ML20029E166
| ML20029E166 | |
| Person / Time | |
|---|---|
| Issue date: | 04/21/1994 |
| From: | Allison D NRC |
| To: | NRC |
| Shared Package | |
| ML20029E167 | List: |
| References | |
| REF-GTECI-023, REF-GTECI-NI NUDOCS 9405170135 | |
| Download: ML20029E166 (6) | |
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APR 211991 MEMORANDUM FOR:
James M. Taylor Executive Director for Operations FROM:
Edward L. Jordan, Chairman Committee to Review Generic Requirements
SUBJECT:
MINUTES OF CRGR MEETING NUMBER 195 The Committee to Review Generic Requirements (CRGR) met on Wednesday, November 28, 1990 from 1:00 - 5:00 p.m.
A list of attendees at the meeting is enclosed (Enclosure 1). The following items were discussed at the meeting:
1.
J. Heltemes, R. Baer and J. Jackson of RES continued the presentation for CRGR review of a proposed resolution for Generic Issue 23, Reactor Coolant Pump Seal Failure.
(The review of this matter was begun at Meeting No. 194). At the conclusion of the discussions the staff agreed to revise the package substantially and provide a revised package to the CRGR. The revised package would then be circulated to CRGR members and a decision would be made as to whether further discussion at another meeting would be needed. This matter is discussed in Enclosure 2.
2.
A briefing on the standard technical specification program and related matters, which had been scheduled for Meeting No.195, was cancelled prior to the meeting.
It was expected that this briefing would take place at Meeting No. 196.
In accordance with the ED0's July 18, 1983 directive concerning " Feedback and Closure of CRGR Reviews," a written response is required from the cognizant office to report agreement or disagreement with CRGR recommendations in these minutes. The response, which is required within five working days af ter receipt of these minutes, is to be forwarded to the CRGR Chairman and if there is disagreement with CRGR recommendations, to the ED0 for decisionmaking.
Questions concerning these meeting minutes should be referred to Dennis Allison (492-4148).
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)i ' l AdwTrd L.
rdan, Chairman Committee o Review Generic Requirements
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Enclosures:
As stated cc/w enclosures: See page 2 h{dhMC b
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l 5
2 cc/w enclosures:
Commission (5)
SECY J. Lieberman P. Norry D. Williams Regional Administrators CRGR Members R. Cunningham J. Murphy S. Lewis A. Gibson 11
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ATTENDANCE LIST CRGR Meeting No.'195-November 28, 1990 CRGR Members NRC Staff E. Jordan J. Heltemes R. Cunningham (for G. Arlotto)
R. Baer
- S. Lewis (for J.~ Moore)
S. K. Skaukat F. Miraglia J. Jackson J. Murphy (for B. Sheron)
B. Richter L. J. Callan P. Boehnert B. Mendelsohn CRGR Staff C. Hendren T. Dipalo J. Conran D. L. Basdekas D. Allison A. Buslik D. Ross A. S. Masciantonio L. E. Kokajko T. E. Collins D. Thatcher
'i M. Taylor B. Hardin G. Holohan A. Thadani i
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1 l to the Minutes of CRGR Meetina No. 195 Proposed Resolution of Generic Issue 23. Reactor Coolant Pump Seal Failure
.I November 28, 1990 TOPIC J. Heltemes, R. Baer and J. Jackson of RES continued the presentation for CRGR review of a proposed resolution for Generic Issue 23, Reactor Coolant Pump Seal Failure. The review of this matter was begun at Meeting'No. 194. The basic proposal was to publish for comment a proposed resolution which would indicate the following new positions in the form of a generic letter and regulatory guide:
1.
Apply Appendix B quality assurance provisions to activities affecting the reactor coolant pump seals.
2.
Establish instrumentation, procedures and training for monitoring reactor coolant pump seal operations and for handling abnormal indications.
3.
Provide alternate seal cooling or some alternative that would ensure seal integrity under station blackout conditions and certain loss of seal cooling events.
It should be noted that the staff was recommending that the above positions be published for comment rather than recommending that the positions be issued for implementation.
The principal comments and questions raised in the initial discussions, at Meeting No. 194, are summarized in the minutes of that meeting.
BACKGROUND The basic review package is described in the Minutes of Meeting No. 194.
The following additional material was obtained as background for the
)
discussions at Meeting No. 195.
1.
Draft Revision 2, Enclosure 1, NUREG-1421, Regulatory Analysis for the Resolution of Generic Issue 130: Essential Service Water System Failures at Multi-Unit Sites, Draft Report for Comment, Manuscript Completed October 1990 (Attachment 1).
2.
Memorandum dated November 26, 1990 for E. Jordan from J. Murphy,
Subject:
Thoughts on the backfit rule 50.109 and the safety goal policy statement (Attachment 2).
I
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3.
Three graphs entitled Safety Goal analysis showing NUREG-ll50 results for core damage frequency and conditional containment failure probability, including uncertairty bands, superimposed on a. template which indicates various regimes for implementing the safety goals (Attachment 3).
CONCLUSIONS / RECOMMENDATIONS 1.
Disposition At the conclusion of the discussions the staff agreed to modify the package substantially and provide a revised package to the CRGR. The revised package would then be circulated to CRGR members and a decision would be made as to whether further discussion at another meeting would be needed. The CRGR was in agreement with this approach.
It was noted that some of the features of the revised package would be as follows:
A.
The package to be published for comment would be integrated into a more coherent whole. The perspective would be that it is a close call and our minds are not made up. The questions upon which we are specifical.y inviting comments would be better articulated.
Although the perspective would be changed as indicated above, the staff believed that it might be beneficial to still include the regulatory guide and/or draft regulatory analysis to provide a concrete basis for comments. Some members had reservations about this approach because it might point narrowly toward three specific fixes rather than encouraging thought about a wide range of possibilities, including better information on current seals, seal redesign and testing, alternate ac electric power or alternate core cooling. However, it might still be feasible to include the regulatory guide and regulatory analysis if they could be put in proper context and the request for comments would focus attention on a wide range of possibilities.
B.
A federal register notice would be used to solicit comments and articulate questions about the issue. A generic letter to licensees would also be used. The generic letter would be similar to Enclosure 8 in the review package. In addition to soliciting comments, the generic letter would inform the licensees of a potential impact of this issue on any coping analyses used to demonstrate compliance with the station blackout rule.
j C.
It would be appropriate to solicit comments on whether the resolution of this issue should be integrated with the resolution of related issues such as Generic Issues 130 and 153.
l D.
It may be appropriate to discuss the draft AE00 study on operating experience which was discussed at Meeting No. 194.
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2.
Safety goals The overall discussion of safety goals was considered adequate. Since the perspective would be changing towards one of requesting information.
rather than proposing requirements or positions, a determination-regarding consistency with the safety goals would not be necessary or e
appropriate.
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DRAFT REVISION 2 ENCLOSURE 1 NUREG-1421 REGULATORY ANALYSIS FOR THE RESOLUTION OF GENERIC ISSUE 130:
ESSENTIAL SERVICE MATER SYSTEM FAILURES AT MULTI-UNIT SITES Draft Report for Comment Manuscript Co=pleted:
October 1990 Date Published:
TBD V.
- Leung, D.
- Basdekas, G. Masatis Reactor and Plant Safety Issu3s Branch Division of Safety Issue Resolution Office of Nuclear Regulatory Research U.S. Nuclear Regulatory Commission Washington, D.C.
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l e-ABSTRACT The Essential Service Water System (ESWS) is required-to provide cooling in nuclear power plants during normal operation and ---
accident conditions.
Typical equipment supported by the ESWS are component cooling water heat exchangers, containment spray-heat exchangers, high pressure injection pump oil coolers, amargency diesel generators, and auxiliary building ventilation coolers.
Failure of the ESWS function could lead to severe consequences.
This report presents the regulatory analysis for GI-130 i
" Essential Service Water System Failures at Multi-Unit Sites."
The risk reduction estimates, cost benefit analyses, and other insights gained during this effort have shown that implamentation of the recommendations will significantly reduce risk and that these improvaments are warranted in accordance with the Backfit Rule, 10 CFR Part 50.109 (a) (3).
l 1
iii
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a TABLE OF CONTENTS Page ABSTRACT iii LIST OF TABLES vii EXECUTIVE
SUMMARY
ix 1.
STATEMENT OF TEE PROBLEM 1
2.
OBJECTIVI 4
3.
ALTERNATIVI RESOLUTIONS 5
3.1 Alternative 1 - No Action 5
3.2 Alternative 2 - Install Additional Crosstie..
5 3.3 Alternative 3 - Provide Electrical Power Cross-Connection 5
3.4 Alternative 4 - Provide Separate Intake Structure 5
3.5 Alternative 5 - Modify Technical Specifications 6
3.6 Alternative 6 - Provide Independent RCP Seal Cooling System 6
3.7 Alternative 7 - Combine Alternativ.es 5 and 6 7
4.
TECENICAL FINDINGS 8
4.1 Core Damage Frequency Analysis B
4.1.1 Initiating Events Frequency 8
4.1.2 ESW and RCP Seal LOCA Recovery 9
4.1.3 Relative Time Fractions 9
4.1.4 Core Damage Frequency 10 4.1.5 Effects of Potential Improvements on Core
)
Damage Frequency 11 4.2 Dose Consequence Analysis 20 4.3 Cost Analysis 23 1
4.3.1 Direct Costs 23 4.3.2 Indirect costs 23 4.3.3 Operating and Maintenance Costs 23 4.3.4 Technical Specifications Costs 24 4.3.5 NRC Costs 24 j
4.3.6 Averted Onsite Costs 24 1
4.3.7 Range of Cost Estimates 25 1
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a TABLE OF CONTENTS (continued)
Page 5.
