Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
TO: All Unresolved Safety Issue (USI) A-46 Plant Licensees Who Are
Members of the Seismic Qualification Utility Group (SQUG)
SUBJECT: SUPPLEMENT NO. 1 TO GENERIC LETTER (GL) 87-02 THAT TRANSMITS
SUPPLEMENTAL SAFETY EVALUATION REPORT NO. 2 (SSER No. 2) ON SQUG
GENERIC IMPLEMENTATION PROCEDURE, REVISION 2, AS CORRECTED ON
FEBRUARY 14, 1992 (GIP-2)
Background
GL 87-02, "Verification of Seismic Adequacy of Mechanical and Electrical
Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46," was
issued on February 19, 1987. The generic letter was issued to implement the
USI A-46 resolution which concluded that the seismic adequacy of certain
equipment in operating nuclear power plants should be reviewed against
seismic criteria not in use when these plants were licensed. GL 87-02
requested that all recipients of the letter provide within 60 days of
receipt of the generic letter a schedule for implementation of the seismic
verification program at their facilities. In its April 10 and October 9,
1987, letters, SQUG, representing its member utilities, committed to a SQUG
generic program to develop a Generic Implementation Procedure (GIP) for use
by its members, and requested a deferment of the generic letter's 60-day
response period until after the NRC issues its final safety evaluation
report (SER) on the final GIP. By letter dated November 19, 1987, the staff
agreed that individual member utility's response to GL 87-02 could be
deferred until 60 days after the issuance of a final SER on a final GIP.
(Subsequently, the staff agreed to increase the response period to 120 days
as described in the corrected Revision 2 of the GIP issued on February 14,
1992).
Since the issuance of the staff Generic Safety Evaluation Report (GSER)
dated July 29, 1988, which contains the staff's evaluation of the GIP,
Revision 0, dated June 1988 for USI A-46, considerable progress has been
made in the development of technical criteria and procedures, as well as
some revisions to the licensing issues. As a result, SQUG incorporated
parts of the changes and additions into the GIP, Revision 1, dated December
1988. The staff reviewed the GIP, Revision 1, and issued Supplemental
Safety Evaluation Report No. 1 (SSER No. 1), on June 29, 1990.
The SQUG completed the final version of the GIP, Revision 2, as corrected on
February 14, 1992 (GIP-2), and submitted it to the NRC for review and
approval
92055190366
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on February 14, 1992. The staff has completed its review of the GIP-2. The
enclosed SSER No. 2 documents the NRC staff evaluation of GIP-2. Because
the GIP criteria and procedures described in Revision 0 and Revision 1 have
been significantly changed and improved on since the GSER was issued in
1988, the staff has modified its earlier position on various technical and
licensing issues. Therefore, this supplement (SSER No. 2) supersedes all
previous staff documents such as the GSER and SSER No. 1.
Discussion
This Supplement No. 1. to GL 87-02 transmits the staff SSER No. 2 on the
SQUG's final GIP, i.e., GIP-2. As such, SQUG's requested deferment related
to individual licensee responses to GL 87-02 has ended on the basis that the
staff has completed the evaluation of the SQUG's final GIP as agreed to with
the SQUG in the staff's November 19, 1987, letter. The staff and SQUG
discussed each of the resolutions in SSER No. 2 needed to close USI A-46.
Upon completion of the plant-specific walkdown and the third-party audit
review, the licensee should submit a plant-specific summary report,
including the results of the third-party audit review and a proposed
schedule for future modifications and replacements, where appropriate. Each
licensee should also provide a completion letter advising the NRC that any
corrective actions identified in the summary report, or agreed to with the
staff as a result of other related correspondence, have been completed. The
staff will review the licensee's submittals, conduct audits as necessary,
and issue a simple plant-specific safety evaluation report (SER) addressing
whether the licensee has complied with its commitments and whether it has
adequately resolved the USI A-46 issue. This plant-specific SER will serve
as the USI A-46 closure document for its respective docket.
Required Response, (10 CFR 50.54f)
Addressees are required to submit, pursuant to the provisions of 10 CFR
50.54(f), the following information within 120 days of the date of this
Supplement No. 1 to GL 87-02:
1. A statement whether you commit to use both the SQUG commitments and the
implementation guidance provided in GIP-2 as supplemented by the SSER No.
2 for the resolution of USI A-46. In this case, any deviation from
GIP-2, as supplemented by the SSER No. 2, must be identified, justified,
and documented. If you do not make such a commitment, you must provide
your alternative for responding to GL 87-02.
2. A plant-specific schedule for the implementation of the GIP and
submission of a report to the staff that summarizes the results of the
USI A-46
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review, if you are committing to implement GIP-2. This schedule shall be
such that each affected plant will complete its implementation and submit
the summary report within 3 years after the issuance of the SSER No. 2,
unless otherwise justified.
3. The detailed information as to what procedures and criteria were used to
generate the in-structure response spectra to be used for USI A-46 as
requested in the SSER No. 2. The licensee's in-structure response
spectra are considered acceptable for USI A-46 unless the staff indicates
otherwise during a 60-day review period.
Your response must be submitted under oath or affirmation and must be
addressed to the U.S. Nuclear Regulatory Commission, ATTN: Document Control
Desk, Washington, D.C. 20555. In addition, a copy of your response must be
submitted to the appropriate regional administrator. The justification for
this information request under 10 CFR 50.54(f) continues to be the same as
that for GL 87-02.
Backfit Discussion
The staff's request for a plant-specific schedule and information related to
the implementation of the resolution of USI A-46 was considered to be a
backfit under 10 CFR 50.109. A backfit analysis was completed and described
in NUREG-1211, "Regulatory Analysis for Resolution of Unresolved Safety
Issue A-46, Seismic Qualification of Equipment in Operating Plant," dated
February 19, 1987. Although the estimated cost to a licensee excluding the
cost of repairs to anchorages and supports of equipment in implementing the
USI A-46 program has since increased approximately 25% in total cost, the
benefits of implementing the program have also greatly increased as
documented in the package of information prepared for the Committee to
Review Generic Requirements, which reviewed this Supplement No. 1 to GL
87-02 at the Meeting No. 212. The staff has placed the package of
information in the Public Document Room for public availability.
This request is covered by Office of Management and Budget Clearance Number
3150-0011, which expires May 31, 1994. The estimated average number of
burden hours is 120 person hours per owner response, including the time
required to assess the new recommendations, search data sources, gather and
analyze the data, and prepare the required letters. These estimated average
burden hours pertain only to the identified response-related matters and do
not include the time for actual implementation of the requested action.
Comments on the accuracy of this estimate and suggestions to reduce the
burden may be directed to Ronald Minsk, Office of the Information and
Regulatory Affairs (3150-0011), NEOB-3019, Office of Management and Budget,
Washington, D.C. 20503, and to the U.S. Nuclear Regulatory Commission,
Information and Records Management Branch, (MNBB 7714), Division of
Information Support Service, Office of Information and Resources Management,
Washington, D.C. 20555.
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If you have any questions about this matter, please contact the NRC
technical contact or the lead project manager listed below.
Sincerely,
James G. Partlow
Associate Director for Projects
Office of Nuclear Reactor Regulation
Enclosures:
1. SSER No. 2
2. Listing of Recently
Issued Generic Letters
cc w/enclosures:
Seismic Qualification Utility Group
ATTN: Richard Schaffstall
Electric Power Research Institute
1019 19th Street, NW
Washington, DC 20336
Technical Contact: Pei-Ying Chen (301) 504-2789
Lead Project Manager: Patrick Sears (301) 504-2021
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SUPPLEMENTAL SAFETY EVALUATION REPORT NO. 2
ON
SEISMIC QUALIFICATION UTILITY GROUP'S
GENERIC IMPLEMENTATION PROCEDURE, REVISION 2,
CORRECTED FEBRUARY 14, 1992
FOR
IMPLEMENTATION OF GL 87-02 (USI A-46),
VERIFICATION OF SEISMIC ADEQUACY OF EQUIPMENT
IN OLDER OPERATING NUCLEAR PLANTS
.
TABLE OF CONTENTS
Page
BACKGROUND 1
GENERAL DISCUSSION 3
GENERAL EVALUATION . . . 3
DETAILED DISCUSSION AND EVALUATION 5
I Licensing and Implementation Guidelines 5
I.1.0 Introduction 5
I.1.1 Background 5
I.1.2 Purpose of the GIP 6
I.1.3 GIP Commitments and Guidance 6
I.2.0 Issues and Positions 6
I.2.1 Introduction 6
I.2.2 Interpretation and Guidelines 7
I.2.3 Compliance With Guidelines 7
I.3.0 Revisions to the GIP 9
II Generic Procedure for Plant-Specific Implementation 9
II.1 Introduction 9
II.2 Seismic Evaluation Personnel 9
II.3 Identification of Safe-Shutdown Equipment 10
II.4 Screening Verification and Walkdown 12
II.4.0 Introduction 12
II.4.1 SQUG Commitments 13
II.4.2 Seismic Capacity Compared to Seismic Demand 13
II.4.3 Equipment Class Similarity and Caveats 16
II.4.4 Anchorage Adequacy 16
II.5 Outlier Identification and Resolution 23
II.6 Relay Functionality Review 24
II.7 Tanks and Heat Exchangers Review 26
II.8 Cable and Cable Raceway Review 29
II.9 Documentation 31
II.10 Reference 31
III Appendices. . . . 32
III.1 Appendix A, Procedure for Identification of 32
Safe Shutdown Equipment
III.2 Appendix B, Summary of Equipment Class Descriptions 32
and Caveats
III.3 Appendix C, Anchorage Data 33
III.4 Appendix D, Seismic Interaction 33
III.5 Appendix E, Preparatory Work Prior to Walkdown 33
III.6 Appendix F, Screening Walkdown Plan 34
III.7 Appendix G, Screening Evaluation Work Sheets 34
CONCLUSION. . . . . . . 34
REFERENCES. . . . . . . 37
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SUPPLEMENTAL SAFETY EVALUATION REPORT NO. 2
ON SEISMIC QUALIFICATION UTILITY GROUP'S
GENERIC IMPLEMENTATION PROCEDURE, REVISION 2,
CORRECTED FEBRUARY 14, 1992,
FOR IMPLEMENTATION OF GL 87-02, USI A-46 PROGRAM
VERIFICATION OF SEISMIC ADEQUACY OF EQUIPMENT IN
OLDER OPERATING NUCLEAR PLANTS
BACKGROUND
In December 1980, the Nuclear Regulatory Commission (NRC) designated
"Seismic Qualification of Equipment in Operating Plants" as an unresolved
safety issue (USI). The safety issue of concern was that equipment in
nuclear plants for which construction permit (CP) applications had been
docketed before about 1972 had not been reviewed according to the
then-current (1980-81) licensing criteria for seismic qualification of
equipment (i.e., Regulatory Guide (RG) 1.100; Institute of Electrical and
Electronics Engineers (IEEE) Standard 344-1975, and Standard Review Plan
(SRP) Section 3.10 (NUREG-0800, July 1981)). Therefore, the seismic
adequacy of the equipment in these older plants may be questionable
regarding their ability to survive and function in the event of a
safe-shutdown earthquake (SSE). Equipment in plants with a CP application
docketed after about 1972 were qualified according to the then-current
licensing criteria and licensee compliance has been audited by the NRC
staff. All operating plants for which equipment seismic qualification could
not be verified to meet the intent of then-current licensing criteria are
subject to the implementation provisions outlined in Generic Letter (GL)
87-02, "Verification of Seismic Adequacy of Mechanical and Electrical
Equipment in Operating Reactors, Unresolved Safety Issue (USI) A-46"
(Reference 1). These plants are identified as "USI A-46 plants" and are
listed in Table A in Section II.4.2 of this report.
