Protecting People and the EnvironmentUNITED STATES NUCLEAR REGULATORY COMMISSION
UNITED STATES
NUCLEAR REGULATORY COMMISSION
WASHINGTON D. C. 20555
September 19, 1991
TO: LICENSEES AND APPLICANTS OF THE FOLLOWING PRESSURIZED-WATER
REACTOR NUCLEAR POWER PLANTS:
1. Braidwood Units 1 and 2
2. Byron Units 1 and 2
3. Catawba Units 1 and 2
4. Comanche Peak Units 1 and 2
5. Cook Units 1 and 2
6. Diablo Canyon Units 1 and 2
7. McGuire Units 1 and 2
SUBJECT: REQUEST FOR INFORMATION RELATED TO THE RESOLUTION OF GENERIC
ISSUE 130, "ESSENTIAL SERVICE WATER SYSTEM FAILURES AT MULTI-UNIT SITES,"
PURSUANT TO 10 CFR 50.54(f) - GENERIC LETTER 91-13
DISCUSSION
The purpose of this letter is to inform affected licensees and applicants of
the technical findings resulting from the NRC resolution of Generic Issue
130 (GI-130), "Essential Service Water System Failures at Multi-Unit Sites,"
and to request information from licensees and applicants at affected
multi-unit sites relating to the applicability of certain findings regarding
their facilities. Affected licensees and applicants are required to respond
to the request for information contained in this letter, but no new
requirements or staff positions are imposed on the affected licensees and
applicants by this letter.
The essential service water system (ESWS) is important in maintaining plant
safety during power operation, shutdown, and accident conditions. As part
of our evaluation of loss of essential service water (LOSW), extensive
analyses of this issue were performed at the Brookhaven National Laboratory
(BNL). The technical findings of this effort at BNL are reported in
NUREG/CR-5526, "Analysis of Risk Reduction Measures Applied to Shared
Essential Service Water Systems at Multi-Unit Sites." In addition, the NRC
staff performed a regulatory analysis to evaluate the safety benefits and
implementation costs associated with various equipment and the
administrative-type improvements that were considered. The staff's
regulatory analysis is contained in NUREG-1421, "Regulatory Analysis for the
Resolution of Generic Issue 130: Essential Service Water System Failures at
Multi-Unit Sites." These analyses assume that the flushing and flow testing
provisions of Generic Letter (GL) 89-13, "Service Water System Problems
Affecting Safety-Related Equipment," will be applied to the crosstie lines
as part of addressees' implementation of the resolution of GI-51, "Improving
the Reliability of Open-Cycle Service Water Systems" (GL 89-13 and
Supplement 1). On the basis of results of these evaluations of this generic
.
Generic Letter 91-13 -2- September 19, 1991
safety issue, the NRC staff has concluded that the following
administrative-type improvements would significantly enhance the
availability of the ESWS in affected plants, and their implementation is
warranted in view of the safety benefit to be derived and the cost of
implementation:
o Technical specification (TS) changes contained in Enclosure 1 to
enhance the availability of the ESWS as applied to the design
configuration of affected plants.
o Improvement of emergency procedures for a LOSW using existing design
features, specifically: (a) operating and maintaining high-pressure
injection (HPI) pump integrity in the event of loss of reactor coolant
pump (RCP) seals as a result of ESWS failure, and (b) testing and
manipulating the ESWS crosstie between the units during a LOSW
accident.
The incorporation of technical specification improvements is consistent with
the Commission's Policy Statement on Technical Specification Improvements.
This policy statement captures existing requirements under Criterion 3
(Mitigation of Design-Basis Accidents or Transients) or under the provisions
to retain requirements that operating experience and probabilistic risk
assessment are shown to be important to the public health and safety.
General Design Criteria 44, 45, and 46 of 10 CFR Part 50, Appendix A, in
conjunction with the probabilistic risk assessment performed under GI-130,
form the technical bases for these TS and procedures improvements.