VALUE/ IMPACT ANALYSIS 31 5.1 Alternative 1 - No Action 31 5.2 Alternative 2 - Install Additional Crosstie 31 5.3 Alternative 3 - Provide Electrical Power Cross-Connection 32 5.4 Alternative 4 - Provide Separate Intake Structure 32 5.5 Alternative 5 - Modify Technical Specifications 33 5.6 Alternative 6 - Provide Independent RCP Seal Cooling System 34 5.7 Alternative 7 - Combine Alternatives 5 and 6 36 5.8 Uncertainty Analysis 36 5.9 Life Extension Considerations 37 6.
DECISION RATIONALE 40 7.
IMPLEMENTATION 46 8.
REFERENCES 47 i
vi
LIST OF TABLES Page TABLE ES.1 Best-Estimate Cost-Benefit Ratios xii 4.1.1 Operational Status of Multi-Unit Sites 13 4.1.2 State Dependent LOSW Initiating Event Frequencies 14 4.1.3 Sequence Conditional Core Damage Probabilities 15 4.1.4 Core Damage Frequency due to Individual Sequences 16 4.1.5 Core Damage Frequency --- Summary.
17 4.1.6 Failure Mode Classification 18 4.1.7 CDF Reduction For Alternatives 19 4.2.1 Benefits of Proposed Alternatives 22 4.3.1 Best Estimate Costs of Proposed Alternatives 26 4.3.2 Direct Cost Estimates 27 4.3.3 Cost Offsets for Proposed Alternatives 28 4.3.4 Onsite Consequences 29 4.3.5 Range of Estimates for the Total Cost and the Net Cost 30 5.1 Best-Estimate Cost-Benefit Ratios 38 5.2 Best-Estimate Cost-Benefit Ratios for 20-years License Renewal 39 vii 8
EXECUTIVE SUMHARY This report provides supporting information, including a value-impact analysis, for the Nuclear Regulatory Commission's (NRC' s )
resolution of Generic Issue 130, " Essential Service Water System Failures at Multi-Unit Sites."
This issue addresses the concerns regarding the Essential Service Water (ESW) system at multi-unit PWR sites having two ESW trains per unit with a crosstie capability (fourteen reactor units at seven sites).
Typical components cooled by the ESW system under normal and accident conditions are the component cooling heat exchangers, containment spray heat exchangers, high pressure injection pump oil coolers, amergency diesel genwrators, and auxiliary building ventilation coolers.
The ESW systam is also used for cooling the reactor coolant pump (RCP) seals, typically indirectly via the component cooling water system (CCWS) or the charging pumps.
This issue was initially identified as a result of the safety evaluation related to the limiting condition for operation (LCO) relaxation program for Byron Unit 1.
ESW systam support from Byron Unit 2 via the crosstie between the two units was not available while Unit 2 was under construction.
To support the LCO relaxation program, Byron Unit 1 performed a probabilistic risk assessment (PRA) of the ESW system.
The insights derived from that study indicated that the core damage f quency due to the unavailability of a two train (one pump per train) ESW systam could present a significant risk to the public health and safety, particularly if one ESW pump from the adjacent unit via an ESW system crosstie is not available.
At multi-unit sites, crossties are usually provided ber. ween the ESW systems of the adjacent unit to enhance operational flexibility; however, the Technical Specifications (TS) for these plants have typically not placed any operability requiraments in the adjacent unit's ESW systam, particularly during shutdown modes 5 and 6.
This regulatory analysis is partly based on a modified reliability analysis performed by Brookhaven National Laboratory (BNL) for the Byron plant.
The PRA model was modified to reflect the multi-unit configuration and the assumption of having an ESW systam failure as an initiating event for the accident sequence.
Also, it was determined that a more recent value for RCP seal LOCA probability based on the data developed in NUREG-ll50 shoeld be established for the present analysis.
A model was developed to incorporate the probability of an RCP seal LOCA as a function ix f
of time and leakage rate of the reactor coolant pump seal.
In addition, both short and long term recovery actions which might affect the final outccme were examined.
The results of the analysis indicate that the core damage frequency (CDF) due to ESW systam failure is estimated to be 1.52E-04 per reactor year.
The staff examined seven possible alternatives to lower the CDF, and estimated that the potential reductions in CDF range from 1.37E-05 to 9.13E-05 per reactor year.
A detailed description of modeling and assumptions used in the analysis are presented in NUREG/CR-5526.
A cost-benefit evaluation of the possible alternatives indicate that cost-effective options are available.
One or more of these alternatives have the potential for significantly reducing the risk due to loss of ESW.
Table ES.1 provides a swanary of the best estimate cost-benefit ratios for each of the alternatives examined.
Comparison of the best estimate cost-benefit ratios for all the alternatives against a guideline cost-benefit ratio of $1000/ person-ram shows that all the alternatives are cost-beneficial except Alternative 4 entailing a separate intake s t ru cture.
The regulatory analysis used these cost-benefit calculations as partial basis for considering a proposed resolution to GI-130.
The proposed resolution is a combination of Alternative 6 (or 6a) plus Alternative 5 to provide a backup means of RCP seal cooling plus additional ESW technical specifications and amargency procedures.
The cost-benefit ration were also calculated for the case of licence renewal for an additional term of 20 years, or a remaining plant life of 50 years.
A comparison of the results shows that the cost-benefit ratios for all analyzed backfit alternatives are considerably lower for extended plant life.
Even so, Alternative No.
4, Separate Intake Strvcture, still remains appreciably higher than the $1,000/ person-rem guideline at a cost-benefit ratio of $2,285/ person-ram.
j 1
of interest to the decision process on this generic issue are the l
insights and views available in related PRA documentation in the i
open literature.
Although still not finalized, the preliminary PRA work available in NUREG-1150, "Severo Accidents Risks: An Assessment for Five U.S. Nuclear Power Plants" (plus supporting documentation) is a source of extensive risk analyses information one might turn to for an understanding of ESW vulnerabilities.
An examination of the NUREG-1150 documentation of the three PWRs that were studied indicates that the analyst considered that the ESW redundancy for two of the three PWRs was large enough that a x
s
a complete loss of ESW as an event-initiator was deemed not credible (eight pumps available in Sequoyah, Unit 1).
None of the five plants in the NUREG-ll50 study is a GI-130 plant; however, it is worthwhile to note that one of the PWRs (Zion),
identified the service water contribution to risk to be substantial (approximately 1.5E-4 /RY).
This contribution for Zion was approximately 42% of the total core damage frequency due to all causes.
Another PRA work available in the open literature is NSAC-148,
" Service Water Systems and Nuclear Plant Safety," dated May 1990.
Although it 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 systam'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 (due to LOSW) over the six plants studied suggests large variability in plant-specific ESW configurations.
The average CDF due to LOSW for the six plants was 6.55E-05/RY, with a range of 2.33E-04/RY-to " negligible" contribution.
Many vetails of these six PRAs are not included in NSAC-148 and, therefc.re, tast be considered to be used only with a great caution.
The overall message that the service water system provides an important safety function which could be a substantial contributor to overall plant risk tends to lend added credence to the GI-130 conclusions.
l xi l
Table ES.1 Best Estimate Cost-Benefit Ratios ($/ Person-Ram)
Total Cost /
Total Cost /
Alternative Benefit without Benefit /Nith Averted onsite Averted onsite Costa Costa 1.
No Action 2.
Additional Crosatie 433 238 3.
Electrical Cross-Connection 80 Note 1 4.
Separate Intake Structuro 3847 3651 5.
Technical Specification Modifications + Procedures 25 Note 1 6.
High Pressure Pump for RCP Seal Cooling 862 684 6a.
Firewater for Thermal Barrier Cooling 37 Note 1 7.
Combination 6 + 5 756 574 7a.
Co=bination 6a + 5 39 Note 1 Note 1:
Including averted onsite costs resulted in a net cost savings.
xii E
1 l.
STATEMENT OF THE PROBLEM This issue was identified in 1986 (Refs.
1,
- 2) as a result of the Byron Unit i vulnerability to core-damage sequences in the absence of the availability of Byron Unit 2 (not operational at the time).
Because of the licensing considerations of Byron Units 1 and 2 and the immediate need to make a third ESW pump available to Byron Unit i via a crosstie with one of the two Byron Unit 2 ESW pumps, the Byron Unit 1 concern was treated as a plant-specific issue.
However, the Byron plant-specific issue raised questions relative to multi-unit sites that have only two ESW pumps / unit with a crosstie capability between tham.
Fourteen units at seven sites having the basic Byron ESW configuration were evaluated as part of this issue.
These multi-unit sites have two ESW pumps per unit (one per train) with a sharing of one of the two pumps with the other unit via a crosstie between the two units.
Evaluation of other design configurations of ESW systems in LWRs, including those of single unit sites, will be performed under GI-153, " Loss of Essential Service Water in LWRs."
It should be noted that the success criteria for the ESW systams in providing adequate cooling capability during normal, accident, and post-accident conditions are design-specific, depending on the plant configuration, the capacities of the ESW pumps, and equipment dependencies on ESW cooling.
Although the success criteria may be as varied as the ESW systams, this evaluation assumed a generic set of success criteria as a representative model for purposes of quantifying the events leading to possible core-damage accidents.
These generic criteria are discussed below and apply only to multi-unit sites having two ESW pumps / plant with a crosstie capability between them.
During normal operation, one ESW pump per unit provides adequate cooling to systems such as CCW, RCP seals and air conditioning and ventilation systems.
The second ESW pump per unit is assumed to be normally in a standby mode.
Because of load shedding (isolati on of non-essential equipment), one ESW pump per unit is assumed capable of handling accident and cooldown heat loads.
Tfpical equipment cooled by the ESW under these conditions are the CCW heat exchangers, containment spray heat exchangers, diesel generators, and auxiliary building ventilation coolers.
With one plant in power operation, and the second plant in the shutdown or refueling modes of operation, the criteria assume one ESW pump can provide adequate cooling to shut down the operating plant through the crosatie connection, should the need arise.
1 l
A survey of operational experience (Refs. 3 and 4) shows that a nesber of different components in the ESW system may fail to perform their intended function in a variety of ways.
- However, review of operating experionce has indicated that there are specific dominant failure modes for the ESW system associated with failures of certain components.