The applicable portions of the NRC's regulations governing the seismic
design of nuclear power plants require that structures, systems, and
components important to safety be designed to withstand the effects of
earthquakes, and that those systems and components be designed to perform
their intended safety functions (Appendix A to 10 CFR Part 50). Appendix A
to 10 CFR Part 100, which was published in the Federal Register on November
13, 1973 (38 FR 31281) and became effective December 13, 1973, requires that
the engineering method, used to insure that the required safety functions of
such structures, systems, and components are maintained during and after an
SSE, must involve the use of either a suitable dynamic analysis or a
suitable qualification test. This engineering method is termed "seismic
qualification method" for the purpose of licensing requirements. No
explicit provisions within the regulations permit the use of experience data
as a means for seismic qualification. However, the NRC has determined that
requiring those older operating plants to comply with the then-current
licensing requirements was not practicable because a literal application of
those criteria to older operating plants could require extensive
modifications of those facilities that could not be justified from the
cost-benefit standpoint.
Although no explicit provisions within the regulations permit the use of
experience data as a means for seismic qualification, the NRC concluded that
the use of earthquake experience data, with appropriate restrictions and
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caveats, supplemented by some test results to verify the seismic adequacy of
equipment within certain specified earthquake motion bounds, represented the
most reasonable and cost-effective means of ensuring that the purpose of the
NRC regulations related to seismic design can be satisfied for those plants.
Therefore, for USI A-46 plants only, rather than requiring compliance with
then-current criteria for seismic qualification of equipment, the staff
requested the USI A-46 licensees only to verify the seismic adequacy of
equipment in these plants as reflected in the title and the intent of GL
87-02 (Reference 1). One of the programmatic restrictions excludes using
earthquake experience data to verify the seismic adequacy of structures and
piping.
To address the USI A-46 issue, some of the affected utilities formed the
Seismic Qualification Utility Group (SQUG) in 1982. In 1983, the SQUG
proposed the formation of a panel of consultants, the Senior Seismic Review
and Advisory Panel (SSRAP), to independently assess and review the viability
of using earthquake experience data and test data to demonstrate equipment
ruggedness, and to provide expert advice and consultation. The SQUG
subsequently developed the "Generic Implementation Procedure (GIP) for
Seismic Verification of Nuclear Plant Equipment" for its members to use.
The SQUG submitted the GIP, Revision 0 (GIP-0), dated June 1988 (Reference
2) and the related documents and reports supporting GIP-0 to the NRC staff
for review. The staff reviewed these documents and issued a Generic Safety
Evaluation Report (GSER) on July 29, 1988 (Reference 3), recognizing that
not all sections of GIP-0 had been developed at that time.
In contrast to the IEEE Standard 344 qualification approach, which in the
past has relied on analysis or testing of each item of equipment, the GIP
methodology relies primarily on the use of existing earthquake and test
experience data to verify the seismic adequacy of generic classes of
equipment. By convention, the IEEE Standard 344 procedures have been termed
"equipment seismic qualification," while the USI A-46 procedures have been
termed "equipment seismic adequacy verification."
In December 1988, Revision 1 of the GIP (GIP-1) (Reference 4) was submitted
for NRC staff review. GIP-1 contained essentially the same technical topics
as GIP-0 except that a new section was added for evaluating tanks and heat
exchangers and some information was added for resolving outstanding issues.
While the staff was reviewing GIP-1, the SQUG was making significant changes
to Part II which meant that Part II of GIP-1 would be virtually obsolete.
Therefore, the staff focused its evaluation of GIP-1 primarily on Part I.
This evaluation can be found in the Supplemental Safety Evaluation Report
(SSER) No. 1 (Reference 5).
In September 1990, the SQUG submitted Revision 2 of the GIP (Reference 6).
The staff reviewed Revision 2 and commented on it in March 1991 (Reference
7). In response to these comments and to subsequent discussions with the
NRC staff, the SQUG further revised Revision 2, and resubmitted it with
corrected pages in February 1992 (Reference 8). This supplement (SSER No.
2) presents the results of the NRC staff's evaluation of this latest
revision. For the remainder of this report, "GIP-2" refers to Reference 8.
.
Because the criteria and procedures described in GIP-0 and GIP-1 have been
significantly changed and improved since the GSER was issued in 1988, the
staff has modified its earlier positions on various technical and licensing
issues. Therefore, this supplement (SSER No. 2) supersedes all such
previous staff documents, i.e, the GSER and SSER No. 1.
This supplement begins with a general discussion and evaluation of the
overall GIP-2 document, followed by a detailed discussion and evaluation of
specific sections of GIP-2. Where it is applicable, the staff discusses
clarifications, interpretations, positions and exceptions. The
clarifications, interpretations, positions and exceptions are not
specifically labeled in the main text of this report, but all exceptions are
specifically identified in the final conclusion section.
GENERAL DISCUSSION
GIP-2 is divided into two major parts: Part I discusses the related
licensing and implementation issues for the USI A-46 program; Part II
contains the technical information necessary for the implementation of the
program. Part II has 10 sections. They are:
1. Introduction
2. Seismic Evaluation Personnel
3. Identification of Safe Shutdown Equipment
4. Screening Verification and Walkdown
5. Outlier Identification and Resolution
6. Relay Functionality Review
7. Tanks and Heat Exchangers Review
8. Cable and Conduit Raceway Review
9. Documentation
10. References
GIP-2 provides the general guidelines in these 10 sections with detailed
procedures, technical data, and implementation work sheets given in seven
appendices. These are:
Appendix A Procedure for Identification of Safe Shutdown Equipment
Appendix B Summary of Equipment Class Descriptions and Caveats
Appendix C Anchorage Data
Appendix D Seismic Interaction
Appendix E Preparatory Work Prior to Walkdown
Appendix F Screening Walkdown Plan
Appendix G Screening Evaluation Work sheets
Parts, sections, and appendices of GIP-2 are discussed in more detail in
this SSER No. 2 in the section "Detailed Discussion and Evaluation."
GENERAL EVALUATION
In general, the NRC staff finds GIP-2 to be a very useful working document
for implementing the USI A-46 program. The information contained in GIP-2
is
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generally acceptable to the staff for a plant-specific implementation of USI
A-46.
The staff discusses clarifications, interpretations, exceptions, and
positions in the section "Detailed Discussion and Evaluation," by
referencing specific parts, sections, or appendices of GIP-2. The staff
clarifications and exceptions that are general in nature and that apply to
the entire GIP-2 are listed as follows:
1. The staff considers GIP-2 acceptable (with the clarifications,
interpretations, exceptions, and positions identified in this SSER No. 2)
for verifying seismic adequacy of equipment in USI A-46 plants only. The
NRC staff expects that, when weaknesses in existing equipment are
identified as a result of the implementation of USI A-46, licensees will
take appropriate corrective actions, including modifications or upgrades,
if necessary, to ensure that those equipment possess an adequate level of
seismic safety.
2. The staff considers GIP-2 to be a method for verifying the seismic
adequacy of equipment rather than a seismic qualification method. This
is because the criteria and many of the practices proposed in GIP-2 are
not equivalent to current seismic qualification requirements. Examples
include the following: (1) A vast amount of the technical information
that is given in GIP-2 was gathered in a very general way rather than for
each item of equipment, and was based on many subjective judgments and
opinions; (2) The implementation methodology proposed in GIP-2 allows the
review engineers to resolve major issues on the basis of their judgments,
in some cases without requiring the engineers to justify or document the
basis for these judgments, rather than on the basis of such standard
engineering practices as calculations and testing; (3) The damping values
used in GIP-2 are, in general, higher than those provided in the current
version of R.G. 1.61; (4) The practice proposed in GIP-2 for evaluating
safe-shutdown paths and identifying safe-shutdown equipment differs from
current requirements in that safety grade equipment is not required to be
available and it is not necessary to cool the reactor beyond hot shutdown
conditions, whereas current designs require safety grade systems to cool
the reactor from normal operating conditions to cold shutdown; and (5)
The practice of allowing the "rule of the box" (GIP-2, page 3-16) and
spot checking of the device mounting in a cabinet also differs from
current practice, which requires such activities as testing, inspection,
documentation, and corrective actions to be covered by a 10 CFR Part 50,
Appendix B, quality assurance program.
On the basis of the differences between current seismic qualification
requirements and the criteria and the procedures provided in GIP-2, the
NRC staff does not consider the USI A-46 methodology given in GIP-2 to be
a "seismic qualification" procedure. Rather, the staff considers the
GIP-2 methodology to be a seismic adequacy verification procedure which
was developed based on generic equipment earthquake experience data
supplemented by generic equipment test data. The implementation of the
GIP-2 approach for USI A-46 plants provides safety enhancement, in
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certain aspects, beyond the original licensing bases. Therefore, GIP-2
methodology is an acceptable evaluation method, for USI A-46 plants only,
to verify the seismic adequacy of the safe-shutdown equipment and to
satisfy the pertinent equipment seismic requirements of General Design
Criterion 2 and the purpose of the NRC regulations relevant to equipment
seismic adequacy including 10 CFR Part 100.
3. The term "licensee," as used in GIP-2, refers only to the licensee of a
plant in the USI A-46 program.
4. Some statements were made in Reference 5 of Part II of GIP-2 about the
aging effects of the database equipment. The NRC staff considers that
the scope of the USI A-46 program excludes the issue of environmental
qualification of equipment in operating plants, because this issue was
addressed by the implementation program under 10 CFR 50.49. GIP-2 does
not address the aging effects of equipment by systematic collection of
quantitative data on the earthquake experience; therefore, the staff will
not accept any claim that the experience data collected by the SQUG for
the USI A-46 program adequately addressed the aging effects of equipment,
as one might incorrectly interpret from the related statement on page 13,
Reference 5 of Part II of GIP-2.
DETAILED DISCUSSION AND EVALUATION
In the details of the staff's evaluation of GIP-2, which follows, parts,
sections, or appendices are briefly discussed, and the staff's positions,
clarifications, interpretations, and exceptions are presented.
I LICENSING AND IMPLEMENTATION GUIDELINES
Part I of GIP-2 describes the genesis of the USI A-46 program and discusses
the role of the GIP in resolving the unresolved safety issue. This part
considers several issues and describes SQUG positions on several aspects
related to licensing and implementation guidelines. These aspects include,
among other things, the interpretation of GIP-2 guidelines, the compliance
with regulations, the selection of equipment, and any future revisions of
GIP-2. The staff finds Part I acceptable subject to the following:
I.1.0 Introduction
I.1.1 Background
The second paragraph of Section 1.1 of Part I of GIP-2 states that "the
purpose of USI A-46 is to verify this conclusion" which is "that there is
adequate seismic capacity of properly anchored equipment in older operating
plants." Although this conclusion is expected to be correct in general,
there may be some pieces of equipment for which proper anchorage alone does
not demonstrate seismic adequacy. For all equipment within the USI A-46
scope, the licensee is responsible for verifying all aspects of seismic
adequacy of the equipment in accordance with GIP-2.
.
I.1.2 Purpose of the GIP
1. The third paragraph of Section 1.2 of Part I states that "Because the NRC
will document its evaluation of the GIP in a safety evaluation report
(SER), the GIP provides an NRC-accepted method to verify the seismic
adequacy of equipment...." The staff concurs with this statement,
provided that GIP-2 is used in its entirety in conjunction with and
supplemented by the clarifications, interpretations, and exceptions
identified in this supplement, and that the application of the GIP is
limited to USI A-46 plants only.
2. Section 1.2 of Part I states, "Every aspect of the Generic Letter
Procedure has been fully considered in development of the GIP.
Therefore, licensees will be guided by the GIP. By satisfying the
provisions of the GIP, licensees will have fully satisfied the guidance
of the Generic Letter...." This is generally acceptable to the staff for
GIP-2. However, deviations from GIP-2 and the SSER No. 2 without the
staff's prior approval may result in the licensee not fully satisfying
the provisions of GL 87-02.
I.1.3 GIP Commitments and Guidance
1. The second paragraph of Section 1.3 of Part I states that "USI A-46
licensees may use the GIP guidance or may substitute clearly equivalent
methods without prior notification of the NRC...." The staff's position
is that if licensees use other methods that deviate from the criteria and
procedures as described in SQUG commitments and in the implementation
guidance of GIP-2 without prior NRC staff approval, the method may not be
acceptable to the staff and, therefore, may result in a deviation from
the provisions of GL 87-02 as stated in item 2 in Section I.1.2 above.
