wt% 235U-Enriched UO2 Rods in Water with Steel Reflecting Walls," Nucl. Technol., 54,
131 (Aug. 1981).
G-JA148. H.K. Clark, "Subcritical Limits for Plutonium Systems," Nucl. Sci. Eng., 79, 65 (Sep.
1981).
G-JA149. R.M. Westfall, "Critical Masses for the Even-Neutron-Numbered Transuranium Actinides,"
Nucl. Sci. Eng., 79, 237 (Oct. 1981).
G-JA150. J. Ernest Wilkins, Jr. and Keshav N. Srivastava, "Minimum Critical Mass Nuclear Reactors.
Part I," Nucl. Sci. Eng., 82, 307 (Nov. 1982).
G-JA151. J. Ernest Wilkins, Jr. and Keshav N. Srivastava, "Minimum Critical Mass Nuclear Reactors.
Part II," Nucl. Sci. Eng., 82, 316 (Nov. 1982).
G-JA152. R.C. Lloyd, R.A. Libby and E.D. Clayton, "The Measurement of Eta and the Limiting
Concentration of 239Pu in Critical Aqueous Solutions," Nucl. Sci. Eng., 82, 325 (Nov. 1982).
G-JA153. Anil Kumar, M. Srinivasan, and K. Subba Rao, "Characterization of Neutron Leakage
Probability, keff, and Critical Core Surface Mass Density of Small Reactor Assemblies Through the
Trombay Criticality Formula," Nucl. Sci. Eng., 84, 155 (Jun. 1983).
G-JA154. P.K. Job, K. Subba Rao, and M. Srinivasan, "Additional Remarks on BeH2 as a Moderator
in Minimum Critical Mass Systems," Nucl. Sci. Eng., 84, 293 (Jul. 1983).
G-JA155. Edward T. Tomlinson and C.L. Brown, "Nuclear Criticality Safety Considerations in Design
of Dry Fuel Assembly Storage Arrays," Nucl. Technol., 63, 347 (Nov. 1984).
G-JA156. Yasushi Nomura and Takanori Shimooke, "A Multiple Regression Equation for Calculated
keff Bias Errors by Criticality Code System," Nucl. Technol., 65, 340 (May 1984).
G-JA157. R.E. Kelley and E.D. Clayton, "Fissible: A Proposed New Term in Nuclear Engineering,"
Nucl. Sci. Eng., 91, 481 (Dec. 1985).
G-JA158. Philip F. Rose, "Monte Carlo Analysis of Aqueous 235U Critical Assemblies with ENDF/B-V
Data," Nucl. Sci. Eng., 94, 36 (Sep. 1986).
G-JA159. Eugene D. Clayton, Hugh K. Clark, Gordon Walker, and Richard A. Libby, "Basis for
Extending Limits in ANSI Standard for Mixed Oxides to Heterogeneous Systems," Nucl. Technol., 75,
225 (Nov. 1986).
G-JA160. Francis Barbry and Raymond Prigent, "The EDAC System and New Developments under
Consideration at the Commissariat A L'Energie Atomique for Criticality Accident Detection," Nucl.
Technol., 78, 320 (Sep. 1987).
G-JA161. Raymond C. Lloyd, E. Duane Clayton, Robert E. Wilson, Robert C. McBroom, and Robert
R. Jones, "Critical Experiments Using High-Enriched Uranyl Nitrate with Cadmium Absorber," Nucl.
Technol., 79, 82 (Oct. 1987).
G-JA162. Tsuyoshi Misawa, Seiji Shiroya, and Keiji Kanda, "Study on Criticality of a Light Water
Moderated and Reflected Coupled Core with Highly Enriched Uranium Fuel," Nucl. Technol., 83, 162
(Nov. 1988).
G-JA163. Masaki Suwa and Atsuyuki Suzuki, "An Inference Model for Predicting a Pinching Effect in
the Co-Decontamination Extraction Process in a Purex Fuel Reprocessing Plant," Nucl. Technol., 85,
187 (May 1989).