Such failures have involved the traveling screens and/or other common cause problems at the intake structure leading to the partial or complete loss of the water supply.
The ESW pumps and their electrical power supply are other important contributors to the partial or total loss' of the ESW systam.
All ESW systems at the GI-130 multi-unit sites are safety systems, and their designs are plant-specific with plant-specific equipment, crosstie capability, and ESW operability needs for successful accident mitigation operations.
A comprehensive review and evaluation of the operating experience with ESWS has been performed and is reported in NUREG/CR-5526 (Ref. 3).
Excluding system fouling (sediment, biofouling, corrosion, erosion), the total number of plant events involving a possible complete loss of the ESW function was 12 (Ref. 3, Appendix B).
System fouling data were noted, but excluded from the current analysis due to the earlier resolution of Generic Issue 51, " Improving the Reliability of Open Cycle Service Water Systems" (see also the discussion in Chapter 6).
The total number of PWR years during this period of data retrieval was calculated to be 667 reactor-years.
In 1980, one event involved a complete loss of ESW at San Onofre, Unit 1.
At 100% power, a shaft on the operating salt water cooling (SWC) pump sheared due to vibration.
This event then involved the additional failure of the normal standby pump (discharge valve failed to open) as well as the failure of a second auxiliary standby pump (lost prime).
This led to a complete loss of ESW flow for about 15 minutes, at which time a fourth pu=p was manually crossconnected from the traveling screen wash system to establish cooling water flow.
A detailed examination of the loss of ESW events indicatos that a number of events occurred in Modes 5 and 6 (shutdown) and some of them may not have been a complete loss of ESW in terms of total stoppage of ESW flow, even though the ESW system might have been declared inoperable.
The difference of the ESW system between power and shutdown operation is primarily the actual heat load and equipment affected by the loss of ESW.
In addition, the actual 2
t
administrative requirements imposed by the technical i
specifications also differ, and make these two operational modes more distinct.
To calculate the initiating event frequency for loss of ESW, the total operating ESW-system-years for all PWRs of 667 reactor-years was divided into two parts as follows:
487 reactor-years-at-power 180 reactor-years-at-shutdown 1
Finally, the respective loss of all ESW frequencies were calculated to be 1.1E-03 per reactor-year-at-power, and 3.2E-02 j
per reactor-year-at-shutdown (with one pump running and one at standby), and 2. 9E-01 per reactor-year-at-shutdown (with one pump running and the other in maintenance).
These numbers then were weighted for the various operational states of each unit and their respective time fractions, before calculating the CDF values, as discussed in Section 4.1.1.
Should a loss of the ESW system function fail to be recovered, the resulting core-damage accident could lead to significant risk to the public.
The most dominant sequence is the reactor coolant pump seal loss of coolant accident (RCP-LOCA).
This specific sequence is the subject of GI-23, "Reector Coolant Pump Seal Failures" (Ref. 7).
This study estimated the total core
{
damage frequency (CDF) attributable to the loss of ESW for seven 1
two-ur't sites (Chapter 4) and the cost-effectiveness of several alternative modifications (Chapter 5) which could lower this CDF.
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o 2.
OBJECTIVE The purpose of the Generic Issue 130 program is to evaluate the safety adequacy of a two-pump ESW system in existing multi-unit FWR power plant sites, and to examine the cost-effectiveness of alternative measures for reducing the overall vulnerability to ESW system failures.
Probabilistic methods were used to assess the CDF, the potential reduction in risk of the modifications, and their cost-effectiveness.
The overall objective for resolution of GI-130 is i
that contribution from loss of the ESW system shov1d be a small percentage of the total CDF due to all causes.
For USI A-45, the staff recommended in NUREG-1289 that the frequency of events related to DER _ failure leading to core damage
{
should be reduced to a level (around 1.0E-5/RY) so that the probability of such an accident in the next 30 years would be about 0.03 based on a population of around 110 plants.
A similar j
core damage objective (1.0E-5/RY) was noted in USI A-44 covering station blackout.
These objectives are also consistent with the recantly issued guidance to the staff (Ref. 6) setting a goal for CDF of less than 1.0E-04/RY from all contributors.
To meet such a goal the staff has aimed for the benchmark that a single contributor to the CDF contributes no more than 10% of the above suggested value, or no more than 1.0E-05/RY, The application of the safety goal guidance and the objectives of previously resolved USIs, as discussed above, to GI-130 was limited to using them as general guidelines to the decision process described in Chapter 6.
Rigid application of such a quantitative objective to define an absolute requirament was not made.
Since the ESW vulnerability issue is only a fraction of the total contribution I
to risk due to all causes, the current-safety goal guidance that the overall mean frequency of a large release should be less than 1 in 1,000,000 per year is not directly usable to this case.
This is partly because an overall PRA due to all causes was not in the scope of GI-130.
However, consistent with current policy guidance in References 5 and 6, a judgement was made that, in light of the safety goals and available knowledge, the recczmendation to backfit selected design and operational improvements to reduce risk due to ESW failures is warranted (Chapter 6).
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3.
ALTERNATIVE RESOLUTIOP3 There were several alternatives considered for the resolution of Generic Issue 130.
These alternatives are described below.
3.1 Alternative 1 - No Action Under this alternative there would be no new regulatory requiraments.
Consistent with existing regulations, this alternative does not preclude a licensee, or an applicant for an operating license, from proposing to the NRC staff design changes intended to enhance the reliability / operability of the Essential Service Water System and its components on a plant-specific basis.
3.2 Alternative 2 - Install Additional Crosstie The ESW systems of the seven multi-unit sites analyzed under GI-130 are cross-connected through pipe connections and isolation valves.
This arrangament allows the operator of one unit to utilize the ESW cooling capacity of the other unit under most circumstances.
In most cases, the crosstie isolation valves can be ramotely operated.
A hardware failure to open the isolation l
valves, should the need arise, could result in adverse I
conditions.
A parallel cross-connection could reduce the possibility of this kind of failure, and in. addition would allow for more flexible maintenance options.
3.3 Alternative 3 - Provide Electrical Power Cross-Connection In general, the electrical power supplies to the ESW trains are separated and have no cross-connection capability, i.e.,
the Train A ESW pump cannot be powered from electrical Trair B (or Diesel B).
This alternative investigated the implamentation of crossties between the electrical trains of the unit with respect to the operation of the two ESW pumps (Trains A and B).
The cross-connection of electrical power supply of other electrical components, such as HOVs was not considered as part of this alternative because of their less significant potential contribution to risk as observed in the operational experience failure data.
3.4 Alternative 4 - Provide Separate Intake Structure The intake structure is usually a single structure divided into separate bays by concrete walls.
There are a number of screens 5
I l
installed to prevent the intake from passing large foreign i
objects.
The common mode failure of these screens may occur as a I
loss of the co= mon inlet and/or common water source.
The whole intake structure or screens could be affected by events such as
)
flooding or freezing.
~
{
I The alternative considered here is a completely separate intake l
structure and swing pump serving as a redundant intake source of ESW water.
It may be located on the same water source, but in a physically separate location.
An alternate design, which would provide additional independence / diversity, would be to install the additional intake structure on a physically separate water l
source (e.g.,
pond or lake).
The separate intake structure alternative includes the structure, screens, associated motors, valves and piping.
A swing ESW pump would also be made availt a to either unit with redundant electrical power supplies.
Common mode failure considerations are assumed,to play a primary role in j
the design and installation of the new structure (such as heated spaces in areas of the country subject to freezing conditions).
3.5 Alternative 5-Modify Technical Specifications (TS)
Re quirement s In operating modes 5 and 6 (shutdown and refueling, respectively), the status of ESW pumps is uncertain because TS typically do not require that the ESW pumps be operational in these shutdown modes.
This alternative partially involves imposing an explicit operability requirement on at least one of the ESW pupps of a unit while in modes 5 and 6 to provide backup for the other unit ESW system.
An additional improvament is the testing of the unit crosstie valves to provide greater assurance of operability, thereby reducing the hardware failure assumptions on the crosstie valves.
Also, this alternative includes additional credit for improvements in amargency procedures for recovering from a LosW accident.
3.6 Alternative 6 -Provide Independent RCP Seal Coolino System This alternative provides an independent water supply and distribution system for backup cooling of the RCP seals in case of ESW loss.
Preventing an RCP seal failure and, hence, a small break LoCA would remove a substantial risk contributor associated with this issue.
This alternative is also a consideration in Generic Issue 23, " Reactor Coolant Pump Seal Failures." A l
proposed resolution for GI-23 has recently been reported (Ref. 7).
An objective of the proposed resolution of GI-23 is to 6
(
l I
I
I 1
reduce the probability of seal failure, thus making it a relatively small contributor to total core-damage frequency.
3.7 Alternative 7 Combine Alternatives 5 and 6 (TS Chances and Independent RCP Seal Coolino)
Under this alternative, a combination of two or more alternatives discussed above could result in greater risk reduction.
The combination of Alternatives 5 and 6, namely technical specifications (TS) changes regarding limits on taking equipment out of service during shutdown operations, cross-tie testing re quirement s, and procedures improvement combined with an independent RCP seal cooling system, could be expected to result in a more substantial CDF reduction and still be' cost-effective.
7 l
- - - - - - =
r 4.
TECHNICAL FINDINGS The BNL evaluation of failures of ESW system at multi-unit sites included a determination of the initiating frequency of loss of ESW system, core damage frequency due to ESW failure, dose consequence analysis and cost benefit analysis.
The detailed evaluation is found in NUREG/CR-5526.(Ref. 3).
4.1 Core Damage Frequency Analysis The core damage vulnerability caused by the failure of the ESW systam may be estimated by developing a full scale PRA model including initiating event frequency categories, event tree and fault tree analysis and incorporation of support system dependencies.
The PRA model was then appropriately modified to reflect various plant operating configurations to analyze the consequences of the loss of ESW function in each operating state as shown in Table 4.1.1.
To facilitate the present analysis, BNL selected an existing Byron Unit 1 PRA model (Ref 2.) which was previously developed and which examined the ESW system of a single unit (Byron Unit 2 was not operational at the time).