2. The third paragraph of Section 1.3 of Part I states that "submittals
which commit to the entire GIP...shall be regarded as accepted by the
Staff upon docketing...." The staff concurs with this statement provided
that the licensee commits to the entire GIP-2 as supplemented by the
clarifications, interpretations, and exceptions identified in this
supplement.
I.2.0 Issues and Positions
I.2.1 Introduction
The third paragraph of Section 2.1 of Part I refers to SQUG documents (e.g.
References 9 and 12 of Part I of GIP-2) that summarize the resolution
histories of many issues. The staff recognizes the importance of these
documents. Although all of these many issues are resolved with or without
conditions or clarifications, these documents reflect SQUG's perception of
the resolution. The final resolutions of all issues are contained in the
GIP-2 as supplemented by this SSER No. 2.
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I.2.2 Interpretation and Guidelines
For a meaningful third-party audit (Section 2.2.7 of Part I), the NRC
expects that the auditor(s) should have broad engineering experience and
have completed the SQUG developed training course on seismic adequacy
verification of equipment in operating nuclear power plants. This is
because the third-party audit will involve substantially less time and
effort than the original walkdown and analyses. Thus, the auditor(s) should
have sufficient qualification and experience to be able to assess the
adequacy of the entire plant-specific implementation program during the
limited time of the audit.
Additionally, to provide a desired degree of assurance concerning the
effectiveness of the third-party review, a process for inter-plant
information exchange and coordination should be implemented to collect,
evaluate, and disseminate generic problems, questions, and lessons learned
during the USI A-46 plant-specific walkdowns and third-party reviews to all
member utilities in a timely manner. The responsibility of carrying out the
above-mentioned process may be charged to the cognizant industry
organization as stated in Section I.3.0 of this supplement.
I.2.3 Compliance With Regulations
1. Section 2.3.3 of Part I, Revision of Plant Licensing Bases, states that,
"a USI A-46 licensee, in accordance with 10 CFR . 50.59, may revise the
plant licensing bases to reflect that the USI A-46 (GIP) methodology may
henceforth be used as the methodology for verifying the seismic adequacy
of mechanical and electrical equipment within the scope of equipment
covered by the GIP...." The staff recognizes that a licensee may revise
its licensing basis in accordance with 10 CFR 50.59 to reflect the
acceptability of the USI A-46 (GIP) methodology for verifying the seismic
adequacy of electrical and mechanical equipment covered by the GIP. The
staff's approval of the implementation of the GIP does not relieve the
licensees from the requirement to address all aspects of unreviewed
safety questions as specified in 10 CFR 50.59 (for example, those plants
where the FSAR has specified damping values which differ from the GIP.)
The staff understands the word "henceforth" to mean, based on SQUG GIP-0
(page 5 of Part I), "after issuance of a final, plant-specific SER
resolving USI-A-46." If this is not the case, the staff requests
licensees intending to change their licensing bases prior to receipt of
the plant-specific SER to inform the staff in their 120-day response
letters.
2. In Section 2.3.3 of Part I, Example 2 and Example 4 may imply that the
seismic requirements (RG 1.100, Revision 1) for RG 1.97 instrumentation
may be changed to the GIP seismic methodology under 10 CFR 50.59. The
staff has stated, and the SQUG has previously acknowledged, that any
previous commitments, such as for RG 1.97 and TMI Action Plan Item
II.F.2, are not superseded by the resolution methods of the GIP. For
Category 1 equipment, as described in RG 1.97, the staff agrees that the
seismic qualification requirements (RG 1.100, Revision 1) will resolve
the USI A-46 requirements for that equipment. The Category 2 and
Category 3 equipment as described in RG 1.97 have no specific seismic
.
qualification provisions. Therefore, if that equipment is used as part
of the USI A-46 safe-shutdown equipment, it will need to be verified for
seismic adequacy using GIP-2 methods or by an acceptable seismic
qualification method.
3. Section 2.3.3 of Part I is acceptable to the staff subject to the
addition of the following phrase to the last sentence of Example 5: "...
for matters related to verifying the seismic adequacy of electrical and
mechanical equipment."
4. Section 2.3.4 of Part I describes the criteria and procedures for future
modification and for new and replacement equipment. The staff position
is that these criteria and procedures may be applied to new and
replacement equipment on a case-by-case (i.e., plant-specific and
equipment-specific) basis only and with the provisions that the seismic
evaluations are performed in a systematic and controlled manner so as to
ensure that new or replacement items of equipment are properly
represented in the earthquake experience or generic testing equipment
classes, and that applicable caveats are met. In particular, each new or
replacement item of equipment and parts must be evaluated for any design
changes that could reduce its seismic capacity from that reflected by the
earthquake experience or generic testing equipment classes, and these
evaluations must be documented. These criteria and procedures as
described are acceptable for verifying the seismic adequacy of
commercial-grade equipment to be dedicated for safety-related purposes;
but, for other (non-seismic) critical characteristics of equipment to be
dedicated, licensees are referred to such applicable guidance and
requirements as GL 89-02, GL 89-09, and GL 91-05, which include
applicable criteria of 10 CFR Part 50, Appendix B.
The staff normally would require that new or replacement equipment be
qualified in accordance with plant-specific licensing commitments or
current criteria (e.g., 10 CFR 50.49) unless the licensee can justify the
use of other acceptable qualification methods. As a result of the
backfit analysis for the USI A-46 program, the staff determined that the
use of USI A-46 approach provides adequate level of safety and that it
was not cost-justifiable for the safety benefit gained to demonstrate the
seismic qualification of equipment in these older operating plants by
using rigorous current qualification requirements. Therefore, the
resolution as described in GL 87-02 and NUREG-1211, "Regulatory Analysis
for Resolution of Unresolved Safety Issue A-46, 'Seismic Qualification of
Equipment in Operating Plants'," was that the criteria and procedures
described herein are determined to be an acceptable evaluation method for
verifying the seismic adequacy of the equipment in USI A-46 plants
including future modifications and replacement equipment in these plants.
The backfit analysis described in NUREG-1211 did not specifically address
new equipment. However, the staff agrees that it is impractical and
inconsistent with the USI A-46 philosophy to require that new equipment
shall meet current seismic qualification requirements, whereas the
seismic adequacy of all other safe shutdown equipment (which will
.
presumably encompass the large majority of all safe shutdown equipment in
the plant) is verified through the USI A-46 procedures. Therefore, the
criteria and procedures described herein are determined to be an
acceptable evaluation method for verifying the seismic adequacy of new
equipment in USI A-46 plants.
I.3.0 Revisions to the GIP
Section 3.0 of Part I mentions that the earthquake experience or generic
testing equipment classes will be periodically modified by a cognizant
industry organization as new information becomes available. Although the
staff does not intend to review every detail of the information collected,
the suggested cognizant industry organization should submit, for NRC staff
review and approval, a procedure for evaluating the acceptability of new
data and a procedure for updating and revising GIP-2 and subsequent
revisions, based on new information including the lessons learned during the
USI A-46 plant walkdowns.
II GENERIC PROCEDURE FOR PLANT-SPECIFIC IMPLEMENTATION
Part II of GIP-2 which provides the implementation guidelines for the USI
A-46 program, contains 10 sections and 7 appendices. These sections and
appendices are given below.
II.1 Introduction
Section 1 of Part II describes the purpose, background and approach used in
GIP-2. This section also introduces other sections and discusses to some
extent the following subjects:
seismic evaluation personnel
identification of safe shutdown equipment
screening verification and walkdown
outlier identification and resolution
relay functionality review
tanks and heat exchangers review
cable and conduit raceway review
documentation
II.2 Seismic Evaluation Personnel
Discussion
Section 2 of Part II defines the responsibilities and qualifications of the
engineers who will perform seismic evaluations of the equipment. The
systems engineers will develop the list of equipment required for safe
shutdown. The systems engineer should be a degree engineer, or equivalent,
and should have had extensive experience with, and broad understanding of,
the systems, equipment, and procedures of the plant. The seismic capability
engineers will conduct the walkdowns and assess the seismic adequacy of
safe-shutdown
.
equipment. The seismic capability engineers should be degree engineers, or
equivalent, who have completed a SQUG-developed training course on seismic
adequacy verification of nuclear power plant equipment. These engineers
should have at least 5 years of experience in earthquake engineering
applicable to nuclear power plants and in structural or mechanical
engineering. At least one of the seismic capability engineers on each of
the seismic review teams should be a licensed professional engineer to
ensure that there is a measure of accountability and personal responsibility
in making the equipment seismic adequacy determination. The relay reviewers
will perform the functionality review of the relays with the safe-shutdown
functions. The lead relay reviewer should be a degree, or equivalent,
electrical engineer with some electrical engineering experience who is
familiar with the relay functionality review procedure described in Section
6 of Part II and Reference 8 of GIP-2. The lead relay reviewer should
successfully complete the SQUG-developed relay training course. The plant
operations staff will review the safe-shutdown equipment list and assist the
seismic capability engineers and the relay review team. The plant
operations personnel should have experience in the specific plant being
seismically verified.
Evaluation and Conclusion
Based on the above discussions, the staff finds the criteria for qualifying
those individuals responsible for implementing the GIP-2 procedure
acceptable in that the required qualifications are adequate to assure that
the GIP-2 is performed in an acceptable fashion.
The staff acknowledges that these responsible individuals must exercise
judgment to implement the USI A-46 program. The review engineers should
utilize the technical information in the GIP-2 and the reference documents
to the maximum extent practicable in determining the seismic adequacy of
equipment. Where judgements are needed to make these determinations, the
assumptions and basis for the judgmental conclusions should be documented as
required in GIP-2 or identified in this supplement.
II.3 Identification of Safe-Shutdown Equipment
Discussion
Section 3 of Part II describes the overall method for identifying the
equipment needed to achieve and maintain safe-shutdown conditions in a
nuclear plant. The SQUG commitments, general criteria and governing
assumptions, scope of equipment, safe-shutdown functions, safe-shutdown
alternatives, identification of equipment, operations department review, and
documentation are the major subjects discussed in this section. Loss of
offsite power as a result of SSE is assumed, and the systems selected for
the safe shutdown should not be dependent on a single piece of equipment
whose failure would preclude a safe shutdown. Based on these assumptions
and others as specified in GIP-2, the licensee will use the following
two-stage approach to identify the equipment needed to achieve and maintain
a safe-shutdown condition:
.
1. The licensee will select a safe-shutdown path which would ensure that the
four essential safe-shutdown functions listed below can be accomplished
following an SSE. The functions are:
reactor reactivity control
reactor coolant pressure control
reactor coolant inventory control
decay heat removal
2. After identifying the safe-shutdown path, the licensee will identify the
individual items of equipment required to accomplish the four essential
safe-shutdown functions.
Evaluation and Conclusion
The staff finds the proposed two-stage approach adequate for identifying the
equipment needed to achieve and maintain a safe-shutdown condition.
Therefore, the staff concludes that Section 3 of Part II and Appendix A of
GIP-2 are acceptable subject to the following:
1. Regarding the safe-shutdown equipment list (SSEL), the "safe shutdown" is
defined as bringing the plant to, and maintaining it in, a hot shutdown
condition during the first 72 hours following an SSE (i.e. within
72 hours, the plant is cooled down to the "hot shutdown" condition in
accordance with the plant-specific Technical Specifications). The intent
of this position is to have pressurized-water reactors (PWRs) lower their
temperature and pressure within 72 hours to the point at which residual
heat removal (RHR) equipment could be used, but does not necessarily
require RHR equipment to be included on the SSEL. The staff does not
intend to require plants to cool down faster than their original design
capability or technical specification limits. Therefore, if a licensee
cannot achieve hot shutdown at a plant within 72 hours, the licensee
should discuss with, and obtain prior written consent from the staff on a
case-by-case basis before implementing the USI A-46 program.