G-JA164. M. Srinivasan, K. Subba Rao, S.B. Garg, and G.V. Acharya, "Systematics of Criticality
Data of Special Actinide Nuclides Deduced through the Trombay Critical Formula," Nucl. Sci. Eng.,
102, 295 (Jul. 1989).
G-JA165. Cecil V. Parks, Robert M. Westfall, and B.L. Broadhead, "Criticality Analysis Support for
the Three Mile Island Unit 2 Fuel Removal Operations," Nucl. Technol., 87, 660 (Nov. 1989).
G-JA166. Gerald L. Palau, "Criticality Prevention during Postaccident Decontamination of Three Mile
Island Unit 2 Plant Systems," Nucl. Technol., 87, 679 (Nov. 1989).
G-JA167. Bernard R. Bandini and Anthony J. Baratta, "Potential for Recriticality of the Relocated
Core," Nucl. Technol., 87, 926 (Dec. 1989).
G-JA168. Daniel S. Williams, John C. Rommel, and Raymond L. Murray, "An Overview of Nuclear
Criticality Safety Analyses Performed to Support Three Mile Island Unit 2 Defueling," Nucl. Technol.,
87, 1134 (Dec. 1989).
G-JA169. Jnos Gad and Istvn Vidovszky, "Influence of Water Density Change on the Criticality
in Special Fuel Lattices," Nucl. Sci. Eng., 104, 217 (Mar. 1990).
G-JA170. T. P. McLaughlin, "Process criticality accident likelihoods, consequences and emergency
planning," Nucl. Energy, 31, 143 (Apr. 1992).
G-JA171. William A. Boyd and Mark W. Fecteau, "Criticality Safety of Transuranic Storage Arrays at
the Waste Isolation Pilot Plant," Nucl. Technol., 104, 207 (Nov. 1993).
G-JA172. C.F. Haught, B. Basoglu, R.W. Brewer, D.F Hollenbach, A.D. Wilkinson, H.L. Dodds, and
R.L. Oxenham, "Criticality Safety Analysis of a Calciner Exit Chute," Nucl. Technol., 105, 3 (Jan.
1994).
G-JA173. B. Basoglu, R.W. Brewer, C.F. Haught, D.F Hollenbach, A.D. Wilkinson, H.L. Dodds, and
P.F. Pasqua, "Simulation of Hypothetical Criticality Accidents Involving Homogeneous Damp Low-
Enriched UO2 Powder Systems," Nucl. Technol., 105, 14 (Jan. 1994).
G-JA174. Tracy R. Wenz and Robert D. Busch, "Modeling of Central Reactivity Worth
Measurements in Lady Godiva," Nucl. Technol., 105, 31 (Jan. 1994).
G-JA175. D.F Hollenbach and L.M. Petrie, "Vectorization Methods Development for a New Version
of the KENO-V.a Criticality Safety Code," Nucl. Sci. Eng., 116, 147 (Mar. 1994).
G-JA176. Edward P. Ficaro and David K. Wehe, "KENO-NR: A Monte Carlo Code for Simulating
252Cf-Source-Driven Noise Measurements to Determine Subcriticality," Nucl. Sci. Eng., 117, 158 (Jul.
1994).
G-JA177. Gary R. Smolen, Sidney R. Bierman, and Nubuo Fukumura, "Criticality Data and Validation
Studies of Arrays of Mixed-Oxide Fuel Pins in Aqueous and Organic Solutions," Nucl. Technol., 107,
285 (Sep. 1994).
G-JA178. Gary R. Smolen, Raymond C. Lloyd, and Hideyuki Funabashi, "Criticality Data and
Validation Studies of Plutonium-Uranium Nitrate Solutions in Cylindrical and Slab Geometry," Nucl.
Technol., 107, 304 (Sep. 1994).
G-JA179. Gary R. Smolen, Raymond C. Lloyd, and Tomozo Koyama, "Criticality Data and Validation
Studies of Plutonium-Uranium Nitrate Solutions in Annular Geometry," Nucl. Technol., 107, 326
(Sep. 1994).