A backfit analysis of the type described in 10 CFR 50.109(a)(3) and 10 CFR
50.109(c) was performed, and a determination was made that these new TS and
procedures improvements would provide a substantial increase in overall
protection of the public health and safety and that the costs of
implementing these improvements are justified in view of this increased
protection (Enclosure 2). It should be noted that for the benefits of these
improvements to be realized, the guidance contained in GL 89-13 and
Supplement 1 should be considered in the context of the inter-unit crosstie.
Namely, GL 89-13 states: "Redundant and infrequently used cooling loops
should be flushed and flow tested periodically at the maximum design flow to
ensure that they are not fouled or clogged. Other components in the service
water system should be tested on a regular schedule to ensure that they are
not fouled or clogged...."
Enclosure 3 contains a discussion of an additional safety enhancement
identified as part of our evaluation of GI-130 involving installation of a
dedicated RCP seal cooling system similar to that identified also under
GI-23, "Reactor Coolant Pump Seal Failures." The final decision on the
possible backfitting of additional plant improvements has been deferred
until completion of GI-23; and that aspect of GI-130 is subsumed by GI-23.
GI-23 will be resolved following the review of comments received based on
the related Federal Register Notice published on April 19, 1991. The
comment period has been extended until September 30, 1991. Enclosure 3 is
provided to you for information only at this time.
.
Generic Letter 91-13 -3- September 19, 1991
INFORMATION REQUEST (10 CFR 50.54(f))
Addressees are requested to review the recommended TS and procedures
improvements described in the preceding discussion and to evaluate the
applicability and safety significance of those improvements at their
respective facilities. On the basis of results of the recommended
plant-specific evaluations, each addressee shall provide a response to the
NRC pursuant to Section 182 of the Atomic Energy Act and 10 CFR 50.54(f)
which indicates whether or not the recommended TS and procedures
improvements are applicable to its facility, and whether or not the
addressee will incorporate the TS (Enclosure 1) into its license and
implement the procedures improvements. The response shall be provided to
the NRC under oath or affirmation within 180 days of the date of this
letter. If an addressee intends to implement the recommended TS and
procedures improvements, the licensee shall include an implementation
schedule as part of the response to this letter. The licensee should retain
supporting documentation consistent with the records retention program at
each facility.
An evaluation of the justification for this information request has been
prepared in accordance with the requirements of 10 CFR 50.54(f). That
evaluation concludes that the information requested is justified in view of
the potential safety significance of the ESW reliability issue to be
addressed with that information (Enclosure 4). Copies of NUREG-1421 and
NUREG/CR-5526 are also enclosed for your information and to assist you in
evaluating the applicability of this issue to your respective facilities
(Enclosures 5 and 6).
A list of recently issued NRC GLs is enclosed for your information
(Enclosure 7).
This request is covered by Office of Management and Budget Clearance Number
3150-0011, which expires May 31, 1994. The estimated average burden hours
is 50 person hours per owner response, including assessment of the new
recommendations, searching data sources, gathering and analyzing the data,
and preparing 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. Send comments
regarding this burden estimate or any other aspect of this collection of
information, including suggestions for reducing this burden, to the
Information and Records Management Branch (MNBB-7714), Division of
Information Support Services, Office of Information Resources Management,
U.S. Nuclear Regulatory Commission, Washington, D.C. 20555; and to Ronald
Minsk, Office of Information and Regulatory Affairs (3150-0011), NEOB-3019,
Office of Management and Budget, Washington, D.C. 20503.
.
Generic Letter 91-13 -4- September 19, 1991
If you have any questions on this matter, please contact your Project
Manager.
Sincerely,
James G. Partlow
Associate Director for Projects
Office of Nuclear Reactor Regulation
Enclosures: 1. Draft Technical
Specifications (3/4.7.4) 2. Backfit Analysis for GI-130
3. Background Discussion of
a Deferred Safety Enhancement
from GI-130 to GI-23
4. Justification Analysis
[10 CFR 50.54(f)] for Generic Letter on GI-130
5. NUREG-1421
6. NUREG/CR-5526
7. List of Recently Issued NRC
Generic Letters
.
ENCLOSURE I
DRAFT TECHNICAL SPECIFICATION
PLANT SYSTEMS
3/4.7.4 SERVICE WATER SYSTEM
LIMITING CONDITION FOR OPERATION
3.7.4 At least two independent service water loops per unit and the
crosstie between the service water systems of each unit (as applicable)
shall be operable. In addition, the crosstie shall be capable of being
opened [from the main control room] as a flow path between the two units.