2. All facilities have Emergency Operating Procedures (EOPs) which address
actions in the event of an accident. As stated in Section 3.2.8 of GIP-
2, the staff expects that plant operators should be trained in the use of
existing normal shutdown procedures or symptom-based EOPs which would be
used if a safe shutdown earthquake were to occur. The compatibility of
these procedures with the USI A-46 safe shutdown equipment list should be
verified by the plant Operations Department, and the results included in
the operator training program. This will ensure that the shutdown path
selected for USI A-46 (and equipment included in the SSEL) is a
legitimate safe-shutdown path consistent with plant procedures and
operator training.
3. Because some components, such as those made of cast iron, are brittle and
are more vulnerable to earthquake damage, any such equipment identified
during the walkdown shall be specifically evaluated for seismic adequacy.
.
4. With regard to Section 3.3.2 of Part II "Exclusion of NSSS Equipment,"
the staff finds that the technical basis provided in Reference 17 of
GIP-2 is acceptable for excluding those items of equipment listed in
Section 3.3.2 with the exception of safety-relief valves. The NSSS
equipment exclusion given in Section 3.3.2 of GIP-2 does not apply to
safety-relief valves included in the USI A-46 scope because Reference 17
of GIP-2 does not provide a basis for excluding the safety-relief valves
from the USI A-46 scope.
5. Section 3.3.3 of Part II requires that any equipment needed for safe
shutdown be evaluated for relay chatter. For example, even if equipment
such as a pump is itself seismically rugged, the effects of relay chatter
on the electric power and instrumentation and control circuits still need
to be evaluated to ensure the equipment functionality.
II.4 Screening Verification and Walkdown
Discussion
Section 4 of Part II describes the screening verification and walkdown
procedures that will be implemented to verify the seismic adequacy of the
equipment. In summary, the licensee should (1) compare the seismic capacity
with the demand, (2) satisfy the caveats of the respective databases, (3)
check the anchorages for adequacy, and (4) consider the seismic
interactions.
Evaluation
Section 4 of Part II provides the first level of screening of the equipment
required for safe shutdown for its seismic adequacy. GIP-2 also provides
criteria and screening procedures for five types of anchorages which have
been used extensively in the nuclear power plants to secure equipment. The
criteria provide guidance for determining the seismic load acting on, and
the allowable load of, individual anchors to be calculated and compared.
Anchors will be classified as outliers if the loads acting on the anchors
exceed their allowable parameters. Some anchors could be identified as
outliers during visual inspection of the screening procedures. The
evaluation of screening verification and walkdown follows in Sections
II.4.2, II.4.3, and II.4.4 of this supplement.
Conclusion
The staff has reviewed the screening procedures and criteria. Based on the
evaluations and findings described in Sections II.4.2, II.4.3, and II.4.4
below, the staff concludes that the screening procedures and criteria are
adequate and acceptable only for verifying seismic adequacy of equipment in
USI A-46 plants, subject to the staff clarifications, interpretations,
exceptions and positions described in the sections that follow.
II.4.0 Introduction
This section provides a summary and organization of Section 4 of Part II of
GIP-2. The staff has no comment on this section.
.
II.4.1 SQUG Commitments
Section 4.1 of Part II provides SQUG general commitments in the areas of
screening verification and walkdown procedures. The SQUG commitments and
the implementation guidance of GIP-2 were developed to form an integral part
to satisfy the guidance of GL 87-02. Therefore, the staff position is that
the licensee must commit to both the SQUG commitments and the use of entire
implementation guidance provided in GIP-2, unless otherwise justified to the
staff as described in GIP-2 and this supplement (see Item 2 in Section I.1.2
and Item 1 in Section I.1.3 of this supplement).
II.4.2 Seismic Capacity Compared to Seismic Demand
1. Section 4.2 of Part II maintains that "... the seismic capacity spectrum
needs only to envelop the seismic demand spectrum for frequencies at and
above the conservatively estimated lowest natural frequency of the item
of equipment being evaluated..." (page 4-10 of GIP-2). The NRC staff
cautions that because an equipment assembly (e.g., electrical cabinet
lineup) may consist of many subassemblies, each manifesting its
fundamental mode of vibration at different frequencies, the GIP-2
approach may be non-conservative unless all such frequencies are
determined with high confidence. In addition, unless the equipment is
tested with a high-level vibratory input, the fundamental frequency is
extremely difficult to estimate, especially for complex structured
equipment. Therefore, the staff position is that the capacity spectrum
should envelope the demand spectrum over the entire frequency range
unless the lowest natural frequency of the equipment, the door panels, or
internal structures and components as described in Section 4.2 of GIP-2
can be conservatively established (see Item 3 of Section III.7 of this
supplement).
2. In regards to the comparison of seismic capacity to demand in methods
A.1, A.2, (page 4-14 of GIP-2), and B.1, B.2 (page 4-17 of GIP-2), the
SQUG proposes in GIP-2 to use 5% damped seismic demand spectra for
comparison with the corresponding seismic capacity spectra. The staff
has examined the damping values listed in the licensing basis documents
of the USI A-46 plants. Several of these plants have been licensed with
equipment damping values of 2% or less. However, the majority of the
plants in this group do not have a clear definition of the damping values
for equipment in their safety analysis reports (SARs). The seismic
capacity spectra of the equipment in the seismic experience database were
established at 5% damping. Although in the amplified range at discrete
frequencies the seismic demand spectrum for equipment at 2% is higher
than that at 5%, the seismic capacity spectrum at 2% is also higher than
that at 5%. The difference in magnitude (in the ordinates) between the
seismic demand spectra at 2% and 5% is comparable to the difference in
magnitude between the seismic capacity spectra at 2% and 5%. For the
purpose of comparison, insofar as verifying seismic adequacy of equipment
using the GIP-2 methodology, it is the judgement of the staff that the
damping level at which the seismic demand spectra are established is of
little significance (for the range of damping values discussed herein,
i.e. approximately 1-5%) provided that the
.
corresponding capacity spectra are established at the same damping
levels. Therefore, the staff finds that the use of seismic demand
spectra for comparison at 5% damping is acceptable for all USI A-46
plants for the purpose of verifying the seismic adequacy of equipment.
3. With respect to the "Definition of Terms" in the last paragraph of page
4-18 of GIP-2, the staff positions on the definition of "conservative,
design" in-structure response spectra are as follows: "Conservative, de-
sign" in-structure response spectra are defined as in-structure response
spectra that have been computed in accordance with current NRC regulatory
guidelines (such as RG 1.60 and RG 1.61) and the Standard Review Plan
(SRP Section 3.7, Rev. 2, August 1989). Alternatively, for post-1976
operating license (OL) plants with non-Housner-type ground response
spectra (Category 1 plants without double asterisks , Table A) and plants
included in the Systematic Evaluation Program (SEP, Category 2, Table A),
the in-structure response spectra included in the licensing-basis (LB)
documents such as final safety analysis reports (FSARs), updated safety
analysis reports (USARs), and other pertinent commitments related to
in-structure response spectra may be used as "conservative, design"
in-structure response spectra. For plants in neither category
(Category 1 plants with double asterisks and Category 3, Table A), the
plant LB in-structure response spectra may be used, provided that the
licensee submits as part of its 120-day response package the detailed
information on which procedures and criteria were used to generate those
in-structure response spectra (see item 5, Section 2.2.1 of Part I of
GIP-2). The staff will review the acceptability of the proposed usage
case-by-case. The staff approval of the proposed in-structure response
spectra is necessary before the commencement of the implementation
program.
As stated in Section 2.2.1 of Part I of GIP-2, each licensee shall submit
its schedule for implementing the resolution of USI A-46 within 120 days
after this supplement is issued. The plant-specific implementation
schedule shall be such that the affected plant should complete its
implementation within 3 years after the issuance of this supplement. For
Category 1 plants with double asterisks and Category 3 plants, however,
the 3-year period will not commence until one of the following conditions
is met:
(1) the receipt of staff approval of the in-structure
response spectra to be used to resolve the USI A-46.
(2) 60 days following the licensee's initial submittal of acceptable
procedures and criteria in generating those in-structure
response spectra.
.
Table A
USI A-46 plants categorized according to
In-structure Response Spectra*
Category 1 Category 2 Category 3
Post-1976 OL SEP plants Pre-1976 OL plants
plants
(15 units) (9 units) (40 units)
Arkansas 2 Palisades Robinson 2
**Crystal River 3 Ginna Point Beach 1/2
**St. Lucie 1 Oyster Creek Monticello
Hatch 2 Dresden 2 Dresden 3
**Calvert Cliffs 2 Millstone Unit 1 Pilgrim 1
**Cook 2 ***Yankee Rowe Quad Cities 1/2
**Salem 1/2 Haddam Neck Surry 1/2
Brunswick 1 Big Rock Point Turkey Point 3/4
Davis-Besse 1 San Onofre 1 Oconee 1/2/3
Beaver Valley 1 Brunswick 2
North Anna 1/2 Trojan
**Browns Ferry 3 Millstone 2
Farley 1 Nine Mile Point 1
Peach Bottom 2/3
Prairie Island 1/2
Duane Arnold
Cooper
Arkansas 1
Calvert Cliffs 1
Cook 1
Hatch 1
FitzPatrick
Three Mile Island 1
Vermont Yankee
Kewaunee
Fort Calhoun
Zion 1/2
Browns Ferry 1/2
Indian Point 2/3
* All plants in this table are the USI A-46 plants. All plants except
St. Lucie 1 and Turkey Points 3/4 are SQUG members. In case more than
one set of in-structure response spectra appear in the LB documents,
use the more conservative set of spectra or justify the use of the
others.
** Category 1 Plants with Housner-type ground response spectra.
*** Yankee Rowe is no longer an operating reactor.
.
4. Regarding the scaling of the Individual Plant Examination of External
Events (IPEEE) spectra for application in the USI A-46 program (page 4-19
of GIP-2), the staff's position is as follows:
The in-structure response spectra for some USI A-46 plants may have been
or may be developed for the IPEEE based on the realistic, median-center
method as described in Section 4.2.4 of Part II of GIP-2. This method
uses the NUREG/CR-0098, "Development of Criteria for Seismic Review of
Selected Nuclear Power Plants," 1978, median rock or soil spectrum
(depending on the primary condition at the site) anchored at the assigned
review level earthquake. For these plants, the IPEEE in-structure
response spectra may be used to generate realistic, median-centered
in-structure response spectra for use in the USI A-46 program by
appropriately scaling down the IPEEE spectra.
If this approach is to be used to resolve USI A-46, the licensee should
submit as part of its 120-day response package the procedure and the
criteria to be used to generate those realistic, median-centered in-
structure response spectra if different than those specified in GIP-2
(see Item 5, Section 2.2.1 of Part I of GIP-2). But, when technical
justifications identified on page 4-19, Item d and the last sentence of
Page 4-20 of GIP-2 need to be developed, then such justifications should
be provided to the NRC in the 120-day response package.
This staff position is intended, for USI A-46, to allow the licensee to
use seismic input for the IPEEE as described in NUREG-1407, "Procedural
and Submittal Guidance for the Individual Plant Examination of External
Events (IPEEE) for Severe Accident Vulnerabilities: Final Report," June
1991.
II.4.3 Equipment Class Similarity and Caveats
The staff interprets Section 4.3 of Part II (as well as Sections 4.1.3 and
4.2.2) regarding the use of caveats to mean that the review engineer will
determine whether the equipment meets both the caveats and their intent and
will report accordingly (i.e., via the Seismic Evaluation Report to be
submitted to the NRC and in Appendix G of the GIP).
II.4.4 Anchorage Adequacy
Regarding anchorage guidelines, GIP-2 provides criteria and screening
procedures for five types of anchorages that are used to secure an item or
equipment: (1) expansion anchors, (2) cast-in-place bolts and headed studs,
(3) cast-in-place J-bolts, (4) grouted-in-place bolts, and (5) welds to
embedded or exposed steel. GIP-2 classifies any other types of anchorage as
outliers.