G-JA180. Gary R. Smolen, Raymond C. Lloyd, and Tadakuni Matsumoto, "Criticality Data and
Validation Studies of Mixed-Oxide Fuel Pin Arrays in Pu+U+Gd Nitrate," Nucl. Technol., 107, 340
(Sep. 1994).
G-JA181. S. Das, "Comment on 'Reactivity Feedback Mechanisms in Aqueous Fissile Solutions,'"
Nucl. Sci. Eng., 118, 127 (Oct. 1994).
G-JA182. Yasushi Nomura and Hiroshi Okuno, "Simplified Evaluation Models for Total Fission
Number in a Criticality Accident," Nucl. Technol., 109, 142 (Jan. 1995).
G-JA183. Bryan L. Broadhead, "Feasibility Assessment of Burnup Credit in the Criticality Analysis of
Shipping Casks with Boiling Water Reactor Spent Fuel," Nucl. Technol., 110, 1 (Apr. 1995).
G-JA184. Jos M. Conde and Manuel Recio, "Evaluation of Burnup Credit for Fuel Storage Analysis -
- Experience in Spain," Nucl. Technol., 110, 22 (Apr. 1995).
G-JA185. Jack K. Boshoven, "Burnup Credit Experiences with the GA-4 Cask," Nucl. Technol., 110,
33 (Apr. 1995).
G-JA186. Yoshitaka Naito, Makoto Takano, Masayoshi Kurosawa, and Takenori Suzaki, "Study on
the Criticality Safety Evaluation Method for Burnup Credit in JAERI," Nucl. Technol., 110, 40 (Apr.
1995).
G-JA187. Stephen M. Bowman, Mark D. DeHart, and Cecil V. Parks, "Validation of SCALE-4 for
Burnup Credit Applications," Nucl. Technol., 110, 53 (Apr. 1995).
G-JA188. R.H. Kimpland and D.E. Kornreich, "A Two-Dimensional Multiregion Computer Model for
Predicting Nuclear Excursions in Aqueous Homogeneous Solution Assemblies," Nucl. Sci. Eng., 122,
204 (Feb. 1996).
G-JA189. Matjaz Ravnik and Bogdan Glumac, "TRIGA Spent Fuel Storage Criticality Analysis," Nucl.
Technol., 114, 365 (Jun. 1996).
G-JA190. William E. Kastenberg, Per F. Peterson, Joonhong Ahn, J. Burch, G. Casher, Paul L.
Cambr, Ehud Greenspan, Donald R. Olander, and Jasmina L. Vujic; B. Bessinger, Neville G. W.
Cook, Fiona M. Doyle, and L. Brun Hilbert, Jr., "Considerations of Autocatalytic Criticality of Fissile
Materials in Geologic Repositories, " Nucl. Technol., 115, 298 (Sep. 1996).
G-JA191. Bogdan Glumac, Matjaz Ravnik, and Marjan Logar, "Criticality Safety Assessment of a
TRIGA Reactor Spent-Fuel Pool Under Accident Conditions," Nucl. Technol., 117, 248 (Feb. 1997).
G-JA192. Yoshinori Miyoshi, Takuya Umano, Kotaro Tonoike, Naoki Izawa, Susumu Sugikawa, and
Shuji Okazaki, "Critical Experiments on 10% Enriched Uranyl Nitrate Solution Using a 60-cm-
Diameter Cylindrical Core," Nucl. Technol., 118, 69 (Apr. 1997).
G-JA193. Lane S. Paschal, C. L. Bentley, Michael E. Dunn, S. Goluoglu, R. E. Pevey, and H. L.
Dodds, "Criticality Safety Evaluation of Shutdown Diffusion Cascade Coolers," Nucl. Technol., 119,
295 (Sep. 1997).
G-JA194. Michael E. Dunn, C. L. Bentley, S. Goluoglu, Lane S. Paschal, L. M. Petrie, and H. L.
Dodds, "Development of a Continuous Energy Version of KENO V.a," Nucl. Technol., 119, 306 (Sep.
1997).