The Byron model was modified by BNL to include the effects of multi-unit configuration, and short term /long-term recovery actions.
Additionally, the probability of RC pump seal LOCA was established based on a more recent pump seal failure model as described in NUREG/CR-4550 (Ref. 8), and incorporated in the present analysis.
4.1.1 Initiating Event Frequency The initiating event frequency representing' the loss of ESW for multi-unit site operations was derived initially from operational experience for single unit PWR operations.
This LOSW initiating event frequency was then modified, ' to reflect multi-unit PWR sites. As the system configuration for various operating states may be different, the respective LOSW initiating event frequency for each of these operating states was determined separately.
An approximation method involving the combination of the experience data with an analytical _ technique was used.
A multi-unit ESW system fault tree was developed similar to the existing model of Byron Unit 1.
This modified model represents the unavailability of the second unit to supply ESW to'the first Unit, given the complete LOSW in the first unit.
The fault tree is'provided in Appendix D of Reference 3.
. Table 4.1.2 lists the initiating event frequency for each operating state.
This frequency was 8
r
,r.~
-.p s
m m
calculated on the basis of the operational experience reflected by the base initiator, and then multiplied by a modifier corresponding to the respective operating states of the two units derived from a fault three analysis (Ref. 3).
i 4.1.2 ESW System and RCP Seal LOCA Recovery The event tree established in Reference 2 indicated that the small LOCA due to RC pump seal failure and AFW system failure are the dominant accident sequences.
It was decided to use a more recent model for seal LOCA probability.
The RC pump seal failure probabilities are based on the model developed in Reference 8 which provides the probability of a seal leakage as a function of the leak rate and elapsed time after the loss of seal cooling.
A simplified recovery model was also developed by BNL in Reference 3 for the sequences relative to the failure of the ESW system.
The recovery model consists of a number of recovery factors which are established based on the particular failure mode and the time available.
Operating experience data bases regarding ESW systams consisting mostly of LER submittals were searched by BNL and, as also confirmed by NUREG-1275 (Ref. 4), the ESW systam failure duration has varied from less than 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> to a few days before EST system recove ry.
The data suggest that there are approximately three characteristic time periods of systam recovery.
The first time period involves ESW failures which may be recovered within one hour and consists of a large fraction of the ESW events (approximately 70% of the total).
The.second time period involving more problematical hardware or other failures, extends up to 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
About 90-96% of all events may be recovered in this time.
The last group of events are such that recovery may take a relatively long time and generally involve the most serious hardware problams.
It is estimated that by the end of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> only about 1% of the events were not recovered.
4.1.3 Relative Time Fractions Since the average time of operation varies with different operating configurations, it is necessary to estimate the relative time fractions for each operating mode.
The relative time fractions essentially represent the average length of time period of the specific multi-unit operating state coupled with the arrangament of the ESW Systams.
Maintenance or test-related outage time of ESW equipment must also be accounted for in the system's average time fraction.
The ESW flow requirement may be 9
i
satisfied through the unit crossties utilizing the ESW pumps of the other unit.
Based on discussions with utilities, it was assumed that the crossties are used about 10% of the time during the shutdown period.
1 The most dominant time fraction is that of the power operating arrangement, i.e.,
both units at power and one ESW train of each unit running with the other in standby.
4.1.4 Core Damage Frequency For each.of the operating states a conditional core damage probability (CDP) _ was calculated by renormalizing the-original base case with the respective configuration-dependent initiating frequency and weighting the state-dependent initiating event frequency.
The total CDP may be expressed as:
CDP = l A (state)
- P (Sequence)
- RT, i
3 i
Where 1, is the state-dependent initiating event frequency given that the unit is in this state for the full year, and RT, is the relative time fraction of the ith state while P is the ith i
sequence probability.
The dominant sequence conditional core damage probabilities are summarized in Table 4.1.1.
The sum of all the sequences.during power operation results in P (power operation) = 1.03E-01 which reflects the conditional probability of core' damage given a complete loss of ESW during power operations.
The corresponding.
value for shutdown is P (shutdown) = 2.82E-02.
The most dominant contributor for all sequences, including shutdown, is the RCP seal LOCA: P (Seal LOCA) = 6.8E-02, which is approximately'65% of-P (power operation).
The core damage frequencies due to various accident sequences are summarized in Tables 4.1.2, and 4.1.3.
The most dominant sequence is the RCP seal LOCA: CDF (Saal LOCA) = 8.BE-05 par reactor-year, which is about 60% of the total CDF due to ESW loss or 1.5E-04 pei reactor-year.
The total CDF due to loss of ESW (1.5E-04 per reactor-year) is judged to be substantial comps-ed to the total due to all causes (typically in a range of about 1.0E-4 to 2.0E-4 per reactor-year)..The next section presents the results of an examination of different alternatives which could lower this core damage fre quency.
10 I
s
4.1.5 Effects of Potential Improvements on Core Damage Fre quency The potential alternatives for improvements were initially selected in NUREG/CR-5526 (Ref. 3) by considering (a) the dominant failure modes of the ESW system (listed in Table 4.1.4) and (b) the dominant accicant sequences contributing to the relatively high CDF.
Since there is no single dominant failure mechanism represented in the initiating event frequency, a number of different options were considered including combinations of particular failure modes to reduce the initiating LOSW frequency.
The failure modes indicated in Table 4.1.4 are based on actual operating experience.
The base case initiating event frequency was modified to take into account the effects of the particular alternative under consideration.
First, the fraction of the initiating event frequency that could be improved by each alternative under consideration was determined using the data listed in Table 4.1.8.
Second, the relative change in the ESW systam reliability with and without the improvament provides an indication of the potential reduction in the core damage f re quency.
Fault tree analyses which included the logic modules and/or additional component failure rates that represent the proposed modification were employed to estimate the total system unavailability.
The reliability analyses of the improvaments were performed for each state or plant configuration, resulting in a calculation of configuration-dependent initiating event frequencies.
As noted in Section 3, the following potential improvaments were i
analyzed regarding their capability to provide a cost-effective reduction in risk due to a LOSW event:
i l
o Additional Crosstie - Reducing the possibility of 1
the malfunction of the cross-connection between l
units.
Electrical Power Cross-Connection-Increasing the o
redundancy of the electrical power supplies to ESW
{
pumps.
11
)
i
-o o
Separate Intake Structure or Bay with an i
Additional Swing ESW Pump - Increasing the redundancy of the ultimate heat sink or source of cooling and increasing the availability of the ESW~
pumps.
L Changing Technical Specification requirements and o
amargency procedures.
o Installation of an independent RCP seal cooling
- system, Combination of RCP seal cooling system and o
Technical Specifications / Procedures changes.
The first three alternatives were selected based on considerations regarding the ESW failure mechanisms observed in the PWR operating history data base.
A particular operating mode when both ESW pumps of the shut.down plant are inoperable (State IId and h) is a concern since there are no explicit Technical Specifications requiraments on the ESW system in this operating moda.
Therefore, the alternative of imposing additional TS requiraments was also analyzed regarding their effect on CDT reduction potential.
This alternative.also considers additional credit for unit crosstie testing and amargency procedures.
The most dominant contribution to the CDF arises from the failure of the RCP seal upon loss of seal cooling due to the unavailability of the ESW.
Therefore, the installation of an independent RCP seal cooling system which would cool the seals in the event of loss of ESW was also evaluated as a potential improvament.
The results are summarized in Table 4.1.5.
12 1
=
Table 4.1.1 Operational Status of Multi-Unit Sites Unit 1 Unit 2 Site's ESW Pump ESW Pump Status Unit 1 1
2 Unit 2 1
2 Ia OP R
AOT OP R
AOT Ib OP R
ACT OP R
SB Ic OP R
SB OP R
ACT Id OP R
SB OP R
SB iia OP R
ACT DN R
ACT iib OP R
AOT DN R
SB IIc OP R
ACT DN SB M IId OP R
AOT DN M
M IIe OP R
SB DN R
ACT IIf OP R
SB DN R
SB IIg OP R
SB DN M
M IIIa DN R
ACT OP R
AOT IIIb DN R
AOT OP R
SB IIIc DN R
SB OP R
AOT IIId DN R
SB OP R
SB IVa DN R
ACT DN R
ACT IVb DN R
ACT DN R
SB ivc DN R
AOT DN M
M IVe DN R
SB DN R
SB IVf DN R
SB DN R
AOT IVg DN R
SB DN M
M OP = Operating.
DN = Shutdown.
R = Pump running.
SB = Pump in standby.
AOT = Pump in test (allowable outage time).
M = Maintenance.