APPLICABILITY: Modes 1, 2, 3, and 4.
ACTION:
A. Both units in Modes 1, 2, 3, or 4.
With one service water loop per unit OPERABLE, restore at least two
loops per unit to OPERABLE status within 72 hours, or for the unit with
the inoperable service water loop, be in at least HOT STANDBY within
the next 6 hours and in COLD SHUTDOWN within the following 30 hours.
2. With one [or both] of the crosstie valve(s) INOPERABLE and not
capable of being opened [from the control room], within 72 hours
restore the valve(s) to OPERABLE status or open the affected
valve(s), and maintain the affected valve(s) open; otherwise be in
at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN
within the following 30 hours.
B. One unit in Modes 1, 2, 3, or 4 and one unit in Mode 5 or 6.
1. Verify that at least one pump in the shut down unit is OPERABLE
and available to provide service water to the operating unit. If
neither service water pump in the shut down unit is OPERABLE,
restore at least one pump to OPERABLE status within 72 hours, or
place the operating unit in at least HOT STANDBY within the next 6
hours and in COLD SHUTDOWN within the following 30 hours.
2. With one service water loop in the operating unit INOPERABLE,
restore two loops in the operating unit to OPERABLE status within
72 hours or be in at least HOT STANDBY within the next 6 hours and
in COLD SHUTDOWN within the following 30 hours.
3. With one [or both] of the crosstie valve(s) INOPERABLE and not
capable of being opened [from the control room], within 72 hours
restore the valve(s) to OPERABLE status or open the affected
valve(s), and maintain the affected valve(s) open; otherwise be in
at least HOT STANDBY within the next 6 hours and in COLD SHUTDOWN
within the following 30 hours.
.
DRAFT TECHNICAL SPECIFICATIONS
PLANT SYSTEMS
SURVEILLANCE REQUIREMENTS
4.7.4 Two service water loops per unit shall be demonstrated OPERABLE:
a. At least once per 31 days by verifying that each valve (manual,
power-operated, or automatic) servicing safety-related equipment
that is not locked, sealed, or otherwise secured in position is in
its correct position.
b. At least once per 92 days by cycling crosstie valves and/or
verifying that valves are locked open with power removed; and
c. At least once per 18 months during shutdown, by verifying that:
1. Each automatic valve servicing safety-related equipment
actuates to its correct position on a test signal;
2. Each service water system pump starts automatically on a
test signal; and
3. Each crosstie valve is cycled or is locked open with power
removed.
BASES
3/4.7.4 SERVICE WATER SYSTEM
The OPERABILITY of the service water system ensures that sufficient cooling
capacity is available for continued operation of safety-related equipment
during normal and accident conditions. The redundant cooling capacity of
this system, assuming a single failure, is consistent with the assumptions
used in the accident conditions within acceptable limits.
In the event of a total loss of service water in one unit of a two-unit site
where backup cooling capacity is available via a crosstie between the two
units, the OPERABILITY of the unit crosstie along with a service water pump
in the shut down unit ensures the availability of sufficient redundant
cooling capacity for the operating unit. These limiting conditions will
ensure a significant risk reduction, as indicated by the analyses of a
loss-of-service water system accident. The surveillance requirements ensure
the short-term and long-term operability of the service water system and the
crosstie between the two units. The service water system crosstie between
the two units consists of appropriate piping, valves, and instrumentation
cross-connecting the discharge of the service water pumps of the two units.
By operating the crosstie, the supply of additional redundant cooling
capacity from one unit is available to the service water system of the other
unit.
.