G-JA195. Ehud Greenspan, J. Vujic, and J. Burch, "Neutronic Analysis of Critical Configurations in
Geologic Repositories: I -- Weapons-Grade Plutonium," Nucl. Sci. Eng., 127, 262 (Nov. 1997).
G-JA196. Yasushi Nomura and Yoshitaka Naito, "Fault-Tree Analysis of Criticality in a Pulsed
Column of a Typical Reprocessing Facility," Nucl. Technol., 121, 3 (Jan. 1998).
G-JA197. Michelle Pitts, Farzad Rahnema, Tom G. Williamson, and Fitz Trumble, "Description of
Critical Experiments in Spherical Geometry Contining UO2F2 Solution," Nucl. Technol., 122, 1 (Apr.
1998).
G-JA198. J. Vujic and Ehud Greenspan, "Neutronic Analysis of Critical Configurations in Geologic
Repositories -- II: Highly Enriched Uranium," Nucl. Sci. Eng., 129, 1 (May 1998).
G-JA199. Marko Maucec, Matjaz Ravnik, and Bogdan Glumac, "Criticality Analysis of the
Multiplying Material Inside the Chernobyl Sarcophagus," Nucl. Technol., 122, 255 (Jun. 1998).
G-JA200. Rene Sanchez, William Myers, David Hayes, Robert Kimpland, Peter Jaegers, Richard
Paternoster, Stephen Rojas, Richard Anderson, and William Stratton, "Criticality Characteristics of
Mixtures of Plutonium, Silicon Dioxide, Nevada Tuff, and Water," Nucl. Sci. Eng., 129, 187 (Jun.
1998).
G-JA201. Kenneth D. Wright, James S. Tulenko, and Edward T. Dugan, "Comparison Between
MCNP and Critical Experiments -- A Determination of Bias Values To Be Utilized in Licensing
Calculations for High-Level Radioactive Waste Disposal," Nucl. Technol., 123, 259 (Sep. 1998).
G-JA202. J. T. Mihalczo, J. J. Lynn, and J. R. Taylor, "The Central Void Reactivity in the Oak Ridge
National Laboratory Enriched Uranium (93.2) Metal Sphere," Nucl Sci. Eng., 130, 153 (Sep. 1998).
G-JA203. Yu. E. Titarenko, O. V. Shvedov, M. M. Igumnov, E. I. Karpikhin, V. F. Batyaev, A. V.
Lopatkin, V. I. Volk, A. Yu. Vakhrushin, S. V. Shepelkov, S. G. Mashnik, and T. A. Gabriel,
"Experimental Determination and Simulation of the Reactivity Effects and Reaction Rate Sensitivity to
Different Ranges of Neutron Energy in Homogeneous Heavy Water Solutions of Thorium," Nucl. Sci.
Eng., 130, 165 (Oct. 1998).
G-JA204. Toru Obara and Hiroshi Sekimoto, "Effect of Low-Energy Resonance Absorber on Positive
Neutron Temperature Coefficient of Dilute Plutonium-Water Solution," Nucl. Sci. Eng., 130, 386
(Nov. 1998).
***
G-PA001. A.Santamarina, "Experimental Qualification of the Calculated Subcriticality in High Density
Fuel Storage," Proc. of the [ANS] Topical Meeting on Advances in Reactor Physics and Core Thermal
Hydraulics, NUREG/CP-0034, 2, 937, U.S. Nuclear Regulatory Commission (September 22-24,
1982).
G-PA002. J.T. Mihalczo, V.K. Par, and E.D. Blakeman, "Fission Chain Length Effects on
Californium-Source-Driven Subcriticality Measurements," Proc. of the [ANS] International Topical
Meeting: Advances in Mathematics, Computations, and Reactor Physics, 1, 5.1 1-1, Pittsburgh (Apr.
28 - May 2, 1991).
G-PA003. W.B. Doub and T.M. Sutton, "Analysis of the 252Cf Noise Technique to Measure
Subcriticality by the Method of Probability Generating Functions," Proc. of the [ANS] International
Topical Meeting: Advances in Mathematics, Computations, and Reactor Physics, 1, 5.1 2-1,
Pittsburgh (Apr. 28 - May 2, 1991).