1. Expansion Anchors
For expansion anchors, GIP-2 provides nominal allowable pullout and shear
loads for various diameters of single anchors for certain concrete
strengths with specified minimum embedments, minimum spacings between
.
anchors, and minimum distances of anchors to a free concrete edge. Also
provided in GIP-2 are load reduction factors for specific types of
expansion anchors by different manufacturers. GIP-2 requires a tightness
check on the anchor head or nut to detect gross installation defects.
Acceptance criteria are provided to assure a 95-percent confidence level
that there are no more than 5-percent anchors which do not meet the
tightness check, as described in GIP-2, for expansion anchors. GIP-2
also requires a check on the anchor projection above concrete to ensure a
minimum anchor embedment in the concrete. Furthermore, checks are
required on the spacing between anchors, the distances from anchors to a
free concrete edge, concrete strength, and concrete cracking conditions,
and reduction factors for nominal allowable loads are specified in the
GIP for each condition which does not meet the minimum requirements for
anchors having nominal allowable loads. On the basis of this
information, the actual allowable load for each anchor can be calculated.
The nominal allowable pullout and shear loads for single anchors in GIP-2
were obtained by dividing the average ultimate loads of test expansion
anchors by a minimum safety factor of three. The reduction factors
specified in GIP-2 for different manufacturers, for closely spaced
anchors, for less edge distance than the specified minimum, for less
concrete strength than that of anchors having nominal loads, and for
cracked concrete were also obtained from test data. The staff concludes
that a minimum safety factor of three in conjunction with appropriate
reduction factors for other conditions as specified in GIP-2 gives
adequate safety margins for allowable loads of expansion anchors.
Furthermore, the staff concludes that the safety margin of expansion
anchors is enhanced by the GIP-2 requirements of 100-percent visual
inspection of accessible anchors and sample tightness checks of expansion
anchors.
2. Cast-in-Place Bolts and Headed Studs
As with expansion anchors, GIP-2 provides nominal allowable pullout and
shear loads for various diameters of single bolts and studs for certain
concrete strengths with specified minimum embedment, minimum spacings
between bolts or studs, and minimum distances of bolts or studs to a free
concrete edge. GIP-2 requires a check on the actual embedment, spacing
between bolts and studs, distances of bolts or studs to a free concrete
edge, concrete strength, and concrete cracking conditions for
cast-in-place bolts and headed studs, and specifies reduction factors for
allowable loads for each condition which does not meet the minimum
requirements for bolts and studs having nominal allowable loads. On this
basis, the actual allowable load for each bolt or stud can be calculated.
The nominal allowable shear loads for single bolts are based on the
nominal bolt area times allowable shear stress of 17 ksi. The staff
compared the allowable shear loads of single bolts specified in GIP-2
.
with test data, and found that safety factors with respect to ultimate
failure loads are greater than three, which is adequate. The nominal
allowable pullout loads for single bolts or studs are based on the
nominal bolt area times allowable tensile stress of 34 ksi, and have a
safety factor of two with respect to a 45-degree concrete cone failure
mechanism. For anchorages with multiple bolts or studs, a minimum safety
factor of one and one-half is provided against a 45-degree concrete cone
failure. The 45-degree failure cone was also assumed for the effects of
bolts spaced close to each other or located close to the concrete free
edge. The 45-degree failure cone shape is hypothetical, and predicts a
lower pullout load for bolts with shallow embedment and a higher pullout
load for bolts with deep embedment than that of test bolts. The staff
has verified that, even for the deepest embedded bolt presented in GIP-2,
the actual safety factor provided by GIP-2 is still slightly greater than
one and, therefore, the staff accepts the allowable loads as specified in
GIP-2. The reduction factor given in GIP-2 for cracked concrete was
based on test data and, thus, is also acceptable to the staff.
3. Cast-in-Place J-Bolts
A J-bolt is a plain steel bar that has a hook (usually in a 90- or 180-
degree form) at the end which is embedded in concrete, and is threaded
with a nut at the other end. GIP-2 provides nominal allowable pullout
and shear loads for various diameters of single J-bolts for concrete
strength equal to or greater than 3.5 ksi with specified minimum
embedments, minimum spacings between bolts, and minimum distances of
bolts to a free concrete edge. GIP-2 specifies reduction factors for
bolts that are embedded less than the specified minimum, that are located
closer to a free concrete edge than the specified minimum, and that are
embedded in concrete with strength less than 3.5 ksi. GIP-2 requires
that J-bolts be classified as outliers if the bolts are spaced less than
three times the bolt diameter, or if they are embedded in certain cracked
concrete.
The nominal allowable pullout and shear loads for single J-bolts are
identical to that of cast-in-place bolts and headed studs. Since the J-
bolts are embedded in concrete much deeper than the cast-in-place bolts
or headed studs, the J-bolts can only fail either in the steel material
or if the J-bolt is pulled out upon failure of the concrete bond. The
specified minimum embedments for J-bolts provide a safety factor of about
two with respect to concrete bond failure, which is acceptable to the
staff. The allowable shear loads have safety factors greater than three
with respect to ultimate shear failure loads, which is also acceptable to
the staff. The reduction factor in GIP-2 for pullout is in proportion to
the reduction in the straight portion of J-bolt embedment. This is
reasonable because bond force from concrete to bolts is proportional to
the embedment length and, therefore, is acceptable to the staff. The
reduction factor in GIP-2 for pullout and shear loads due to concrete
strength less than 3.5 ksi is proportional to the square root of the
ratio of the actual strength to the nominal strength of 3.5 ksi. This is
also reasonable because this reduction represents concrete
.
tensile strength reduction, and thus reduces the holding power of bolts.
Therefore, the staff concludes that the use of appropriate safety factors
for single J-bolts in conjunction with appropriate reduction factors
applied to various conditions as specified in GIP-2, would provide
adequate safety margins for allowable loads of cast-in-place J-bolts.
4. Grouted-in-Place Bolts
GIP-2 provides nominal allowable pullout and shear loads for various
diameters of single, grouted-in-place bolts for concrete strength equal
to or greater than 3.5 ksi with specified minimum spacings between bolts,
and minimum distances of bolts to a free concrete edge. GIP-2 requires a
check on the actual embedment, spacing between bolts, distance of bolts
to a free concrete edge, concrete strength, and concrete cracking
conditions for grouted-in-place bolts, and specified reduction factors
for allowable loads for each condition that does not meet the minimum
requirements for bolts having nominal allowable loads. On this basis,
the actual allowable load for each bolt can be calculated.
The provisions for grouted-in-place bolts in GIP-2 are identical to
provisions for cast-in-place bolts and headed studs if bolts were found
to be installed using certain installation procedures. However, if such
installation procedures cannot be verified to have been used, GIP-2
reduces the nominal allowable pullout loads to one-tenth of that of
cast-in-place bolts and headed studs with other provisions remaining
unchanged.
Test results have indicated that grouted-in-place-bolts, installed
properly, can develop the same allowable loads as cast-in-place bolts.
However, test results also show that pullout loads of grouted-in-place-
bolts drop substantially if the bolts were not installed properly. The
staff believes that the use of 10 percent of the allowable loads (as
specified in GIP-2) of properly installed grouted-in-place-bolts for the
bolts for which proper installation procedures cannot be verified is
conservative. The staff also believes that the allowable shear loads and
other phenomena of grouted-in-place bolts should be similar to that of
cast-in-place bolts and headed studs. Therefore, the staff concludes
that the provisions in GIP-2 for grouted-in-place bolts are adequate.
5. Welds to Embedded or Exposed Steel
GIP-2 provides allowable loads for welds of various sizes, and requires
an inspection of the weld size and quality. The minimum effective length
of fillet welds should not be less than four times the nominal size of
the weld, or else the size of the weld should be considered not to exceed
one-fourth of its effective length. The allowable loads are based on the
weld size times an allowable weld stress of 30.6 ksi. The staff
concludes that the allowable loads so determined for such weld
calculations are conservative and provide adequate safety margins against
failure for welds to embedded or exposed steel.
.
6. Determination of Seismic Load for Individual Anchor
GIP-2 states that the seismic load on anchorages can be calculated by
assuming an equivalent static load acting on the center of gravity of the
equipment, with the load being equal to the input seismic accelerations
times the mass of the equipment. GIP-2 further states that the seismic
accelerations can be obtained from any one of the following three types
of response spectra: (1) a "conservative, design" horizontal,
in-structure response spectrum for SSE as defined in GIP-2, and modified
by Item 3 of Section II.4.2 of this supplement with no modification
factor, (2) a median-centered, horizontal, in-structure response spectrum
for SSE as defined in GIP-2, and (3) a 1.5 times SSE horizontal ground
response spectrum for equipment mounted 40 feet above the effective grade
and having its lowest natural frequency greater than 8 Hz. If option (2)
or (3) is selected, the acceleration is increased by a modification
factor of 1.25. The vertical component of acceleration is assumed to be
two-thirds of the horizontal component of acceleration. The
square-root-of-the-sum-of-the-squares (SRSS) method is used to combine
the load components from three directional accelerations. The final load
on each anchor is calculated by adding the combined seismic loads to the
equipment deadweight loads and any other loads on the anchor. The staff
concludes that the procedures specified in the GIP for determining loads
on individual anchors provide adequate safety margin against failure and
are, therefore, acceptable.
7. Modification and Replacement of Expansion Anchors
GIP-2 states the following:
The GIP-2 criteria may be applied to modification or repair of
existing anchorages (e.g., anchor bolts or welds) including one-
for-one component replacements (e.g., replacing bolts in one-for-one
component replacements).
For new installations and newly designed anchorages in modifications
or replacements, the GIP-2 criteria and procedures may also be
applied, except that the factor of safety currently recommended for
new nuclear power plants in determining the allowable anchorage loads
shall be met.
It is generally recommended that if expansion anchors need to be used
for vibrating equipment, then the undercut type of expansion anchors
should be installed.
The staff concurs with these statements because they are practical and
reasonably conservative.
8. Identification and Resolution of Outliers
Anchors are classified as outliers if the loads acting on the anchors
exceed their allowable loads, or if anchors fail to pass certain
screening guidelines specified in GIP-2. GIP-2 requires the licensee to
.
assign qualified persons to the task of outlier resolution. Although
GIP-2 provides recommendations on generic methods for resolving outliers,
it states that the details for resolving outliers are beyond its scope.
GIP-2 further states that the utility is responsible for resolving
outliers using its existing engineering procedures as it would resolve
any other seismic concern. The staff considers the task of outlier
resolution to be plant specific, and agrees that the acceptability of the
outlier resolution should be addressed individually by each licensee.
The staff will, in its plant-specific SER, present its evaluation of the
licensee's proposed method for resolving any outliers identified in the
plant-specific walkdown inspection summary reports.
9. Verification of Anchorage Capacity by Computer Codes
Two computer codes, EBAC code (Reference 7 of GIP-2) and ANCHOR code
(Reference 14 of GIP-2), were developed and referenced in the GIP-2 for
verifying anchorage capacity of equipment in USI A-46 plants.
The EBAC 1.0 and ANCHOR 3.0 anchorage evaluation computer codes use
somewhat different analysis approaches to determine the adequacy of a
given anchorage arrangement and postulated seismic loading. Although
both codes use a seismic equivalent static load approach to evaluate the
equipment anchorages, the application of the equivalent static loads
differs between these two codes. The EBAC code applies the seismic
equivalent static loads in one direction at a time to an anchorage and
then takes the square root of the sum of the squares (SRSS) of the bolt
reactions from the three-directional seismic inputs. The ANCHOR code
allows an SRSS combination of three-directional seismic equivalent static
loads applied simultaneously and then the combined load is applied at the
equipment center of gravity. The staff finds both methodologies for load
application acceptable.
A major difference between these two computer codes was noted in the
selection of the equipment overturning axis (i.e., the neutral axis).