13 E
v
Table 4.1.2 State Dependent LOSW Initiating Event Frequencies ESW Unit Initiating States Event Unit 1 Unit 2 Frequency /
Pumps Pumps Reactor-Year I - Unit 1-Up/2-Up R/AOT R/AOT 1.6E-01 R/SB 1.4E-02 R/SB R/ACT 1.2E-02 R/SB 1.1E-03 II - Unit 1-0o/2 Down R/AOT R/ACT 1.2E-02 R/SB 1.1E-02 SB/M 1.4E-02 M/M 1.6E-01 R/SB R/AOT 9.7E-04 R/SB 8.9E-04 SB/M 1.1E-03 M/M 1.2E-02 III - Unit 1-Down/2-Up R/AOT R/AOT 2.3E-02 R/SB 2.1E-02 R/SB R/AOT 2.6E-02 R/SB 2.3E-03 IV - Unit 1-Down/2-Down R/AOT R/AOT 2.3E-02 R/SB 2.1E-02 SB/M 2.6E-02 M/M 2.9E-01 R/SB R/AOT 2.6E-03 R/SB 2.3E-03 SB/M 2.9E-03 M/M 3.2E-01 14
Table 4.1.3 Sequence Conditional Core Damage Probabilities Sequences Conditional Core Damage Probability Power Operations RCP Seal LOCA - P (Seal LOCA) 6.BE-02 Auxiliary Feedwater - P,
2.3E-02 u
Long Term ATK - P 9.lE-03 un Other Sequences - P 3.2E-03 a o, Total - P (Operation) 1.03E-01 Shutdown - P (Shutdown) 2.82E-02 15 I
Table 4.1.4 Core Damage Frequency Due to Individual Sequences Initiating Event Core Damage Frequency Sequence Frequency Sequences A*RT Probability-P CDF/R-YR Seal LOCA - P (SL) 1.3E-03 6.8E-02 8.8E-05 AFW - Pm 1.3E-03 2.3E-02 3.0E-05 Long Term - Pw, 1.3E-03 9.1E-03 1.2E-05 Other - P 1.3E-03 3.2E-03 4.2E-06
- m.,
Total Power Operation
- P (Power Operation) 1.3E-03 1.03E-01 1.3E-04 Shutdown - P (Shutdown) 7.1E-04 2.82E-02 2.0E-05 TOTAL 1.5E-04 i
16
I Table 4.1.5 Core Damage Frequency - Summary Initiating Sequence Core Damage States Event-Frequency Probability Frequency CDF/RYR A*RT P
I + II 1.30E-03 1.03E-01 1.3E-04 III + IV 7.1E-04 2.82E-02 2.0E-05 TOT 1J.,
1.SE-04 17 i
l Table 4.1.6 Tailure Mode Classification Relative Contribution Failure Mode to Initiating Frequency i
Intake structure unavailable 35%
Loss of electrical power supply 35%
)
Loss of ESW pumps 20%
Other log i
-18 1
o Table 4.1.7 CDP Reduction For Alternatives Alternative ACDF 1.
No Action N/A 2.
Additional Crosstie 1.60E-05 3.
Electrical Power Cross Connection 1.4E-05 4.
Separate Intake Structure 9.13E-05 5.
Technical Specifications Modifications and Procedures 2.55E-05 6.
Independent RCP Seal Cooling 7.82E-05 7
Coe.bination of Alt. 6 + Alt. 5 9.10E-05
/
/5//0
7 kfu,5
& x /O 'S,Dm 19
~
l
4.2 Dose Consequence Analysis For purposes of this study, consequences are measured in person-ram and benefits in person-rem averted.
Once the core damage frequency (CDF) and changes in CDF due to a potential resolution alternative have been calculated (Section 4.1), the next step is to calculate the corresponding consequences in person-ram, and hence, benefits in person-rem averted.
The reactor safety study (WASH-1400) first attempted to evaluate containment performance for a number of accident sequences. As part of that attempt a set of radioactive release parameters was developed corresponding to specific containment failure modes.
More recently, the NRC has documented in NUREG-ll50 a detailed assessment of the risk associated with five nuclear power plants.
This study (NUREG-ll50) represents the most updated analytical framework for the assessment of containment performance including source terms and off-site consequences.
It was decided to use NUREG-1150 as the basis for the evaluation of the seven two-unit sites of this issue. A more dotailed description of these calculations and their application to this study is given in Reference 3.
The consequence model specific to the Zion site was used as the starting point of the consequence assessment of the seven sites of this issue because of the availability of its detailed modeling and evaluation in the NUREG-ll50 effort.
The multi-unit sites evaluated in the GI-130 study would be expected to produce average consequences smaller than those calculated for the Zion site because of their location and respective population densities within their evacuation zones.
For this reason, adjustments were made to the Zion consequences as discussed in detail in Reference 3, and summarized in the following paragraph.
A comparison of the Zion-based results was made with those of the Surry and Sequoyah plants, and it was concluded that the consequences of an ESW induced core-damage at a large, dry containment plant, typical of the GI-130 plants, to be 47% of the total consequences for Zion, or 8.0E+06 person-ram.
It should be noted that this is for power operation only and without taking containment systems recovery into consideration.
When recovery actions are taken into consideration this number is modifica to 5.5E+06 person-ram.
A calculation of the consequences associated with shutdown operations was also performed.
While the use of power operation release categories for consequence calculations at shutdown may appear to overestimate consequences, Reference 3 indicates that the person-rem consequences are relatively insensitive to the 20
e
~
source term.
This is because of interdiction criteria and because of the relatively high contribution of long-lived isotopes to the long term dose.
The total consequences for shutdown operations were calculated in NUREG/CR-5526-(Ref. 3)-to be 3.1E+06 person-ram.
Hence, the overall benefit for each alternative considered in terms of averted consequences in person-rem may be estimated by multiplying the power consequences with the power ACDF and the shutdown consequences with the shutdown ACDF, adding the two products and multiplying by 30 years, the assumed lifetime of the average GI-130 plant.
Hence:
Total Benefit = 30 X (ACDF X 5.5E+06 + ACDF, X y,
3.1.E+0 6).
Table 4.2.1 shows the benefits (or consequences reduction) in person-ram that was calculated for each proposed alternative.
21 k
Table 4.2.1 Benefits of Proposed Alternatives (Person-Ram)
Alternative Low Best High Estimate Estimate Estimate 1.
No Action 2.
Additional Crosstie 739 2,635 4,951 3.
Electrical Cross-Connection 645 2,349 4,467 4.
Separate Intake Structure 3,992 14,324 27,004 5.
Technical Specifications Modifications 1,150 4,141 7,825 6.
Independent RCP Seal Cooling 3,510 12,870 24,570 7.
Conbination of Alternativas 6 and 5 4,063 14,821 28,211 22
43 Cost Analysis To calculate costs for the various alternative backfits, several sources were consulted (Ref. 3).
Some cost estimates were __
derived from an NRC-sponsored research report (Ref. 9).
Another source was the computer printout for the Energy Economic Data Base (EEDB) and supporting documents (Ref. 10).
Still another source was various discussions with utilities.
An initial overall assumption was that the backfits can be accomplished outside of the critical path.
Consultation with utility personnel confirmed that this should be possible.
Otherwise, the direct costs will rise substantially, at the rate of $400K for each day that replacement power is needed.
For each resolution alternative, the costs noted in Subsections 4.3.1-4.3.7 were considered.
4.3.1 Direct Costs This cost category includas factory purchases, installation and ensite labor and materials, but excludes indirect costs (e.g.,
engineering, administrative, etc.).
It is given in the first column of Table 4.3.1 as a best estimate.
Table 4.3.2 shows the best estimate and the range of estimates in the direct cost.
Alternative 5 (techn. cal specification modifications including procedures and crosstie testing) shows a zero in the direct cost because this itam was already included in Column 4 (technical specification costs) of Table 4.3.1.
4.3.2 Indirect Costs The indirect costs are usually a certain fraction of the direct cost.
As recommended in NUREG/CR-4627 (Ref. 9), 30% was used (the range is frem 25% to 33% for engineering and quality assurance costs for in-place structures).
Column 2 of Table 4.3.1 includes this cost component.
4.3.3 Operatine and Maintenance Costs Usually, these costs annually equal 3% of total " overnight" costs.
Overnight costs represent the sum of total direct and indirect costs assuming that the modification was completed
)
overnight (e.g., excluding the time costs of capital).
To arrive at the total operating and maintenance (O&M) cost, the annual value was integrated and discounted over the remaining plant life
{
23 I
,. - ~
~
4 (30 years).
Alternative 5 (modify Technical Specifications) was assumed not to involve any O&M costs.
Column 3 of Table 4.3.1 includes this cost component.
In calculating OsM costs, a 5 per cent discount rate was assumed, consistent with the NRC
.9 commended practice.
4.3.4 Technical Specifiestions Costs Each alternative involves modifying technical specifications to a certain extent. According to NUREG/CR-4627 (Ref. 9), these costs are $1BK per reactor for a simple-case and $35K per reactor for a complicated or controversial one..It was assumed that each alternative will result in a simple technical specification change.
Neither choice includes the cost of a public hearing.
The fourth column of costs in Table 4.3.1 includes this component of cost.
4.3.5 NRC Costs NRC costs include the development and implementation costs.
The development costs should be about $11K/ reactor for a simple case and $21K/ reactor for a complicated one.
Neither case includes the cost of a public hearing.
The former figure was chosen here.
Operating costs would be incurred after the resolution's implementation and they would cover ensuring compliance with the new requiraments. The operating costs have to be integrated and discounted, since they are recurring.
The implementation and operating costs were estimated at $50K per reactor.
Total NRC costs would then be $11K + 50K = $61K per reactor.
Celumn 5 of Table 4.3.1 includes the NRC costs.
For a technical specification and procedures change, the total NRC costs would be
$21K per reactor (Ref. 9).
4.3. 6 Averted Onsite Costs Averted onsite costs are taken into account as cost offsets (Table 4.3.3) to the calculated cost of the proposed resolution alternatives, consistent with NRC policy.
Table 4.3.4 lists the averted consequences.
It can be seen that the onsite personnel
't exposure per accident will be low, compared to the offsite exposure, and other onsita consequences, so this component was noc considered further.
The numbers are from NUREG/CR-3568 (Ref. 11) as best estimate numbers.
Averted onsite exposure would be added to the offsite person-rem exposure as part of the benefits, but the effect is negligibly small.
For cleanup and replacement power, the integrated and discounted costs is then
.24 t
t 9
t
cultiplied by the 3CDF to arrive at the offset cost of each alternative.
The calculated as follows: cleanup and replacement power costs were u=
(C, + C,)
1_
( 1 - e '"')
(1 - e -")
(Ref. 11).
r*
where: u = integrated and discounted cost C. = cost of cleanup ($100M/yr)
C, = cost of replacement power ($400K/ day) r = discount rate (0.05/yr) at = ramaining plant life (30 yr m = duration of cleanup / power re) placement (10 yr)
Table 4.3.1 shows components of the total cost and th for the best estimate case e not cost cost is the total cost minus the cost offset (the costs are per reactor).