G-PA004. T.M. Sutton and W.B. Doub, "Stochastic Transport Theory Analysis of the 252Cf Source
Driven Noise Technique," Proc. of the [ANS] International Topical Meeting: Advances in
Mathematics, Computations, and Reactor Physics, 1, 5.1 3-1, Pittsburgh (Apr. 28 - May 2, 1991).
G-PA005. J.P. Weinman, M.R. Mendelson, and T.M. Sutton, "KAPL Analysis of Oak Ridge
Subcritical Reactivity Measurements for Uranyl Nitrate Solutions," Proc. of the [ANS] International
Topical Meeting: Advances in Mathematics, Computations, and Reactor Physics, 1, 5.1 4-1,
Pittsburgh (Apr. 28 - May 2, 1991).
G-PA006. Edward P. Ficaro and David K. Wehe, "Monte Carlo Simulations of the Cf-252-Source-
Driven Noise Analysis Measurements for Determining Subcriticality," Proc. of the [ANS] International
Topical Meeting: Advances in Mathematics, Computations, and Reactor Physics, 1, 5.2 2-1,
Pittsburgh (Apr. 28 - May 2, 1991).
G-PA007. W.G. Winn and R.C. Hochel, "Confirming Criticality Safety od TRU Waste with Neutron
Measurements and Risk Analyses," Proc. of the 1992 [ANS] Topical Meeting on Advances in Reactor
Physics, 1-135, Charleston (Mar. 8-11, 1992).
G-PA008. S.C. Chay, "Probabilistic Risk Assessment of Drum and Culvert Containing Suspect FB-
Line TRU Waste," Proc. of the 1992 [ANS] Topical Meeting on Advances in Reactor Physics, 1-147,
Charleston (Mar. 8-11, 1992).
***
G-P001. Proc. of the 1993 [ANS] Topical Meeting on Physics and Methods in Criticality Safety, 38
papers, 260pp, Nashville (Sep. 19-23, 1993).
G-P002. Proc. of the Fifth International Conference of Nuclear Criticality Safety, 141 papers, 2
vols., Albuquerque (Sep. 17-21, 1995).
G-P003. Proc. of the 1997 [ANS] Topical Meeting on Criticality Safety Challenges in the Next
Decade, 62 papers, 424pp, Chelan (Sep. 7-11, 1997).
INDEX
Absorber . . . . . . . 5, 6, 10, 21, 35, 73, 89, 90, 100, 102, 205
Accident Yield Estimationxvii, 4-7, 10, 12-16, 18, 20, 22, 26, 29, 31, 35-40, 42-44, 46, 48, 51,
55-59, 61, 63-65, 69-71, 73, 75, 77, 78, 80, 81, 83, 88, 92-96, 98, 101, 102, 107,
118, 119, 123, 141, 143, 148, 149, 157, 205-207
AEG. . . . . . . . . . . . . . . . . . . . xvii, 178-180, 192, 193
American National Standards Institute. . . . .xvii, 6, 8, 115, 134
Analysisxvii, 2, 4, 6, 10, 12, 15, 18, 25, 26, 29, 30, 39, 40, 45, 48, 51, 55-57, 59, 61, 69-71,
74, 77, 78, 88, 94-96, 101-108, 110, 111, 114-116, 120, 122-126, 128, 130-133,
135-138, 140-143, 1, 147, 148, 153-155, 176, 178, 197, 198, 204-209
Areas of Applicabilityiii, 1, 6, 11, 24, 61, 78, 81, 112, 113, 127, 128, 175, 179-182, 184
Average Energy Group . . . . . . . . . . . . . xvii, 178, 180, 192
Calculation. . . . . . . . .2, 19, 65, 85, 175, 179, 183, 184, 196
CFR. . . . . . . . . . . . . . . . . . . . . . . xvii, 3-5, 37, 51
Code of Federal Regulations. . . . . . . . . . . . . . xvii, 5, 37
Computation. . . . . . . . . . . . . . . . . . . . . . . .138, 179