The EBAC code performs a linear elastic analysis in assuming that plane
sections remain plane as the overturning moments are applied. This
assumption leads to a linear distribution of the tension forces on the
anchors due to an overturning moment. No compressive forces in the
concrete are considered. The EBAC code asks the user to input the
equipment overturning axis locations based on perceived equipment base
flexibility. The staff finds that the lack of specificity regarding the
location of the overturning axis could lead to underestimation of anchor
loads. Furthermore, concrete crushing strain limit could be exceeded if
they are not verified by the user. In contrast, the ANCHOR code uses an
approach which assumes an anchorage to exhibit an elastic-perfectly-
plastic behavior for the anchor bolts and the concrete. The ANCHOR code
computes the overturning axis based on the overturning moments being
resisted by tensile forces in the anchors and compressive forces in the
concrete, assuming the base plate to be rigid. The user should consider
the limit for concrete crushing strain (strength) and verify the
applicability of the rigid base plate assumption.
.
The EBAC code provides for a bilinear and an exponential tension/shear
interaction formulation for anchor bolt strength evaluation. The ANCHOR
code allows selection of tension/shear interaction factors to represent a
bilinear interaction formulation. For both computer codes, the selection
of tension/shear interaction formulation must be consistent with those
given in Appendix C of GIP-2.
Therefore, the EBAC code as given in Reference 7 of GIP-2 and the ANCHOR
code as given in Reference 14 of GIP-2 are acceptable provided that the
items of concern discussed above are adequately considered prior to their
applications. Other computer codes may be used for anchorage evaluations
if demonstrated to be acceptable.
Furthermore, for overall anchorage design and analysis, the equipment
anchorage attributes listed in Section 4.4.1 of Part II of GIP-2 and the
concerns described on page 49 of Reference 5 in GIP-2 must also be taken
into consideration.
10. Anchorages in Inaccessible Areas
Regarding the verification of anchorages in inaccessible areas, GIP-2,
states on page 4-28 that "inaccessible anchorages not required for
strength...need not be inspected...." This is, in general, acceptable to
the staff because, if the inaccessible anchorages are not required for
strength, it implies that the structural integrity of the anchorage is
already adequate. However, to ensure the relay functionality, the
licensee should try all practicable means to inspect all the anchorages
of the cabinets having essential relays to avoid impact or excessive
cabinet motion.
11. Minimum Spacing Between Anchors
The sentence "The minimum spacings given in Appendix C are for distances
between adjacent anchors in which the cones of influence just touch each
other at the surface of the concrete...." which starts at the end of
page 4-39 in GIP-2, is incorrect. This is because the values of minimum
spacing given in Appendix C of GIP-2 were directly taken from Volume 1 of
GIP-2 Reference 7, and these values correspond to a 13-percent shear cone
overlapping as stated on page 2-81 of Volume 1 of GIP-2 Reference 7.
Therefore, this quoted statement should be corrected to be consistent
with the statement given in GIP-2 Reference 7.
12. Use of ACI 349
On page 4-32, GIP-2 states that any anchorage other than the five types
of anchorages covered in GIP-2 should be classified as an outlier and be
resolved in accordance with the guidelines described in Section 5 of
GIP-2. Therefore, other types of cast-in-place embedments as stated in
"Check 14 - Embedment Steel and Pads" in Section 4.4.1 of GIP-2 should be
treated as outliers and resolved in accordance with the guidelines
described in Section 5 of GIP-2. The proposed ACI-349 Code method of
resolving these special outliers is beyond the scope of GIP-2 and also,
.
the entire second paragraph except for the first two sentences on page
4-49 of GIP-2 is not acceptable to the staff. The licensee should use
Appendix C to GIP-2 for guidance.
13. Frequency Shifting
On pages 4-10 (fourth paragraph) and 4-56 (2nd from last paragraph), when
an unbroadened seismic demand response spectrum is used for comparison, a
reference or basis should be provided by the licensee for methods of
"frequency shifting" for addressing the uncertainty in natural frequency
of the building structure.
II.5 Outlier Identification and Resolution
Discussion
Section 5 of Part II defines an outlier as an item of equipment that does
not meet the screening guidelines provided in GIP-2. Several generic
methods for resolving outliers are summarized in Section 5 of Part II.
Evaluation
As noted in Section 5.3 of Part II, the details for resolving outliers are
beyond the scope of the GIP. It is the responsibility of the utility to
resolve outliers, using existing procedures (e.g., plant-specific procedural
controls and QA requirements) as it would resolve any other seismic
concerns. Therefore, the methods and results of outlier resolutions will be
treated on a plant-specific basis.
It should be noted that one of the methods suggested in GIP-2 for resolving
outliers is to use the earthquake experience data documented in References 4
and 5 of GIP-2. However, GIP-2 Reference 4 has been modified extensively by
the addition of unreviewed new information. Although the staff and GIP-2
Reference 5 (SSRAP report) have used the information contained in a previous
draft version (dated February 1987) of GIP-2 Reference 4 to assist the staff
in arriving at a decision for resolution of some technical issues, the staff
has not reviewed this GIP-2 Reference 4, and the GIP-2 Reference 5 does not
endorse the entire GIP-2 Reference 4 (see Reference 9). Therefore, any
specific application of the detailed information documented in GIP-2
Reference 4 for the implementation of USI A-46 resolution should be
submitted to the NRC staff for review and approval before it is used.
Regarding the evaluation of the acceptability of new data, the staff
position is described in Section I.3.0 of this supplement.
Conclusion
Subject to the above clarifications, the staff concludes that the procedures
for outlier identification and the general approach for outlier resolution
are adequate and acceptable.
.
II.6 Relay Functionality Review
Discussion
Section 6 of Part II provides an overview of the relay evaluation procedure
and describes the relationships between other GIP activities and the relay
evaluation which is contained in a separate reference document, "Procedure
for Evaluating Nuclear Power Relay Seismic Functionality," Reference 8 of
GIP-2.
This section was revised from GIP-0, primarily to include a multilevel
screening approach for comparing relay seismic capacity to demand. The
evaluation procedure described in GIP-2 is a summary; the details of the
method are contained in the above-referenced SQUG relay procedures.
Evaluation and Conclusion
The relay review requires the use of the generic equipment ruggedness
spectra (GERS) to assess the relay ruggedness. The staff had concerns about
the amount of relay data that were available in the GERS (open issue C.2.1
of the original GSER). Additional information has been considered in the
GERS and data will be added as needed if the walkdowns identify relays not
currently addressed. The staff considers the SQUG approach practical and,
therefore, acceptable and this issue resolved.
The GERS were constructed using test data from relays of vintages newer than
those that are currently installed in the USI A-46 plants. In the GSER,
open issues C.2.2 and E.2.2 described the concern that the testing of newer
equipment may not be applicable to the older equipment. The SQUG initiated
a program to test a sample of older relays which were of the same type as
those covered by the GERS, and compared the results to the more recent test
results. The test results demonstrated that the difference in seismic
ruggedness between relays of different vintages was not significant. GIP-2
considers this issue resolved, and the staff concurs. If additional testing
in the future, by the NRC, SQUG, or others, provides evidence to change this
conclusion, the staff will take appropriate action at that time.
Open issue E.2.5 of the GSER discussed the inclusion of relay mountings in
the walkdown inspection and the number of relays to be inspected. The staff
concurs with the SQUG position to review a sample of the relay mountings to
ascertain that the relays are mounted in conformance with the vendors'
recommendations. If any abnormality exists, the licensee shall increase the
number of samples for inspection.
In conclusion, on the basis of its review of Section 6 of Part II, the staff
agrees with the approach of evaluating systems and electrical circuits to
determine the effect of relay chatter and endorses the review procedure as
given in GIP-2. Therefore, the staff concludes that the procedure, if
properly implemented, is an acceptable method of verifying the seismic
adequacy of relays for the resolution of USI A-46 subject to the following:
.
1. Use of Zero Period Acceleration Capacities
Regarding the acceptability of a relay, because of the important effects
of zero period acceleration (ZPA) on relay chatter, the staff position is
that in addition to the comparison of the spectral accelerations, the ZPA
capacities should be compared and shown to be adequate (page 6-18 of
GIP-2.)
2. Development of In-Cabinet Amplification Factors
Section 6 of Part II includes the use of a single number amplification
factor which is applicable to a given class of equipment for what is
defined as Screening Level 2. With this concept, an in-cabinet demand
spectrum is estimated by multiplying the base excitation demand spectrum
by an effective amplification factor that is representative of the given
class of equipment. The result is then compared with device ruggedness
spectra to verify device capability.
The amplification factors for motor-control-center-type cabinets, for
control room benchboards and panels, and for switchgear-type cabinets or
similar panels are presented in Table 6-2 of GIP-2. Because these
amplification factors were determined based on test data and some
empirical parameters specific to a certain type of cabinets or panels,
the staff concludes that these amplification factors are reasonable and
acceptable. However, the use of the 0.6 reduction factor for narrow peak
amplification spectra for other types of cabinets, panels, or enclosures
must be justified by the user and documented using procedures described
in Reference 2 of Section 4 of GIP-2 Reference 33 because this 0.6 factor
is an empirical value derived from specific types of cabinets, panels, or
enclosures.
3. Development of In-Cabinet Response Spectra
The use of the in-cabinet amplification factors is intended for initial
screening purposes. Should the result not produce a positive equipment
seismic verification in a given case, then the next level of screening
presents a more definitive methodology developed by EPRI (GIP-2 Reference
33) for generating in-cabinet response spectra. The staff has reviewed
the procedures described in GIP-2 Reference 33 and the results
specifically applicable to control room benchboards and panels. The
staff finds that the approach includes development of conservative
estimates for a single generic lowest natural frequency and a
corresponding single generic high participation factor for this class of
equipment. Therefore, the staff concludes that GIP-2 Reference 33
constitutes an approximate method of generating an in-cabinet demand
response spectrum for devices which will be attached to control room
benchboards and panels that are subject to a given site-specific floor
spectrum.
The EPRI methodology (GIP-2 Reference 33) includes a combination of in
situ experimental tests, modal analysis, linear power spectral density
(PSD) response prediction, response spectrum/PSD transformation, and
.
statistical methods in various combinations to generate the final,
generic, elevated-demand spectrum. The staff finds the use of the
computer program GENRS, as documented in Reference 33 of GIP-2 for the
calculation of in-cabinet response spectra, acceptable only for control
room benchboards and panels as defined in GIP-2 Reference 33 because the
parametric values, such as those for natural frequency and the
corresponding participation factor used in the computer code (GENRS),
were derived specifically from the control room benchboards and panels.
Therefore, the use of GENRS should not be extended to other classes of
equipment without the review and approval of the NRC staff.
The EPRI methodology includes direct generation of a PSD from a required
response spectrum (RRS) and vice versa. Because, the current NRC staff
position on this approach is that the direct-generation method can be
considered only case by case, the staff performed some additional
investigation concerning the viability of this approach and its
applicability in the GENRS computer code. The results of the staff
investigation support the viability of the direct-generation method in
general and its application in the GENRS computer code in particular.
II.7 Tanks and Heat Exchangers Review
Discussion
Section 7 of Part II gives guidelines for evaluating the adequacy of tanks
and heat exchangers. The SQUG commitments, evaluation methodology, vertical
tanks, horizontal tanks, outliers, and documentation are the main topics in
this section.
Evaluation
1. Vertical Tanks
The procedure given in Section 7 of Part II of GIP-2 and discussed in the
subsequent paragraphs covers the screening guidelines for flat-bottom
vertical tanks supported on a concrete pad or floor, and anchored to the
pad (or floor) by means of cast-in-place anchor bolts. The screening
guidelines are applicable when the tank dimensions, anchor bolt
configurations, and materials of fabrication are within the range and
assumptions given in Table 7-1 in Part II.
The last paragraph in Section 7 of Part II indicates that the successful
completion of the review described in Section 7 has been accepted by the
NRC as resolving the seismic issues related to these types of tanks for
USI A-40. However, the SQUG commitments in Section 7.1 and evaluation
methodology in Section 7.2 do not address the screening guidelines for
ensuring the adequacy of the foundation structures of the vertical tanks.