The not If the net cost is negative, (from Table 4.3.3).
regardless of the cost benefit ratio.the alternative is cost-beneficial column and includes an additional indicated cost c evious
- instance, and the indirect costs of an alternative. column " inclu For 4.3.7 Range of cost Estimates Table 4.3.5 presents the range of estimates obtained f total cost (corresponding to Column 6 of Table 4.3 1)(correspond or the cost and the not various cost componentswere calculated by taking the lowest estimates in the The low values a a of (mainly direct costs) co=putation through to the final number.
and carrying the calculated by taking the highest estimates in the data of thThe high value var 3ous cost components and carrying the computation thou h t e
the final number.
g o
25 I
i l
i Table 4.3.1 Best Estimate Costs of Proposed Alternatives,
($ Per Reactor)
Colur.n Numbe r 1
2 3
4 5
6 Include Include Onsite Include NRC Conseq.
Include Include Tech.
Cost-Offset Direct Indirect O&M Spec.
Total Net Alternatives Cost Cost Cost Cost Cost Cost 1.
No Action 2.
Additional Crosstie
$557K
$724K
$1.05M
$1.08M
$1.14M
$627K 3.
Electrical Cross-Connection
$50K
$65K
$94K
$128K
$189K -$246K 4
Separate Intake St ructure
$29M
$38M
$55M
$55M
$55.1M $52.3M 5.
Technical Spec.
Modifications
$0
$0
$0
$83K
$104K -$684K 6.
High Pressure Pump for RCP Seals
$5.9M
$7.7M
$11M
$11M
$11.1M
$8.8M 6a. Firewater for Thermal Barrier Cooling
$200K
$260K
$378K
$412K
$473K
-$1.9M 26 F
Table 4.3.2 Direct Cost Estimates ($ Per Reactor)
Alternatives Low Best Estimate Estimate High Est4= te 1.
No Action 2.
Additional Crosstie 250K 3.
Electrical Cross-Connection 557K 50K 1M 4.
Separate Intake Structure 50K 7M 50K 5.
Technical Specifications 29M 3BM Modifications (see text) 0 6.
High Pressure Pump for RCP 0
0 Seals 1M 6a.
Firewater for Thermal Barrier 5.9M 15M Cooling 127K 200K 273K l
i i
27 1
i
Table 4.3.3 Cost Offsets for Proposed Alternatives ($ per Reactor)
Alternatives Cost Offset ($)
1.
No Action 2.
Additional Crosstie 3.
Electrical Cross-Connection 513K 4.
Separate Intake Structure 435K 5.
2.75M Technical Specifications Modifications 6.
788K Independent RCP Seal Cooling 7.
Combination of Alternatives 5 &
2.34M 6
2.73M 28 P
Table 4.3.4 Onsite Consequences TYPE Amount Occupational Domes:
-Immediate:
1,000 Person-Ram
-Long Term:
20,000 Person-Ram Total 21,000 Person-Rem x 30 yr x $1,000/p-ram = $6.3E+08 yr Replacement Power Cleanup
$1.8E+10 yr
$1.2E+10 yr Total Onsite Consequences
$3.0E+10 yra
- This number to be multiplied by A CDF for each alternative 9
29 I
O Table 4.3.5 Range of Estimates for the Total Cost and the Not Cost ($)
Total Cost Net Cost Alternatives Iow Best High Low Best High Estimate Estimace Estimate Estimate Est4==te Est4=?te 1.
No Action 2.
Additional Crosatie 550K 1.14M 2M 37K
.627K 1.5M 3.
Electrical Crosh-Connection 173K 189K 205K
-262K
-246K
-230K 4
Separate Intake Structure 14M 55.1M 72M 11M 52.3M 69M 5.
Technical Specificat, ions Modifications 40K 104K 171K
-740K
-684K
-617K 6.
High Pressure Pump for RCP Seal Cooling 2M 11.1M 29M 1.2M 8.8M 28.2M 6a. Firewater for Thermal Barrier Cooling 318K 473K 624K
~2M
-1.9M
-1.7M 30 8
f e
5.
VALUE/ IMPACT ANALYSIS The value/ impact (v/I) alternatives examined under this study is based on themethodology in a requiraments of the backfit rule (10 CFR Part 50.109) and related implementing guidance contained in References 11, 13.
cost / benefit ratios for each alternative evaluated in term 127~and cost in $ per person-ram averted, guideline such as $1,000/ person-ram.which may be compared to a Deterministic considerations on the merits ef a prop s.
alternative resolution are also a part of the decision with respect to a given alternative (Chapter 6).
In the following value/ impact assessment are presented. sections a description of e results of this assessment Table 5.1 summarizes the for tha various alternatives analyzed.
5.1 Alternative 1 - No Action Under this alternative there would be no new regulatory requiraments.
alternative does not preclude a licensee, Consistent with existing regula this operating license, or an applicant for an intended to enhance the reliLbility/ operability of the Essentialf Service Water Systam and its components on a plant-specific basis.
the various alternative analyzed. Table 5.1 summarizes the results of th 5.2 Alternative 2 - Install Additional Crosstie The ESW systems of the seven multi-unit sites analyzed under GI-130 are cross-connected through pipe connections and isolation valves.
This arrangement allows the operator of one unit to utilize the ESW cooling capacity of the other unit.
In most A hardware failure to open the isolation valves,the crosstie isol
- cases, arise, could result in adverse conditions.
should the need 1
i A parallel cross-and in addition would allow for more flexible maintenanceco options.
The effects of the isolation valve failures on the CDF were not large due to the relatively low observed isolation valve failure ratos indicating that other hardware components are more significant in reducing the overall system unavailability.
core damage frequency reduction of this alternative was estimated The to be 1. 6 E-05/RY, 31 l
i
The cost-benefit
$433/ person-rem, ratio for this alternative was calculated to be onsite costs.
or $238/ person-rem taking into acccunt averted 5.3 Alternative 3 - Provide Electrical Power Cross-Connection One of the observed contributors to the unavailability of th supply and control systam. system is related to the reliability of the electrica e ESW Reference 3, Based on the data reported in various causes was relatively high;the loss of the electrical power supp however, the recovery times associated with these events indicate a relacively faster average recovery observed during losses of the ESW system.
In general, the electrical power supplies to the ESW trains are separated and have no cross-connection capability, ESW pu=p cannot be powered from electrical Train B. i.e.,
Train A alternative therefore investigated the implamentation of a cross-This to the operation of the two ESW pumpsconnection between the e (Trains A and B).
The cross-connection of electrical power supply of other electrical components, such as HOVs was not considered as part of this alternative because of their less significant potential to risk contribution as observed in the operational data.
envisioned that It is exclusively manual operation.the electrical power cross-connection would be an
- However, adverse interactions between electrical trains A and B,the possibility of the inadvertent transfer of faults from one train to the othe such as and hence, the loss of both trains, make this alternative of r,
questionable value.
Even if this contribution of possible adverse interactions between trains is set aside, observed during losses of electrical pcwer. reduction is not sig the CDF ratio without taking into account the potential adverseThe cost-benefit interactions for this alternative was calculated to be
$80/ person-rem, and, the not cost becomes negative,if the averted onsite costs are taken
- account, i.e.
resulting in a net savings.
5.4 Alternative 4 - Provide Separate Intake Structure A review of the failure modes of the intake structure indicates that one of the observed ESW failure mechanisms is the failure certain intake components (such as travelling screens or strainers).
This type of failure within the intake structure produces a general stopping or restricting the flow of cooling water to the plant.
A separate intake structure, either located 32 I
i
ll would make a backup cooling capability availabison the same bod t
j The intake structure is usually a single structure divided int separate bays by concrete walls.
o installed to prevent the intake blockage by large foreignThere are a numbe objects.
The collat s or plugging of these screens may occur as a common mode fai3',
The whole.ntaka structure coulci; due to the common inlet and/or commo source.
events such as flooding or freezing.
also be affected by The alternative considered here is a completely separate i t k structure serving as a redundant intake source of ESW.
nae located on the same water source, It may be but on a physically separate location.
An alternate design, which would provide additional independence / diversity, would be to install the additional int k s
structure on a physically separate water source.
ae there are sites where this would not be feasible. Naturally, The separate intake structure alternative includes the structu screens and the associated motors, valves and piping.
re, ISW pump would be made available to either unit with redund A swing electrical power supplies.
ant This arrangement is intended to electrical supply failures, reduce the probability of two failure mechanisms; o failures of the ESW pumps.
and the other involving operating The additional ESW pump would be a swing pump serving either unit depending on.the current needs of both units.
additional swing pump with redundant electrical power supp would affect a large fraction of the initiating event frequen y
pumps, and their power supplies.
The calculated reduction in CDF associated with this alternati was 9.13E-05/RY.
The respective cost-benefit ratio was ve calculated to be
$3,847/ person-rem, into account averted onsite costs.
and $3,651/ person-ram taking 5.5 Alternative 5 - Modify Technical Specifications Reavirements There are certain operating modes, Modes 5 and 6 (shutdown and l
specific requirements in the Technical Specificationsrefueling m i
l-the status of its ESW pumps is uncertain.these operating modes the (TS).
In I
l that any of the ESW pumps be operational in these modesThe TS do not require implicit requirement is imposed on the ESW trains through the An 33 I
I
r.
erplicit requirement to operate the RER system to remove decay heat.
In esrence, the operator of the i
unit in shutdown may utilize the unit's own ESW pumps to provide the necessary heat ramoval fuaction, but may just as well decide to use the unit crossties N
to supply ESW flow from the other unit.
requirements.on tha ESW pumps, In the absence of any made inoperable at the s ame time.both pumps could be maintained or Although this is not a universal practice, certain modeling assumptions were made based on information gathered from plant sites representing a more conventional practice involving the administrative control of crobatie use, sad. the ESW pump maintenance schedule.
In the basic analyticel model it was assumed that the simultaneous shutdown of both ESW pumps could occur only randomly.
The unavailability of the Unit 2 ESW pumps to provide backup for the Unit 1 ESW sy$ tam may be reduced by imposing an explicit operability requirement on at Unit 2 while the latter is in Modes 5 and 6.least one of the ESW pumps of An additional improvement is the testing of the unit-to-unit crosatie valves to provide greater assurance of operability.
Also, this alternative includes additional credit for improvements in amargency-procedures for-recovering from a LOSW accident.