Concentration11, 47, 69, 70, 73, 74, 77, 81, 82, 84, 89, 90, 92, 93, 97-102, 106, 108, 115,
119, 140, 156, 157, 160, 197, 200, 204
Configuration Control12, 19, 23, 24, 26, 29, 30, 32, 37-42, 61, 78, 106, 112, 116, 119,
161-164, 166, 174, 176
Contractor President . . . . . . . . . . . . . . . . . . . .28, 31
Credible10, 12, 13, 15, 22, 38-40, 46, 48, 51, 55-57, 63-67, 70-76, 81, 84, 86, 87, 92, 96,
101, 107, 109, 110, 114, 115, 138, 140, 141, 148, 149, 157, 179
Criticality Accident Alarm System. . . . . . . . . xvii, 6, 35, 44
Criticality Accident Detection System. . . . . . . . . . .xvii, 14
Criticality Safety Controls6, 26, 35-37, 42, 69, 72, 78-81, 103, 105-107, 109, 111, 138, 148
Criticality Safety Organization21, 28, 30-32, 34-36, 40, 41, 43, 46, 54, 55, 61, 72, 101, 116,
119, 161-164, 171
DBA. . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii, 10
Density11, 20, 47, 53, 55, 56, 64, 67, 69, 74, 81, 82, 86, 89, 90, 93, 115, 119, 139, 140,
145, 175, 177, 178, 193, 196, 197, 200-202, 204, 205, 209
Design Basis Accident. . . . . . . . . . . . . . . . . . .xvii, 96
Design Reviews . . . . . . . . . . . . . . . . . . 32, 40, 77, 101
Emergency Planning . . . . . . . . . . . . . . .7, 32, 36, 51, 205
Engineering and Projects Organization. . . . . . . . . . .xvii, 35
Enrichment16, 20, 24, 49, 54, 64, 69, 76, 77, 81, 82, 84, 88, 92, 93, 97, 99, 100, 107, 115,
119, 140, 178, 196, 197
Evaluationxvii, 10, 12, 13, 15, 18, 23, 27, 28, 38-40, 44, 46, 55, 56, 65-68, 72, 77, 83, 89,
110, 111, 114, 115, 116, 120, 122-139, 141, 149, 168, 175, 177-179, 197, 199,
201, 202, 206, 207
Facility Operations Managers3-5, 12, 13, 15, 16, 18, 19, 22, 23, 29-35, 37-46, 48-51, 53, 55,
56, 58, 61, 64-66, 68, 70-73, 77, 78, 80, 81, 83, 84, 86, 89, 92, 94-96, 98, 99, 101,
105-111, 115, 117, 119-126, 128, 129-142, 148, 149, 154, 156, 159, 200, 207
Failure Modes and Effects Analysis . . . . . . . . . .xvii, 45, 95
FEM. . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii, 17
Fire fighting. . . . . . . . . . . . . . . .50, 51, 53, 75, 76, 90
First Line Supervision . . . . . . . . . . . . . . . . . . .33, 40
Fissible Material. . . . . . . . . 16, 17, 20, 76, 77, 89, 91, 205
Fissile Material5, 6, 8, 49, 50, 66, 91, 99, 143, 148, 196, 200, 202
Fissionable Equivalent Mass. . . . . . . . . . . .xvii, 17, 55, 56
Fissionable Materialiii, xvii, 1, 7, 11-18, 20-22, 28, 30, 31, 33-44, 46-57, 65-67, 69-77, 80-89,
91-93, 97, 99, 100, 102, 106-108, 110, 111, 114-119, 121, 122, 138-141, 156, 160, 175
FMCA . . . . . . . . . . . . . . . . . . . . . . .xvii, 12, 61, 62
FMEA . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii, 95
Geometry16, 18, 20, 65, 68, 72, 73, 77, 80-92, 97, 99, 100, 115, 119, 140, 148, 156, 179,
199, 203, 206, 207
Graded Approach. . . . . . . . . 18, 26, 38, 39, 77, 129, 138, 141
HE . . . . . . . . . . . . . . . . . . . . . . . . . . . xvii, 213