ENCLOSURE 2
BACKFIT ANALYSIS (REFERENCE 10 CFR 50.109)
FOR GENERIC ISSUE 130
A.1 INTRODUCTION
This enclosure presents the backfit analysis for Generic Issue 130 (GI-130),
"Essential Service Water System Failures at Multi-Unit Sites." The
technical findings for GI-130 are presented in NUREG/CR-5526, and the
regulatory analysis is presented in NUREG-1421. The studies apply to 14
reactor units at seven sites and indicate that essential service water
system (ESWS) failures at these plants are a significant contributor to the
overall plant risk. As a consequence of these technical findings, and based
on the cost/benefit analyses performed, the staff has determined that these
14 plants may need to modify technical specifications (TS) to enhance the
availability of the ESWS and to institute procedures to assure the integrity
of the high-pressure injection (HPI) pump in the event of RCP seal failure
as a result of loss of essential service water (LOSW), as well as procedures
to test and manipulate the ESWS crosstie between the two units during a LOSW
accident.
The estimated benefit from the identified safety enhancements is a reduction
in the core damage frequency and a reduction in the associated risk of
offsite radioactive releases as a result of ESW failure. The reduction of
risk to the public (per plant lifetime) is estimated to be 4141 person-rem
(best estimate numbers used) and supports the conclusion that these safety
enhancements provide a substantial increase in the overall protection of the
public health and safety. Also, the direct and indirect costs of
implementation are justified in view of this increased protection.
As discussed in NUREG-1421, when considered individually, most of the
alternatives analyzed for reducing the risk associated with this issue would
be cost-effective in meeting the $1000/person-rem guideline. The objective
of the GI-130 resolution is that the risk from loss of the ESWS be reduced
consistent with the two basic requirements of the backfit rule that the
corrective alternatives be both substantial and cost-effective.
One of the potential improvements consisting of improvements in TS and
emergency procedures was shown to be capable of reducing the core damage
frequency (CDF) from loss of ESW (1.5E-04/RY) by 17 percent (or by
approximately 3.0E-05/RY) in a cost-effective manner. The staff recognizes
the uncertainties in these estimates, and in recognition of the potentially
substantial risk reductions (over 4000 person-rem per plant lifetime), the
staff believes that significant safety improvements can be achieved by low
cost changes in TS and procedures. This is deemed to be consistent with the
provisions of the backfit rule.
The overall approach to arriving at the proposed resolution considered both
the numerical results of the cost-benefit analysis and the spectrum and type
of potential improvements available for potential risk reduction for
.
loss-of-service-water sequences. Those alternatives that could reduce the
number of occurrences of the LOSW initiators would be desirable from the
prevention perspective. Those alternatives that would help to reduce the
consequences of an LOSW would be desirable from the mitigation perspective.
The improvements in the TS would assist on the prevention side, while the
improved procedures would provide a blend of both prevention and mitigation
capabilities.
The conclusion of this backfit analysis is that a substantial increase in
the protection of the public health and safety will be derived from
backfitting of the ESWS improvements and that the backfit is justified in
view of the favorable cost/benefit ratios. In the following sections of
this backfit analysis, the nine factors stipulated by 10 CFR 50.109(c) to be
used in the determination of backfitting are addressed.
A.2 ANALYSIS OF 10 CFR 50.109(c) FACTORS FOR "ALTERNATIVE 5"
A.2.1 Objective
The objective of Alternative 5 (the proposed backfit) is to improve the
performance of the ESW system by providing a blend of both prevention and
mitigation capabilities. This backfit will be applicable to all the
pressurized-water reactor (PWR) plants (14 units) covered by GI-130.
A.2.2 Licensee Activities
To implement "Alternative 5," each licensee would modify TS in accordance
with Enclosure 1 to this generic letter, as well as implement procedures for
operating and maintaining HPI pump integrity and testing and manipulating
the ESWS crosstie between units during a LOSW event.
A.2.3 Public Risk Reduction
Backfitting in accordance with the proposed alternative will yield a
reduction in the incidence of public risk from the accidental offsite
release of radioactive materials of 4141 person-rem (best-estimate) per
plant with an average remaining life of 30 years. This backfit will reduce
the core damage frequency from an LOSW by 17 percent (or by approximately
3.0E-05/RY).
As detailed in Chapter 6 of NUREG-1421, the staff recognizes the
uncertainties in these estimates and has considered both the numerical
results of the cost-benefit analysis as well as the spectrum and type of
potential improvements for risk reductions associated with LOSW sequences.