The resulting CDF calculations indicated that the CDF would be reduced by 2.55E-05/RY.
The respective cost-benefit ratio for this alternative was determined to be $25/ person-rom, and, if the averted onsite costs are taken into account, the net cost becomes
- negative, i.e. resulting in a net cost savings.
5.6 Alternative 6 - Provide Independent RCP Seal Coolino System The technical findings reported in Chapter 4 and Reference 3 indicate that the major contributor to the ESW-related component of CDF comes from the failure of the RCP seals.following a loss of ESW.
Specifically, the RCP seal LOCA sequence contributes about 60% of the total CDF attributable to ESW failures.
if the likelihood of a RCP seal induced LOCA may be reduced,
- Hence, proportionately significant reduction in CDF may be achieved.
a This alternative provides for a dedicated seal cooling system that would continue to provide heat removal capability after a loss-of-ESW event.
The cooling requiraments of the RCP seals are relatively small, capable of delivering about 50-100 gpm was judged to beand a single sufficient.
The pump may be driven either by an electric motor for electrical independence from the point of view of other o r, i
l 34 t
i
e accident scenarios (such as station blackout), a diesel-driven pump option may also be considered.
The single high pressure pump and diesel would provide flow via the cooling header to the four injection lines (one to each RCP seal).
It was assumed that the pump and diesel would not require auxiliary cooling for the lube oil, bearings, etc., as the suction flow or air cooling would be sufficient to provide all their heat ramoval requiraments.
It was also assumed that the return flow from the RCP seals would not be recycled.
In other
- words, a once-through cooling cycle would be used with a sufficient water supply to last about 8-10 hours.
It is assumed that a dedicated tank will be installed, with a capacity satisfying 8-10 hours of seal cooling.
After this time, added cooling could be provided by other available water supplies, such as the refueling water storage tank.
In modeling the systam, the following assumptions were made:
1.
single high pressure pump, 50-100 gpm capacity, 2.
dedicated water storage tank with capacity to last at least 8-10 hours, 3.
ac-independent (non-seismic) diesel-driven pump, 4.
no support system cooling required, and 5.
once-through RCP seal heat ramoval.
Other design alternatives may also be considered utilizing arrangaments different from that of the high pressure pu=p injection.
One less costly alternative would provide flow through the RCP thermal barrier heat exchangers by connecting the firewater system into the CCW lines.
Most firewater systems have one diesel-driven firewater pump which usually is independent of the ESW system.
i The CDT reduction for this alternative involving a high pressure seal cooling system was calculated to be 7.82E-05.
The respective cost-benefit ratio for this alternative involving a high pressure seal cooling system was calculated to be SB62/ person-rem, or $684/ person-rom if the averted onsite costa 4
are taken into account.
The cost-benefit ratio for this j
alternative involving a connection to the fire water system for thermal barrier cooling was calculated to be $37/ person-rem, or, 35
if the averted onsite costs were taken into account, this alternative would result in a not cost savings.
5.7 Alternative 7-Combine Alternatives 5 and 6 (TS Chances and Independent RCP Seal Coolino)
As shown in Table 5.1, most of the analyzed alternatives have favorable cost-benefit ratios (presented as $/ person-rem).
In these cost-benefit calculations, it was assumed that each of the alternatives (1 through 6a) was utilized individually and independently from the other alternatives.
For the combination case, the CDF reduction is calculated when two alternatives are combined and utilized together to reduce the risk due to tha loss of ESW function.
The alternative with the highest ACDF and favorable cost / benefit ratio was ranked first and served as the starting basis point.
This was Alternative 6 (or 6a), the dedicated cooling system for the RCP seals.
When the next alternative was considered, the CDF reduction was calculated fro = the case where Alternative 6 (or 6a) was already incorporated.
The combined CDF reduction resulting from the implamentation of alternatives 5 and 6 was calculated to be 4
9.12 X 10 /RY, and the respective cost-benefit ratio of
$756/ person-ram, or $574/ person-ram with the averted onsite costs taken into account with a RCP seal cooling systam involving a high pressure cooling systam.
The cost-benefit ratio for this combination of alternatives with a RCP thermal barrier cooling system utilizing the fire water supply was calculated to be
$39/ person-ram, and if the averted onsite cost were taken into consideration a not gain would be achieved (i.e.,
a negative cost of implamentation).
5.8 Unce rtainty_Analys i s This section discusses, the sources and treatment of uncertainty for the GI-130 study.
Uncertainty is expressed as a quantitative bounding of the mean value.
Uncertainty arises due to the selection of the data base used to determine parameter values, modeling assumptions, and completeness of the analysis.
Although a complete analysis of all data uncertainties was not conducted, uncertainty studies were performed on selected issues that were important to the results.
Uncertainty data were gathered, evaluated, and reported in the form of distributions for these selected issues.
This data-gathering and reduction was used to gauge the effects of the individual data uncertainty _ on the final core damage frequency results of the analysis.
36
4 The primary areas of uncertainty exist in the determination of the initiating frequency values, modelling and data uncertainties.
Each of these particular areas were addressed and the final result combines these issues to present the uncertainty of the core damage frequency.
All other parameters were treated as point-estimates.
The results of the uncertainty analysis show a mean value of CDF due to LOSW of 1.49E-4 per reactor-year, with a value of 5% and 95% of 3.99E-5/RY and 3.73E-04/RY, respectively.
5.9 Life Extension Considerations The regulatory process by which license renewal may be accomplished is currently under development by the NRC.
It is envisioned that a license renewal for an additional term of 20 years may be achievable based on satisfying specific requirements still to be established.
Hence, for considerations regarding the effect of license renewal on the results of the evaluation of GI-130, a reanalysis of the cost-benefit ratio parameters for each backfit alternative was performed.
The results of this reanalysis show that the benefits will increase by factor of 1.67, while the costs, both incurred and averted will increase by a factor of about 1.2 for most of the backfit alternatives analyzed.
Table 5.2 summarizes the cost-benefit ratios based on a license renewal of 20 years or a remaining plant life of 50 years.
A comparison of these numbers with those listed in Table 5.1 shows that the cost-benefit ratios for all analyzed backfit alternatives are considerably lower for extended plant life of 50 years vis a vis a plant life of 30 years, corresponding to licenses in force currently.
Even though all alternatives listed in Tables 5.1 and 5.2 become more cost-effective with life extension, Alternative No.
4, Separate Intake Structure, still remains appreciably higher than the $1,000/ person-rem guideline at a cost-benefit ratio of
$2,285/ person-rem.
37 I
i Table 5.1 Best Estimate Coat-Benefit Ratios ($/ Person-Ram)
Alternatives Total Cost /Banafit Net Cost / Benefit 1.
No Action 2.
Additional Crosstie 433 238 3.
Electrical Cross-connection 80 4.
Separate Intake Structure 3847 3651 5.
Technical Specifications Modifications 25 6.
High Pressure RCP Seal Cooling 862 684 6a. Firewater for Thermal Barrier Cooling 37 7.
Combination of 6 and 5 756 574 7a. Cochination of 6a and 5 39 1
- Including averted onsite costs re-ults in a not cost savings.
38 I
Table 5.2 Best Estimate Cost-Benefit Ratios ($/ Person-Rem) for 20 year License Renewal llternatives Total Cost /Banafit Net Cost /Banefit 1.
No Action 2.
Additional cromatie 271 133 3.
Electrical Cross-Connection 50 4.
Separate Intake Structure 2421 2285 5.
Technical Specifications Modifications 16 6.
Eigh Presstre RCP Seal Cooling 541 412 6a. Firewater for Thermal Barrier Cooling 23 7.
Combination of 6 and 5 474 343 7a. Combination of 6a and 5 24
- Including averted onsite costs results in a net cost savings.
l 1
i 39 f
6.
DECISION PATIONALE This generic issue was identified as a consequence of the Byron Unit 1 evaluation with respect to its vulnerability to core-dacage sequences in the absence of a crosatie from the ESW of Unit 2.
This configuration existed because Unit 2 was under construction, and was eventually supplamented by the crosstie between units.
There are fourteen units at seven sites having two service water pumps per unit (one per train) with a sharing of one pump between units via a crosatie between them, in a similar manner as currently in the two Byron units.
It was decided to focus the attention of this study on these seven two-unit sites because the design of their ESW system was expected to show the most vulnerable configuration to risk-significant sequences.
The remaining LWRs will be evaluated under GI-153,
" Loss of Essential Service Water in LWRs. "
As discussed in Chapter 5, most of the alternatives for reducing the risk associated with this issue would be cost-effective in meeting the $1,000/ person-rem guideline.
Furthermore, the objective of the GI-130 resolution is that the risk contributions from loss of the ESW system be reduced consistent with the backfit rule's two basic requirements that the improvament be both a substantial increase in protection, and be cost-effective.
A combination of potential improvements consisting of the installation of a dedicated RCP seal cooling system, and improvements in Technical Specifications with respect to ESW system operation, including crosstie testing and improvements in procedures, was shown to be capable of reducing the total CDF by 60% (to 6.1E-05/RY) in a cost-effective manner.
Hence, this is deemed to meet 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.
From the prevention perspective of a LOSW, it would be desirable to choose those alternatives which could reduce the number of occurrences of the LOSW initiators.
From the mitigation perspective, it would be desirable to choose those alternatives which would help to reduce the consequences of a LOSW.
The proposed resolution (Alternative 7) was selected to achieve some balancing of both these views; that is, the improvaments in technical specifications would assist on the prevention side, while the improved emergency procedures and 40 I
backup seal cooling would provide a blend of both prevention and mitigation capabilities.
The BNL analysis (Ref. 3) shows that after implementation of Alternative 7 there ramains a residual component of CDF of 6.lE-05/RY due to ESW loss which, on the face of it, would tend
~
to indicate the need for additional risk reduction.
We have reviewed this aspect of our evaluation of GI-130 and have concluded that additional improvaments beyond Alternative 7 cannot be justified at this time based on the following considerations:
1.