HEU. . . . . . . . . . . . . . . . . . . . . . . . . xvii, 18, 180
High Enriched Uranium. . . . . . . . . . . . . . . . . . .xvii, 18
High Explosives. . . . . . . . . . . . . . . . . . . . . . . .xvii
IEZ. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii
Immediate Evacuation Zone. . . . . . . . . . . . . . . . . . .xvii
Incident . . . . . . . . . . . . . . . 14, 18, 38, 42, 94, 95, 166
INCSRC . . . . . . . . . . . . . . . . .xvii, 28-30, 32, 40-42, 75
Installation Nuclear Criticality Safety Review Committeexvii, 28, 30, 40, 41
Interaction14, 16, 17, 20, 57, 73, 77, 84, 85, 87, 92, 97, 99, 119, 140, 177, 196
Job/Task Analysis. . . . . . . . . . . . . . . . . . xvii, 45, 125
Labeling . . . . . . . . . . . . . . .31, 34, 52-54, 106, 165, 173
LCO. . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii, 58
LCS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii
LEU. . . . . . . . . . . . . . . . . . . . . . . . . xvii, 19, 180
Limiting Control Settingxvii, 5, 7, 19, 36, 75, 81, 83, 86, 87, 93, 110, 112, 139, 156, 175, 177,
201, 204
Line/Production Management . . . . . . . . . . . . . . . 31-34, 39
Low Enriched Uranium . . . . . . . . . . . . . . . . .xvii, 19, 74
Lower Tolerance Band . . . . . . . . . . . . . . . .xvii, 180, 193
LSSS . . . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii
LTB. . . . . . . . . . . . . . . . . .xvii, 180-184, 191, 193, 194
Maintenance2, 3, 6, 25, 29, 30, 32, 33, 35, 37, 39, 41-43, 46, 50, 58, 78, 83-85, 87, 94, 99,
105, 106, 109, 111, 112, 118, 132, 154, 174
Massxvii, 6, 11, 16, 17, 19-21, 44, 53, 55, 56, 65, 67, 69, 74, 76, 77, 80-84, 86, 88, 91-93,
97, 99, 119, 138, 140, 148, 195-198, 202-204
Moderation7, 20, 47, 49, 50, 53, 54, 66-69, 75-77, 80-82, 84, 86, 88, 90-93, 97, 99, 102,
107, 115, 119, 140, 148, 157
NCSxvii, 1, 13, 25, 29-31, 35-42, 44, 46, 77, 108, 110, 111, 116, 120, 122, 141, 163, 166,
170, 172, 174
NCSS . . . . . . . . .xvii, 23, 44, 120-122, 124-132, 134-137, 173
NCSSST . . . . . . . . . . . . . . . . . . . . . . . . . xvii, 174
Nevada Test Site . . . . . . . . . . . . . . . . . . . . . . .xvii
Nonfissionable Material. . . . . . . . . . . . . . . . . . 20, 140
NTS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii
Nuclear Accident Dosimeter . . . . . . . . . . . . . . . . . . .39
Nuclear Criticality Safetyiii, iv, xvii, 1, 2, 4-10, 12, 13, 15, 17-19, 22, 23, 25-47, 50-52, 54,
64-66, 71-73, 75-81, 83-86, 89, 94-97, 100-102, 105-111, 114, 116, 118-125, 127,
129-131, 134, 138-142, 1, 148, 154, 156, 157, 159-163, 165, 171, 173, 174, 176,
178, 198, 199, 204, 205, 209
Nuclear Criticality Safety Analysis. . . 15, 18, 40, 111, 123, 141
Nuclear Criticality Safety Evaluationxvii, 10, 15, 18, 27, 46, 77, 89, 110, 114, 116, 122
NCSE. . . . xvii, 15, 27, 76, 110, 114, 116, 117, 138, 141, 142
Nuclear Criticality Safety Software System Team. . . . . xvii, 174