A.2.4 Occupational Exposure
The radiological operational exposure is negligible and, therefore, the
implementation of Alternative 5 will not result in any increase in the
radiological exposure to facility employees.
.
A.2.5 Installation Costs
The best estimate total cost per reactor associated with Alternative 5 is
$83,000. When the onsite averted costs are taken into account, this
alternative results in a net savings.
A.2.6 Potential Safety Impact
A number of generic safety issues related to GI-130 have been in various
stages of resolution, including some that have already been resolved. The
relation of these issues to GI-130 is as follows:
o GI-23, "Reactor Coolant Pump Seal Failures" -- This generic safety
issue addresses the same possible improvements as Alternative 6
and, in part, Alternative 7 of GI-130. The staff's current
understandings, technical findings, and potential recommendations
regarding GI-23 were issued for public comment. On the basis of
the staff's current knowledge and perspective, the staff has
identified an approach for the resolution of GI-23. This approach
is contained in Draft Regulatory Guide DG-1008.
An objective of the identified approach for the resolution of
GI-23 is to reduce the risk of severe accidents associated with
RCP seal failure by reducing the probability of seal failure, or
to demonstrate that the risk is not significant, thus assuring
that it is a relatively small contributor to total core damage
frequency. The proposed means of doing so entails the
installation of a separate and independent cooling system for the
RCP seals. Hence, implementation of the proposed GI-23 resolution
could provide a substantial portion of the proposed GI-130
resolution. As such, the resolution of GI-130 is coordinated with
the resolution of GI-23 by allowing the installation of a backup
RCP seal cooling system to be deferred to the resolution of GI-23
pending the receipt and review of public comments. It is expected
that information developed as a result of the submittal of public
comments will be helpful in our efforts to better understand the
performance of the RCP seals under loss of seal cooling
conditions.
o GI-51, "Improving the Reliability of Open-Cycle Service-Water
Systems" -- The resolution of this generic safety issue was
reported in August 1989 and its imposition began with the issuance
of Generic Letter 89-13 and Supplement 1. Implementation of the
GI-51 entails the implementation of a series of surveillance,
control, and test requirements to ensure that the ESWS of all
nuclear power plants are in compliance with all applicable
licensing requirements.
During the review of the operational experience data of GI-130,
credit was taken for a corrective measure as a result of the
resolution of GI-51 by excluding those events that involved
biofouling of the ESW. Hence, GI-51 has no direct impact on
GI-130.
.
o GI-153, "Loss of Essential Service Water in LWRs" has been
assigned NRC staff resources for its resolution. Its purpose is
to assess this issue for all light-water reactors (LWRs) not
already covered by GI-130. Insights gained by the evaluation of
GI-153 are expected to be useful in confirming and/or
supplementing the technical findings of GI-130.
Of interest to the decision process on this generic issue are the insights
and reviews available in related probabilistic risk assessment (PRA)
documentation in the open literature. The PRA work available in NUREG-1150,
"Severe Accident Risks: An Assessment for Five U.S. Nuclear Power Plants"
(plus supporting documentation) is a source of extensive risk analyses
information that might be used for an understanding of ESW vulnerabilities.
An examination of the NUREG-1150 documentation of the three PWRs that were
studied indicates that the analyst thought that the ESW redundancy for two
of the three PWRs was large enough that a complete loss of ESW as an event
initiator was deemed not credible (eight pumps are available at Sequoyah,
Units 1 and 2). None of the five plants in the NUREG-1150 study is a GI-130
plant; however, it is worthwhile to note that one of the PWRs (Zion)
identified the service water contribution to CDF to be substantial
(approximately 1.5E-04/RY). This contribution for Zion was approximately 42
percent of the total core damage frequency from all causes.
Another PRA work available in the open literature is NSAC-148, "Service
Water Systems and Nuclear Plant Safety," dated May 1990. Although NSAC-148
is only a compilation of earlier PRA results for six plants performed by the
industry, it is useful to note that a greater appreciation of the service
water system's contribution to plant risk has moved the industry to initiate
a program to improve service water performance. The limited guidance
available in NSAC-148 is a step in the right direction. The wide range of
core damage frequencies (from LOSW) at the six plants studied suggests the
large variability in plant-specific ESW configurations. The average CDF
from LOSW for the six plants was 6.55E-05/RY, with a range of
2.33E-04/RY-to-"negligible" contribution. Although many details of these
six PRAs are not included in NSAC-148, and therefore, must be considered to
be used only with great caution, the overall message that the service water
system provides an important safety function that could be a substantial
contributor to overall plant risk tends to lend added credence to the GI-130
conclusions.