As the tank foundation is subjected to higher loads than those determined
using the rigid tank assumption, SRP Section 3.7.3.II.14.i recommends
that the tank foundation be designed to withstand the seismic forces
imposed on it. The SQUG commitments in Section 7.1 are not consistent
with the guidelines in SRP Section 3.7.3.II.14.i. However,
.
the resolution to this issue as discussed in the later part of this
section is acceptable to the staff for tanks in USI A-46 plants.
Because the screening guidelines are to be used for the as-built vertical
tanks, the staff strongly recommends that the input data required in Step
1 of Section 7.3.2 be based on the pertinent as-built drawings and
verification through walkdowns of the condition of the tanks and the
supporting foundations. Steps 2 through 6 provide the guidelines for
determining the seismic demand applied to a specific tank, in terms of
the overturning moment and the shear load. The seismic demand is based
on the response value of the fluid-structure model at the impulsive modal
frequency (Step 4). The calculated frequency is varied by ñ 20 percent
to account for the uncertainties involved in the calculations. The
maximum responses from the applicable ground or floor response spectrum
at 4-percent damping are used to calculate the seismic demand.
Guidelines are provided to account for soil-structure interaction effects
on the frequency and the response. On the basis of its review of
procedures described in Steps 1 through 6 of Section 7.3.2 of Part II as
summarized above, the staff finds that the seismic demand so determined
is adequate and, therefore, concludes that these steps are logical and
acceptable.
Steps 7 through 18 of Section 7.3.2 of Part II provide a method for
computing the overturning moment capacity of the tank. The method
considers the complex interactions between the anchor bolt capacity, the
anchorage connection capacity, and the allowable buckling stress. Steps
19 and 20 require the users to compute the shear load capacity provided
by the weight of the fluid on the base of the tank, and the frictional
resistance between the base of the tank and the foundation surface. The
formula to compute shear load capacity also reduces the fluid weight to
account for the 40 percent of the vertical component of the earthquake.
Steps 21 and 22 require the users to evaluate fluid level against the
slosh height computed for the postulated earthquake. Section 7.3.6 of
Part II requires the users to check the effect of the flexibility of
attached piping.
In reviewing the earlier version of GIP-2 (Reference 6), the staff
identified the following concerns:
a. The SQUG commitments do not require the users to check the adequacy of
the supporting foundation which are likely to be subjected to higher
loads than the original design that was based on a rigid-tank
assumption.
b. The allowable buckling stress criteria provided in Step 11 are not
sufficiently conservative to account for the out-of-roundness of the
tank, local imperfections, material nonlinearities, the secondary
effects due to shearing stresses, and rotation of the shell wall at
the base. Without considering the uncertainties induced by these
inherent characteristics, the seismic adequacy of the tanks cannot be
assured.
.
In order to resolve the concern regarding the adequacy of the tank
foundation, the SQUG proposed to include the evaluation requirements for
ring foundations of the vertical tanks. The SQUG justified the narrow
scope of the requirements by pointing to the experience data regarding
tank foundation failures.
The staff agrees that ring foundations, when subjected to loads higher
than the design loads, are likely to be more susceptible to failure than
other types of foundation such as foundation mats on ground or floors
supporting the tanks. Therefore, the proposed resolution is acceptable
to the staff, and the concern is resolved with the inclusion of
instructions to the users in Section 7.3.7 of Part II to identify ring
foundations as outliers.
In order to resolve the concern regarding allowable buckling stress
capacity, the staff has proposed to reduce the capacity reduction factor
in Step 16 of Section 7.3.2 Part II from 0.9 to 0.72. The SQUG has
adopted the staff recommendation in GIP-2. Therefore, this concern is
resolved.
During the discussion related to the resolution of USI A-40, "Seismic
Design Criteria," the method of analysis of above-ground, flexible,
vertical tanks was identified as a topic requiring technical resolution.
USI A-40 is resolved in Standard Review Plan (SRP) Sections 2.5.2, 3.7.1,
3.7.2, and 3.7.3 (Revision 2, August 1989). The guidelines related to
the seismic analysis of the above-ground vertical tanks are included in
SRP Section 3.7.3.II.14. As part of the resolution of USI A-40, a number
of tanks at nuclear power plant sites are required to have confirmatory
checks to ensure that the safety-related, above-ground, vertical tanks
are adequately designed. Most of the licensees of newer plants have
incorporated the flexible tank design concept in the design of their
above-ground tanks. Some licensees have committed to make confirmatory
checks of their design using the procedures developed by the SQUG under
the resolution of the USI A-46 program. The implementation of criteria
and procedures described in GIP-2, supplemented by the staff evaluations
described in this supplement for large, flat-bottom, cylindrical,
vertical tanks which are needed for safe shutdown and for refueling water
storage in PWRs, is considered an acceptable method for resolving the
seismic issues related to these types of tanks for both USI A-46 and USI
A-40, as it applies to USI A-46 plants.
2. Horizontal Tanks
The screening guidelines provided in Section 7.4 of Part II of GIP-2 are
applicable when a horizontal tank or a heat exchanger shell satisfies the
following criteria:
Its longitudinal axis (axis of symmetry) is horizontal.
It is supported on its curved bottom by steel saddle plates.
.
It is anchored to a stiff foundation having adequate strength to
resist the seismic loads applied to the tank.
All the baseplates under the saddle have slotted anchor-bolt holes in
the longitudinal direction except the one for an end saddle support.
Its layout and dimensions satisfy the range of parameters and
assumptions listed in Table 7-6 of Part II.
Steps 2 through 7 of the screening guidelines described in Section 7.4 of
Part II provide guidelines for evaluating the resistance of the existing
tank in terms of the anchorage capacity. Steps 8 through 10 provide
guidelines for evaluating the seismic demand of the tank anchorage
system. Step 11 provides instructions for evaluating the tank saddle
stresses. The staff finds the screening methodology to evaluate the
seismic adequacy of horizontal tanks consistent with engineering practice
and, therefore, acceptable for existing installations only.
Conclusion
On the basis of its review of Section 7 of Part II of GIP-2, the staff
concludes that the methodology provided for the seismic adequacy evaluation
of the safety-related horizontal and vertical tanks and heat exchangers
existing at the USI A-46 plants is acceptable. However, the criteria for
evaluating tanks and heat exchangers, as defined herein, are not acceptable
for new installations.
II.8 Cable and Conduit Raceway Review
Discussion
Section 8 of Part II of GIP-2 describes the screening guidelines for cable
and conduit raceway review. The screening procedure is based primarily on
earthquake experience data and some shake-table test data. Several types of
raceway configurations and support systems are covered in this section. The
guidelines consist of a set of walkdown guidelines and a set of limited
analytical review guidelines.
The walkdown guidelines provide guidance for the seismic review teams (SRTs)
to: (1) perform direct in-plant screening reviews of raceway systems against
a set of inclusion rules, (2) assess other seismic performance concerns not
covered by the inclusion rules, and (3) select, during the walkdown, 10 to
20 representative, worst-case samples of raceway supports for analytical
review. The systems which are identified to be within the boundaries of the
inclusion rules would be considered to be within the applicability limits of
the experience database. If violations of the inclusion rules are observed,
the SRT should investigate the specific conditions of the cable tray systems
with proper assessment methodology to verify their seismic adequacy.
.
The purpose of the limited analytical review is to ensure that the selected
worst-case, representative samples of the raceway support systems in the
plant are at least as rugged under the required seismic loadings as those in
the earthquake experience and shake-table test databases that performed
well. Section 3.3 of GIP-2 Reference 9 should be used for selecting samples
for the limited analytical review. If these samples do not pass this
limited analytical review, further evaluations should be conducted and the
sample should be expanded as appropriate. The analytical reviews are
primarily based on the back-calculated capacities of raceway supports in the
seismic experience database. They are formulated with the use of static
load coefficients, plastic behavior structural theory, and professional
engineering judgment to ensure that cable tray and conduit supports are
seismically adequate and as rugged as those in the seismic experience
database. The main feature of the reviews is that all supports in the
selected worst-case samples are checked for deadload (DL) vertical capacity
using the working stress criteria given in Part 1 of the American Institute
of Steel Construction (AISC) Specification. All supports in the selected
worst-case samples must pass the DL check, otherwise the supports must be
treated as outliers and disposed of as such. However, isolated cases of a
support not meeting the one DL criterion could be accepted if the raceway
support system has high redundancy; this can be demonstrated by showing that
the adjacent supports are capable of satisfying the walkdown guidelines,
including the inclusion rules and the analytical review guidelines. In
addition to the DL check, all of the cable tray supports in the selected
worst-case samples suspended from overhead must satisfy three times the DL,
otherwise the supports must be treated as outliers. This check is designed
to ensure that the anchorage supporting the cable trays and conduit raceway
in the USI A-46 plants is as strong as those in the experience database in
sustaining the vertical loads.
The raceway hardware becomes an outlier if it does not meet the walkdown
guidelines (inclusion rules and other seismic performance concerns), or the
limited analytical review guidelines. When an outlier is identified,
additional evaluations as described in GIP-2 Reference 9, or alternative
methods, are required to demonstrate seismic adequacy of the raceway
hardware and to resolve the outlier issue. The evaluations and
justifications to be used to resolve the outlier issue should be based on
mechanistic principles and sound engineering judgment and should be
thoroughly documented for NRC staff review.
Evaluation and Conclusion
The staff has reviewed the guidelines proposed by the SQUG for evaluating
the seismic adequacy of cable and conduit raceway systems. The main
objective of the proposed guidelines was to develop a cost-effective means
of verifying the seismic adequacy of raceway supports in USI A-46 plants.
These guidelines were developed on the bases of analytical studies,
shake-table experimental model tests, and assessment of the performance of
cable and conduit support systems in past earthquakes.
The staff considers that the plant walkdown guidelines represent an
acceptable approach for evaluating the seismic adequacy of existing cable
and conduit raceways in USI A-46 plants. Also, the staff agrees that the
proposed
.
analytical procedure is a reasonable approach to ensure that the cable and
conduit raceways and supports in USI A-46 plants, when all the guidelines
are satisfied, are as rugged as those observed in the past earthquake
experience data. Although the proposed guidelines would not require
detailed analyses and, therefore, would not predict the structural response
of the raceway support systems, they should provide the needed rationale to
judge the seismic adequacy of the raceway support systems with a reasonable
factor of safety. Therefore, the staff concludes that the proposed
guidelines for evaluation of seismic adequacy of cable and conduit raceways
and their supports are acceptable subject to the staff evaluations described
in this supplement.
II.9 Documentation
Section 9 of Part II describes the documentation that is to be submitted to
the staff upon completion of the plant-specific review and includes the
documentation available at the plant site for audit. The major document
types are:
safe-shutdown equipment list report
relay evaluation report
seismic evaluation report
completion letter
The staff has reviewed the outlines of each report as given in GIP-2. The
information to be submitted to NRC for review will provide overall results
of the implementation program. Therefore, the staff finds the proposed
plant-specific information to be submitted to the NRC for resolution of USI
A-46 acceptable.
However, GIP-2 recommends documentation (not required to be submitted to the
NRC) of only the results from several evaluations (e.g., Sections 9.3 and
9.4) and not the assumptions and judgments used for the respective
evaluations. The staff recommends documentation of the assumptions and the
judgments as previously mentioned in Section II.2 of this supplement. The
documentation of assumptions and judgments, in addition to the results of
evaluations, will facilitate the reconstruction of relevant basis for the
licensee's evaluations.
II.10 References
Section 10 of Part II contains a list of references that are the source of
information for the criteria and procedures described in GIP-2. During the
course of its review, the staff consulted References 5, 6, 7, 8, 9, 10, 26,
32, and 33, among others, of GIP-2, in order to develop the bases for
accepting the criteria and procedures presented in GIP-2 for implementing
USI A-46 resolutions. Therefore, the evaluations and conclusions presented
in this SSER No. 2 are based on the information provided in each reference
as dated in GIP-2 with the exception of Reference 4 for the reasons stated
below. If any updated references are to be used for the USI A-46 program,
they must be submitted for staff review and approval.
.