Nuclear Criticality Safety Specialistxvii, 23, 78, 94, 101, 121, 122, 124, 129, 138, 142
On-the-Job Training. .xvii, 114, 120, 122, 124, 125, 129, 132, 133
Operating Plans and Procedures . . . . . . . . . . . . . . . . .46
Operations Personnel . . . . . . . . . . . . . . . . . 34, 78, 110
PAG. . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii, 18
Peer . . . 21, 39, 64, 111, 114, 116, 120, 129, 132, 134, 135, 141
Poison . . . . . . . . . . . .21, 73, 74, 82, 84, 89, 97, 100, 160
Posting. . . . . . . . . . . . . . . . . . . . .31, 34, 52, 53, 92
PRA. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii
Process Limits . . . . . . . . . . . . . . . . . . . . . . .34, 53
Processing1, 3, 5, 17, 20, 21, 35, 44, 45, 47, 87, 88, 92, 107, 115, 138, 140, 148, 159, 201
Receiving and Inspecting . . . . . . . . . . . . . . . . . . . .48
Records Retention. . . . . . . . . . . . . . . . . . . . . . . .39
Reflection20, 66, 68, 69, 75-77, 81, 82, 84, 85, 91, 92, 97, 99, 115, 119, 140
Safety Analysis Report . . . . . . . . . . . . . . . . . . . . 123
Safety Limit . . . . . . . . . . . . . . . . . . xvii, 53, 70, 136
Seismic Resistance . . . . . . . . . . . . . . . . . . . . . . .58
Shielding6, 15, 20, 36, 38, 39, 56, 57, 59, 61, 66, 73, 119, 177, 202
SL . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii
SME. . . . . . . . . . . . . . . . . . . . . . . . .xvii, 132, 133
SNR. . . . . . . . . . . . . . . . . . . . . . . . .xvii, 167, 169
Software Nonconformance Reportxvii, 1, 6, 10-12, 19, 22-24, 26, 37-40, 112, 116, 161-174
Spacing6, 13, 20, 48, 50, 51, 53, 59, 61, 73, 75-77, 80-82, 84, 85, 87-89, 92, 93, 96, 97, 99,
108, 110, 115, 140
Stop Work Policy . . . . . . . . . . . . . . . . . . . . . . . .29
Storing. . . . . . . . .1, 5, 16, 26, 46, 48-50, 92, 115, 138, 140
Subject Matter Expert. . . . . . . . . . . . . . . .xvii, 127, 133
Technical Safety Requirements. . . . . . . . . . . . . . . xvii, 4
Technical Specifications . . . . . . . . . . . . . . . . . .32, 40
Trainingxvii, 1, 3, 5, 7, 25, 30-38, 40-45, 51, 62, 68, 83, 106, 114, 118-127, 129-137, 164
TSR. . . . . . . . . . . . . . . . . . . . . . . . . . . . . .xvii
Upper Subcritical Limit. . . . . . . . .xvii, 19, 70, 113, 179-181
USL. . . . . . . . . . . . .xvii, 19, 179, 181, 183, 184, 192, 193
Validationxvii, 7, 24, 26, 38, 45, 64, 112, 113, 120, 123, 126, 138, 142, 166, 171, 175, 176,
178-184, 206, 207
Verificationxvii, 7, 11, 23, 24, 33, 38, 45, 48, 89, 112, 114, 116, 138, 154, 162-168
Volume14, 16, 53, 57, 65, 67, 74, 76, 84, 86-88, 93, 96, 119, 140, 148
CONCLUDING MATERIAL
Standards Manager:
DOE-NE Preparing Activity:
DOE-OR
Review Activity:
DOE-DP, -EM, -ER,
-NE, -NS,-Al, -CH,
-ID, -NV, -OR,
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-SR,
M&O-BMI, -EG&G,
-LANL, -LLNL,
-MMES, -M&H,
-WHCO, -WINCO,
-WSR, -PNL Project Number:
SAFT-0006
User Activity:
DOE-DP, -EM, -ER,
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