A.2.7 NRC Costs
Implementation of Alternative 5 is estimated at $21,000 (best estimate).
This estimate assumes minimal resources for review of the generic letter
responses.
A.2.8 Facility Differences
Alternative 5 is applicable to all 14 plants covered by this study,
regardless of age or design. Other PWR and BWR plants that are not included
under the resolution of GI-130 will be evaluated under GI-153, "Loss of
Essential Service Water in LWRs."
.
A.2.9 Term of Requirements
This represents the final resolution of GI-130. Alternative No. 6 entailing
the installation of an independent RCP seal cooling system has been subsumed
under the resolution of GI-23.
.
ENCLOSURE 3
BACKGROUND DISCUSSION OF A DEFERRED SAFETY ENHANCEMENT
FROM GI-130 TO GI-23
(INSTALLATION OF A DEDICATED RCP SEAL COOLING SYSTEM)
As identified in NUREG-1421, "Regulatory Analysis for the Resolution of
Generic Issue 130: Essential Service Water System Failures at Multi-Unit
Sites," a combination of potential improvements consisting of the
installation of a backup, dedicated RCP seal cooling system, and
improvements in technical specifications (TS) and procedures are shown to be
capable of substantial risk reduction. The specific features of such a
backup, dedicated RCP seal cooling system would be as follows:
o Single high pressure pump, 50-100 gpm capacity
o Dedicated water storage tank with capacity to last at least 8-10
hours
o AC-independent (non-seismic) pump
o No support system cooling required
o Once-through RCP seal heat removal
Limited plant-specific information obtained through the existing literature
(FSARs, and so forth), site visits, or discussions with licensees have
indicated that a number of the units covered by GI-130 already have
plant-unique features that could be responsive to this generic safety
enhancement. Rather than attempting to perform a series of PRAs tailored to
each of the 14 units, the NRC encourages each licensee or applicant to
review the plant-specific features (if any) that could be credited with
departing from the generic (representative) base case plant configuration
modelled in NUREG/CR-5526. In addition, other design alternatives may also
be considered utilizing arrangements different from that of the
high-pressure pump seal injection.
One such alternative would provide flow through the RCP thermal barrier heat
exchangers by connecting the fire water system into the component cooling
water (CCW) lines. Most fire water systems have one diesel-driven fire
water pump, which usually is independent of the ESWS.
Generic Issue 23, "Reactor Coolant Pump Seal Failures," deals with this
recommendation also, and specific guidance for resolving that generic issue
is given in proposed Regulatory Guide DG-1008. While awaiting completion of
public review and comment on draft Regulatory Guide DG-1008, resolution of
this GI-130 item has been deferred until GI-23 is resolved. The reason for
this deferral relates to the earlier development and promulgation of 10 CFR
50.63 (station blackout rule), which was based on an assumption regarding
the magnitude of RCP seal leakage during a station blackout event. While it
was
.
- 2 -
left to GI-23 to validate that assumption, the resolution of GI-130 is also
based on a RCP seal failure LOCA model very similar to that of GI-23, but
different from the leakage assumption in 10 CFR 50.63.
.
ENCLOSURE 4
JUSTIFICATION ANALYSIS [10 CFR 50.54(f)]
FOR GENERIC LETTER ON GENERIC ISSUE 130
Section 50.54(f) of 10 CFR Part 50 requires that "... the NRC must prepare
the reason or reasons for each information request prior to issuance to
ensure that the burden to be imposed on respondents is justified in view of
the potential safety significance of the issue to be addressed in the
requested information." Further, Revision 4 of the Charter of the Committee
To Review Generic Requirements (CRGR), dated April 1989, specifies that, at
a minimum, such an evaluation shall include the following:
a. A problem statement that describes the need for the information in
terms of potential safety benefit,
b. The licensee actions required and the cost to develop a response
to the information request, and
c. An anticipated schedule for NRC use of the information.