When the possibility of additional corrective measures (beyond Alternative 7) was considered, the resulting reduction in CDF was either too small (i.e.,
approached diminishing returns), or the cost / benefit ratio too high to be consistent with the backfit rule.
The examination for added corrective measures focused on those systems which are dependent on ESW, and which performed a role in several of the more dominant event sequences.
For example, the alternative of including a recommendation for a design change to make the Auxiliary Feedwater System (ATWS) independent of ESW cooling did produce a modest CDF reduction (CDF was reduced from 6.1E-05/RY to 4.8,E-05/RY).
Even further reduction is theoretically possible by ramoving dependence on ESW of each system and component, one-by-one until virtually complete independence is achieved; this is the ideal maximum reduction in vulnerability to LOSW; however it is judged that going further in this generic, representative plant calculation is pressing the limits of precision beyond what is warranted for plant-specific application to these 14 units.
In addition, such an alternative (ATWS upgrade), would be applicable only to some of the plant sites evaluated under this issue; three of the seven sites are k.nown to have already AFW systams independent of ESW cooling.
In another case, Alternative 4, involving the installation of a separate intake structure and a swing pump to be shared by the two units, was determined to be capable of providing a substantial risk reduction, but was estimated to be not cost-effective.
41 l
Q C
2.
As part of the implementation phase of resolving this issue, we recommend that the licensees /
applicants of the fourteen plants evaluated under GI-130 perform a review of their respective plant-specific designs vis-a-vis the recommendations of Alternative 7, (combination of Alternatives 5 and 6 as discussed earlier in this chapter and in Chapter 5) and report, pursuant to 10 CFR 50.54 (f), whether and how these recommendations would be implemented.
This licensee / applicant effort would take into consideration the existing plant-specific design features, which, in some cases, would be different from those assumed in the generic model used in the evaluation of this issue.
Hence, as a result of this effort, it is expected that individval licensees / applicants will submit a description of the measures taken as a result of the resolution of this generic issue, considering producing at least a comparable CDF reduction as has been calculated for the Alternative 7 combination in the GI-130 generic calculations.
The results of some plant-specific PRA evaluations reported by EPRI in Reference 14 supports the view that plant-specific designs incorporating features recommended by the resolution of this generic issue would recult in significant reductions in CDF.
For some p?. ants, the licensee or applicant may find it desirable or necessary to propose other design features, such as providing ATWS cooling independent of ESW, to improve on the assurance that the risk due to loss of ESW will result in a small fraction of the total risk for their individual plants.
3.
A number of generic safety issues related to GI-130 have been in various stages of resolution, including some that have already been resolved.
Their impact on 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 evaluation i
of GI-23 has been completed and a I
o proposed resolution has been reported (Ref. 7).
An objective of the proposed resolution of GI-23 is to reduce the risk of severe accidents associated with RCP seal failure by reducing the probability of seal failure, thus making it a relatively small contributor to total core-damsge frequency.
The proposed means of doing so entail the installation of a separate and independent cooling system for the RCP seals.
Hence, implamentation of the proposed GI-23 resolution could provide a substantial portion of the proposed GI-130 resolution.
As such, the proposed resolution of GI-130 will be coordinated with the resolution of GI-23.
o GI-51, " Improving the Reliability of Open-Cycle Service Water Systems" - The resolution of this generic safety issue has been reported in August 1989 (Ref.
- 15) and its implementation began with the issuance of Generic Letter 89-13 (Ref. 16), and Supp1 ament 1 (Ref. 17).
The GI-51 implamentation entails the implamentation of a series of surveillance, control and test recommendations to ensure that the ESW systems of all nuclear power plants meet applicable licensing guidelines.
During the review of the operational experience data for GI-130, credit was taken for corrective measures as a result of the GI-51 resolution by excluding those events that involved fouling of the ESW (sediment, biofouling, corrosion, etc.).
Hence, there is no direct impact of GI-51 on GI-130.
o GI-153, " Loss of Essential Service Water in LWRs" is under prioritization review and expected to be assigned NRC staff resources (Ref. 18) for its resolution.
Its purpose is to assess this issue for all LWRs not already covered by GI-130.
)
Insights gained by the evaluation of i
43 e
o o
generic safety issue 153 are expected to be useful in confirming and/or supplementing the technical findings of GI-130.
On the basis of the considerations discussed in Items 1-3 above and the technical findings of this study, including the value/ impact analysis of Chapter 5, it is concluded that the combination of Alternatives 5 and 6, namely, the augmentation of technical specifications and procedures along with the installation of an independent RCP seal cooling backup system are the appropriate risk reduction measures that are recommended.
These measures provide a substantial increase in overall protection of the public health and safety, and are cost-effective.
Of interest to the decision process on this generic issue are the insights and views available in related PRA documentation in the open literature.
Although still not finalized, the preliminary PRA work available in NUREG-1150, " Severe Accidents Risks: An Assessment for Five U.S. Nuclear Power Plants" (plus supporting documentation) is a source of extensive risk analyses information one might turn to for an understanding of ESW vulnerabilities.
An examination of the NUREG-ll50 documentation of the three PWRs that were studied indicates that the analyst considered that the ESW redundancy for two of the three PWRs was large enough that a complete loss of ESW as an event-initiator was deamed not credible (eight pumps available in Sequoyah,, Unit 1).
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 risk to be substantial (approximately 1. 5E-4/RY).
This contribution for
" ion was approximately 42% of the total core damage frequency due to all causes.
Another PRA work available in the open literature is NEAC--148,
" Service Water Systems and Nuclear Plant Safety," dated May 1990.
Although it 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 systam's contribution to plant risk has moved the industry to initiate a program to improve service water performance.
The limited guidanco available in NSAC-148 is a step in the right direction.
The wide range of core damage frequencies (due to LOSW) over the six plants studied suggests large variability in plant-specific KSW configurations.
The average CDF due to LOSW for the six plants was 6.55E-05/RY, with a range of 2.33E-04/RY-to " negligible" l
44 i
f
o o
contribution.
While many details of these six PRAs are not included in NSAC-148 and, therefore, must be considered to be used only with a great caution, the overall message that the service water system provides an important safety function which could be a substantial contributor to overall plant risk tends to lend added credence to the GI-130 conclusions.
45 i
d 7.
IMPLDENTATION The staff proposes to implement the resolution of Generic Issue
--130 by issuing a generic letter, under 10 CFR 50.54 (f), to the licenses and applicants of the fourteen plants involved in this evaluation.
The con. tent of the generic letter will address both the preventive and mitigative aspects of the proposed resolution s
as discussed in Chapter 6.
The implementation phase of Generic Issue 130 will be closely coordinated with that of Generic Issue 23, which deals with the RCP seal reliability for both normal operation and accident conditions.
i i
j 46 I
a
8.
REFERINCES 1.
Hemorandum from T.
P.
Speis to H.
L. Thompson, " Safety Evaluation Report Related to the LCO Relaxation Program _for the Byron Generating Station," January 15, 1986.
2.
- Cho, N.
Z.
et al.,
" Analysis of Allowed Outage Times at the Byron Generating Station," NUREG/CR-4404, June 1986.
3.
P. Kohut, et al.,
" Analysis of Risk Reduction Heasures Applied to Shared Essential Service Water Systems at Multi-unit Sites," NUREG/CR-5526, BNL, June 1990.
4.
- Lam, P.
et al., " Operating Experience Feedback Report -
Service Water System Failure and Degradation", NUREG-1275, Volume 3, Novamber 1988.
5.
U.S. Nuclear Regulatory Commission Report, 10 CFR Part 50,
" Safety Goals for the Operations of Nuclear Power Plant,"
Policy Statament, dated August, 1986.
6.
Hamorandum from Edward L. Jordan (CRGR) to Eric S. Beckjord (RES) " Implementation of the Safety Goals,"
September 6, 1990.
7 Hamorandum from C.
J.
Heltames, Jr. to.F.
P.
Gillespie, et al., " Proposed Resolution of GI-23, " Reactor Coolant Pump Seal Failures," June 20, 1990.
8.
" Analysis of Core Damage Frequency from Internal Events:
Expert Judgement Elicitation," NUREG/CR-4550, Volume 2, April 1989.
9.
E.
Claiborne et al.,
" Generic Cost Estimates," NURIG/CR-4627, Rev.
1, Science Engineering Associates, February 1989.
10.
" Complete CONCISE Printout for Model 148-PWR-HI (A3-ME)
Energy Economic Date Base (EEDB) Phase IX Update," volume 2 of 9, Plant Set 2 of 3, 7697.900, DOE Report 871102, (Philadelphia, PA; Oak Ridge, TN; United Engineers and Const ructors, Inc.), 1987.
11.
S. W.
Heaberlin et al.,
"A Handbook for Value-Impact Assessment, NUREG/CR-3568, PNL, Decamber 1983.
47 l
12.
Memorandum from E.
S.
Beckjord to Distribution, RIS Office Letter No.
2, " Procedures for Obtaining Regulatory Impact Analysis Review and Support," November 18, 1988.
13.
Memorandum from E.
S.
Beckjord to Distribution, RES Office Letter No. 3, " Procedure and Guidance for the Resolution of Generic Issues", May 10, 1988.
14.
" Service Water Systams and Nuclear Plant Safety" NSAC/148, Ma y 1990, prepared by Pickard, Lowe and Garrick, Inc. for Nuclear Safety Analysis Center and Electric Power Research' Institute.
15.
Mamorandum from E.
S. Beckjord to J. H.
Taylor, " Closeout of GI-51, " Improving the Reliability of Open-Cycle Service Water Systam'", August 10, 1989, 16.
Generic Letter 89-13, " Service water System Problams Affecting Safety-Related Equipment", July 18, 1989.
17.
Generic Letter 89-13, Supplement 1,
" Service Water System problems Affecting Safety-Related Equipment", April 4, 1990.
18.
Memorandum from K.
Kniel to C.
E. Ader, " Request for Prioritization of New Generic Safety Issue ' Loss of Essential Service Water in LWRs', " May 2, 1990.
48 e