As noted in Section II.5 of this supplement, Reference 4 of GIP-2 has been
modified extensively by the addition of unreviewed new information. Because
the staff has not reviewed this particular version of the reference, any
specific application of the detailed information documented in Reference 4
for the implementation of USI A-46 should be submitted to the staff for
review and approval before it is used. For the evaluation of the
acceptability of new data, see the staff position described in Section I.3.0
of this supplement.
III APPENDICES
III.1 Appendix A, Procedure for Identification of Safe-Shutdown Equipment
Appendix A of GIP-2 amplifies the method described in Section 3 of Part II
for identifying safe-shutdown equipment. The staff incorporated its
evaluation of Appendix A into its discussion in Section II.3 of this
supplement.
III.2 Appendix B, Summary of Equipment Class Descriptions and Caveats
Appendix B of GIP-2 incorporates information regarding the seismic
capacities of 20 equipment classes. This information was extracted
principally from GIP-2 Reference 5 and partially from GIP-2 Reference 4 for
earthquake experience data, and from GIP-2 Reference 6 for the test data.
The staff evaluation of this appendix shall be used in conjunction with the
staff evaluations presented in Sections II.4, II.5, and II.6 of this
supplement.
In GIP-2 Reference 5, SSRAP documented its review of GIP-2 Reference 4,
GIP-2 Reference 6, and other supporting documents. After a detailed and
careful review of the full range of the available experience database,
combined with the general experience of the SSRAP members, the SSRAP
concludes that the equipment (20 classes) presented in Appendix B of GIP-2,
when properly anchored, and with some reservations as discussed in GIP-2
Reference 5 and Appendix B of GIP-2, have an inherent seismic ruggedness and
a demonstrated capability to withstand seismic motion bound as specified
without significant structural damage and malfunction. The staff concurs
with this conclusion.
On the basis of the discussion described above and the review of information
presented in Appendix B of GIP-2 and other supporting documents, the staff
concludes that Appendix B is generally acceptable, subject to the following:
1. Throughout Appendix B, such statements as "equipment determined to be
seismically rugged" are repeatedly used. The staff considers such
statements ambiguous unless the appropriate vibration level for which the
equipment is rugged is given. In addition, the first sentence of each
equipment class states that the equipment "has been determined to be
seismically rugged...provided the intent of each of the caveats listed
below is met...." The staff also finds such statements to be incomplete
and misleading because according to Appendix B, a user can simply meet
the caveats and declare the equipment to be rugged (and therefore
acceptable) for its application without even comparing it with the demand
vibration level. The staff takes the position that in addition to
meeting the caveats, the user must demonstrate the demand level is
appropriately satisfied by the capacity level before the
.
equipment can be considered to be rugged and acceptable for its
application.
2. Regarding the attachment weight of 100 pounds, GIP-2 uses the term "a
cabinet assembly" (e.g., page B.1-4, MCC/BS caveat 4; page B.2-3, LVS/BS
caveat 5). The staff understands this term to mean a combination or
lineup of a number of individual cabinets, bays, or frames.
3. GIP-2 includes a caveat for many equipment classes that the sections of
the multibay cabinet should be bolted together only "if any of these
cabinets contain essential relays...." Since the database cabinets in
GIP-2 Reference 6 were bolted during testing, the adjacent cabinets at
the plants should be bolted for applicability of the GERS level, even
though these cabinets do not contain relays. Otherwise, the responsible
review engineer should justify the use of GERS level for those cabinets.
4. The capacity levels for motor operators on valves presented in GIP-2, and
in GIP-2 Reference 6, appear to be high compared to the levels reported
in NUREG/CR-4659, Vol. 4 (Reference 10). Note that new data from this
report were not available to the SQUG at the time GIP-2 was being
prepared.
Therefore, the information presented in Reference 10 should be considered
in the evaluation of motor operators on valves, especially for some
earlier models.
III.3 Appendix C, Anchorage Data
Appendix C of GIP-2 includes information necessary for verifying the
adequacy of anchorage. The staff incorporated its evaluation of this
appendix in Section II.4 of this supplement.
III.4 Appendix D, Seismic Interaction
Appendix D of GIP-2 describes seismic interaction as the physical
interaction of any structures, piping, or equipment with nearby
safe-shutdown equipment caused by relative seismic motions. Three seismic
interaction effects are covered in the GIP, namely: proximity, structural
failure and falling, and flexibility of the attached lines and cables. The
staff finds the guidelines for reviewing seismic interaction adequate and,
therefore, this appendix is acceptable and should be used with the staff
evaluation presented in Section II.4 of this supplement.
III.5 Appendix E, Preparatory Work Prior to Walkdown
Appendix E of GIP-2 describes the experience gained from previous walkdowns
to maximize the effectiveness of the walkdown.
The GIP states that most of the equipment "has been shown to be seismically
rugged...." As explained in Section III.2 above, the staff considers this
statement ambiguous unless the appropriate vibration level is associated
with it.
.
III.6 Appendix F, Screening Walkdown Plan
Appendix F of GIP-2 describes the organization and approach that can be used
by the seismic review team, the degree of inspection to be performed, the
walkdown logistics to be followed, and the screening walkdown to be
completed. The staff finds the screening walkdown plan adequate for
accomplishing the objectives of the walkdown inspection and, therefore, the
staff concludes that the GIP screening walkdown plan is acceptable.
III.7 Appendix G, Screening Evaluation Work Sheets
Appendix G of GIP-2 provides the work sheets to be used by the walkdown team
to document their review. The staff finds these work sheets to be a useful
summarized checklist that briefly documents the responses to essential
screening questions addressing the seismic adequacy of each piece of
equipment and, therefore, the staff concludes that this appendix is
acceptable subject to the following staff positions:
1. Since the screening evaluation work sheets in Appendix G contain summary
information presented in Appendices B and C, in case there is any
conflict between these pieces of information, the information in
Appendices B and C should be used. For example, for motor control
centers, the weight of 800 pounds should be considered maximum instead of
average (pages B.1-7 and G.1-2).
2. The screening evaluation work sheets do not require documentation of
manufacturer, model, etc. For information purposes only, the staff
strongly recommends that such information should be recorded if readily
available.
3. Since GIP-2 allows the demand level to exceed the capacity level under
certain conditions, in response to the question "Does capacity exceed
demand?" on the Screening Evaluation Work Sheet (SEWS) for each piece of
equipment, the reviewer must also identify whether the exceptions
described on page 4-10 of GIP-2 were used in the comparison (see Item 1
of Section II.4.2 of this supplement).
CONCLUSION
The staff concludes that GIP-2, dated February 1992 (Reference 8),
supplemented by the staff positions, clarifications, and interpretations,
stated herein for each section of GIP-2, constitutes an acceptable method
for the implementation of the resolution of USI A-46 as specified in Generic
Letter 87-02, subject to the following exceptions:
1. Section II.3, Identification of Safe-Shutdown Equipment: Evaluation and
Conclusion, Item 4
The NSSS equipment exclusion given in Section 3.3.2 of GIP-2 does not
apply to safety-relief valves included in the scope of USI A-46.
.
2. Section II.4.1, SQUG Commitments
If the licensee commits to use GIP-2 for the implementation of USI A-46,
it must commit to both the SQUG commitments and the use of the entire
implementation guidance provided in GIP-2, unless otherwise justified to
the staff as described in GIP-2 and this supplement.
3. Section II.4.4, Anchorage Adequacy, Item 9, Verification of Anchorage
Capacity by Computer Codes
The EBAC code as given in Reference 7 of GIP-2 and the ANCHOR code as
given in Reference 14 of GIP-2 are acceptable provided that the items of
concern discussed in Item 9 of Section II.4.4 of this SSER No. 2 are
adequately considered prior to their applications.
4. Section II.4.4, Anchorage Adequacy, Item 10, Verification of Anchorages
in Inaccessible Areas
To ensure relay functionality, the licensee should try all practicable
means to inspect all of the anchorages of cabinets having essential
relays.
5. Section II.4.4, Anchorage Adequacy, Item 11, Minimum Spacing Between
Anchors
On page 4-39 of GIP-2, the sentence, "The minimum spacings given in
Appendix C are for distances between adjacent anchors in which the cones
of influence just touch each other at the surface of the concrete..." is
incorrect. The minimum spacing values in GIP-2, Appendix C correspond to
a 13-percent shear cone overlapping. The quoted statement should be
corrected.
6. Section II.4.4, Anchorage Adequacy, Item 12, Use of ACI 349
The use of ACI 349, Appendix B, for the resolution of anchorage outliers
is not acceptable to the staff.
7. Section II.5, Outlier Identification and Resolution, and Section II.10,
References
The staff has not reviewed the current version of GIP-2 Reference 4. Any
specific application of this reference for resolving USI A-46 should be
submitted to the NRC staff for review and approval before it is used.
8. Section II.6, Relay Functionality Review, Item 1, Use of Zero Period
Acceleration Criteria
For the relay functionality review, in addition to the comparison of the
spectral accelerations, the ZPA capacities should be compared and shown
to be adequate.
.
9. Section III.2, Appendix B, Summary of Equipment Class Descriptions and
Caveats, Item 3
GIP-2 includes a caveat for many equipment classes that states that the
adjacent sections of multibay cabinets should be bolted together only "if
any of these cabinets contain essential relays." For multibay cabinets,
all cabinets, regardless of whether or not a cabinet contains essential
relays, should be bolted together in order for the GERS to be applicable,
unless justification can be provided for applying these GERS to multibay
cabinets that are not bolted together.
10. Section III.2, Appendix B, Summary of Equipment Class Descriptions and
Caveats, Item 4
The information presented in Reference 10 of this SSER No. 2 should be
considered in the evaluation of motor operators on valves, especially for
some earlier models.
11. Section III.7, Appendix G, Screening Evaluation Work sheets, Item 1
In any case where the information presented in Appendix G conflicts with
the information in Appendices B and C, the information in Appendices B
and C should be used.
12. Section III.7, Appendix G, Screening Evaluation Work sheets, Item 3
In response to the SEWS question, "Does capacity exceed demand?", the
reviewer must identify whether the exceptions described on page 4-10 of
GIP-2 were used in the comparison.
.
REFERENCES
1. Generic Letter 87-02, "Verification of Seismic Adequacy of Mechanical
and Electrical Equipment in Operating Reactors, Unresolved Safety
Issue (USI) A-46," U.S. Nuclear Regulatory Commission, February 19,
1987.
2. "Generic Implementation Procedure (GIP) for Seismic Verification of
Nuclear Plant Equipment," Revision 0, Seismic Qualification Utility
Group, June 1988.
3. "NRC Generic Safety Evaluation Report on the Seismic Qualification
Utility Group Generic Implementation Procedure, Revision 0, for
Implementation of USI A-46," July 29, 1988.
4. "Generic Implementation Procedure (GIP) for Seismic Verification of
Nuclear Plant Equipment," Revision 1, Seismic Qualification Utility
Group, December 1988.
5. "Supplemental Safety Evaluation Report No. 1 on SQUG Generic
Implementation Procedure, Revision 1," by NRC, June 29, 1990.
6. "Generic Implementation Procedure (GIP) for Seismic Verification of
Nuclear Power Plant Equipment," Revision 2, Seismic Qualification
Utility Group, September 1990.
7. Letter, James A. Norberg, NRC, to Neil P. Smith, dated March 11, 1991,
"Comments on Revision 2 of Generic Implementation Procedure (GIP) for
Use in USI A-46 Programs, dated September 21, 1990."
8. "Generic Implementation Procedure (GIP) for Seismic Verification of
Nuclear Power Plant Equipment," Revision 2, Corrected 2/14/92, Seismic
Qualification Utility Group, February 1992.
9. Letter, W. A. Von Reisemann, SSRAP, to Pei-Ying Chen, NRC, dated July
23, 1991, "EQE Class of Twenty Report."
10. NUREG/CR-4659, Vol. 4, "Seismic Fragility of Nuclear Power Plant
Components (Phase II): A Fragility Handbook on Eighteen Components,"
June 1991.