The staff's 10 CFR 50.54(f) evaluation of the information request addressing
the above elements follows:
a. Problem Statement That Describes the Need for the Information in
Terms of Potential Safety Benefit
The recommended resolution of Generic Issue 130 (GI-130),
"Essential Service Water System Failures at Multi-Unit Sites,"
applies to 14 reactor units at seven sites and indicates that
essential service water system (ESWS) failures at these plants may
significantly contribute to the overall plant risk. As a
consequence of these technical findings, and based on the
cost/benefit analyses performed, the staff has determined that
these 14 plants may need to modify technical specifications (TS)
to enhance the availability of the ESWS and to institute
procedures to assure the integrity of the HPI pump in the event of
RCP seal failure as a result of loss of essential service water
(LOSW), as well as procedures to test and manipulate the ESWS
crosstie between the two units during a LOSW accident.
The estimated benefit from the identified safety enhancements is a
reduction in the core damage frequency and a reduction in the
associated risk of offsite radioactive releases as a result of ESW
failure. The reduction of risk to the public (per plant lifetime)
is estimated to be 4141 person-rem (best estimate numbers used)
and supports the conclusion that these safety enhancements provide
a substantial increase in the overall protection of the public
health and safety. Also, the direct and indirect costs of
implementation are justified in view of this increased protection.
The staff recognizes the uncertainties in these estimates, and in
recognition of the potentially substantial risk reductions, the
staff believes that significant safety improvements can be
achieved by low cost changes in TS and procedures, consistent with
the provisions of the backfit rule.
.
As discussed in NUREG-1421, when considered individually, most of
the alternatives analyzed for reducing the risk associated with
this issue would be cost-effective in meeting the $1000/person-rem
guideline. The objective of the GI-130 resolution is that the
risk from the loss of the ESWS be reduced consistent with the two
basic requirements of the backfit rule that the corrective
alternatives be both substantial and cost-effective.
One of the potential improvements consisting of improvements in TS
and emergency procedures was shown to be capable of reducing the
CDF as a result of loss of ESW (1.5E-04/RY) by 17 percent (or by
approximately 3.0E-05/RY) in a cost-effective manner. As
discussed earlier, this is deemed to be consistent with the
provisions of the backfit rule.
The overall approach to arriving at the proposed resolution
considered both the numerical results of the cost-benefit analysis
and the spectrum and type of potential improvements available for
potential risk reduction for loss-of-service-water sequences.
Those alternatives that could reduce the number of occurrences of
the LOSW initiators would be desirable from the prevention
perspective. Those alternatives that would help to reduce the
consequences of a LOSW would be desirable from the mitigation
perspective. The improvements in the TS would assist on the
prevention side, while the improved procedures would provide a
blend of both prevention and mitigation capabilities.
The conclusion of our analysis is that a substantial increase in
the protection of the public health and safety will be derived
from the improvements in the TS and procedures, which are
justified by the favorable cost/benefit ratio. Hence, in view of
the safety significance of the recommended resolution of GI-130,
the issuance of this generic letter under 10 CFR 50.54(f) is
justified. (See also Item b. below.)
b. The Licensee Response Required and the Cost to Develop the
Responseto the Information Request
All the recipient licensees or applicants of this generic letter
would be requested to review the TS and procedures improvements
identified as part of our evaluation of GI-130 and to assess the
applicability of these improvements to their respective
facilities.
We estimate that the cost of reviewing and evaluating the contents
of this generic letter and preparing a response will cost no more
than $2500 per licensee or applicant. It is expected that this
cost may
.
vary from site to site, depending on the degree to which the TS
and procedures improvements apply to individual plants. This cost
is insignificant compared to the cost-justified improvements (see
cost estimates presented in NUREG-1421), which represent a
substantial safety improvement.
c. An Anticipated Schedule for the NRC Use of the Information
We expect that the responses to this generic letter would be
submitted within the 180-day schedule required by the generic
letter, and that NRC staff review of the responses will be
completed within 180 days from their receipt.
