NRC Generic Letter 1991-13

NRC Generic Letter 1991-013: Request for Information Related to Resolution of Generic Issue 130, Essential Service Water System Failures at Multi-Unit Sites Pursuant to 10 CFR 50.54(f)

ML031140524
Person / Time
Site:	Beaver Valley, Millstone, Hatch, Monticello, Calvert Cliffs, Dresden, Davis Besse, Peach Bottom, Browns Ferry, Salem, Oconee, Mcguire, Nine Mile Point, Palisades, Palo Verde, Perry, Indian Point, Fermi, Kewaunee, Catawba, Harris, Wolf Creek, Saint Lucie, Point Beach, Oyster Creek, Watts Bar, Hope Creek, Grand Gulf, Cooper, Sequoyah, Byron, Pilgrim, Arkansas Nuclear, Braidwood, Susquehanna, Summer, Prairie Island, Columbia, Seabrook, Brunswick, Surry, Limerick, North Anna, Turkey Point, River Bend, Vermont Yankee, Crystal River, Haddam Neck, Ginna, Diablo Canyon, Callaway, Vogtle, Waterford, Duane Arnold, Farley, Robinson, Clinton, South Texas, San Onofre, Cook, Comanche Peak, Yankee Rowe, Maine Yankee, Quad Cities, Humboldt Bay, La Crosse, Big Rock Point, Rancho Seco, Zion, Midland, Bellefonte, Fort Calhoun, FitzPatrick, McGuire, LaSalle, Fort Saint Vrain, Washington Public Power Supply System, Shoreham, Satsop, Trojan, Atlantic Nuclear Power Plant, Clinch River, Crane
Issue date:	09/19/1981
From:	Partlow J Office of Nuclear Reactor Regulation
To:
References
GL-91-013, NUDOCS 9109160253
Download: ML031140524 (18)

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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

130,

SUBJECT: REQUEST FOR INFORMATION RELATED TO THE RESOLUTION OF GENERIC ISSUE PURSUANT

• ESSENTIAL SERVICE WATER SYSTEM FAILURES AT MULTI-UNIT SITES,"

TO 10 CFR 50.54(f) - GENERIC LETTER 91-13 DISCUSSION

applicants of The purpose of this letter is to inform affected licensees and Issue 130

the technical findings resulting from the NRC resolution of Generic at Multi-Unit Sites," and to (GI-130), "Essential Service Water System Failures multi-unit sites request information from licensees and applicants at affected findings regarding their facilities.

relating to the applicability of certain request for Affected licensees and applicants are required to respond to the staff posi- no new requirements or information contained in this letter, but this letter.

tions are imposed on the affected licensees and applicants by plant The essential service water system (ESWS) is important in maintaining As part of safety during power operation, shutdown, and accident conditions.

service water (LOSW), extensive analyses of our evaluation of loss of essential (BNL). The this issue were performed at the Brookhaven National Laboratory technical findings of this effort at BNL are reported in NUREG/CR-5526, Service Water

"Analysis of Risk Reduction Measures Applied to Shared Essential a Systems at Multi-Unit Sites.* In addition, the NRC staff performed costs regulatory analysis to evaluate the safety benefits and implementation that associated with various equipment and the administrative-type improvements in NUREG-1421, were considered. The staff's regulatory analysis is contained

'Regulatory Analysis for the Resolution of Generic Issue 130: Essential assume that Service Water System Failures at Multi-Unit Sites." These analyses 89-13, "Service the flushing and flow testing provisions of Generic Letter (GL) applied to Water System Problems Affecting Safety-Related Equipment," will be of the crosstie lines as part of addressees' implementation of the resolution (GL 89-13 GI-51, 'Improving the Reliability of Open-Cycle Service Water Systems' of this generic and Supplement 1). On the basis of results of these evaluations UIA-

L)n ?I

p -

Generic Letter 91-13 September 19, 1991'.'...

safety issue, the NRC staff has concluded that the following administrative-type improvements would significantly enhance the availability of the ESWS

affected plants, and their implementation is warranted in view of the in benefit to be derived and the cost of implementation: safety o Technical specification (TS) changes contained.in Enclosure 1 to-enhance the availability of the ESWS as applied to the design configuration affected plants. of 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 retain requirements that operating experience and probabilistic risk to 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 probabilistic risk, assessment performed under GI-130, form the technicalwith the for these 1S and procedures improvements. bases 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 and procedures improvements-would provide a substantial increase in.overall.new TS

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 uRedundant and infrequently,used cooling loops should be flushed and states::

tested periodically at the maximum design flow to ensure that they are flow fouled or clogged. Other components in the service water system should not 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 dedicated RCP seal cooling system similar to that identified also underof a NReactor Coolant Pump Seal Failures." The final decision on the possibleGI-23, backfitting of additional plant improvements has been deferred until completion f

of GI-23; and that aspect of GI-130 is subsumed byGI-23. GI-23 will be resolved following the review of comments received based on the related Re ister Notice published on April 19, 1991. The comment.period has Federal extended until September 30, 1991. Enclosure 3 is provided to you for been informa- tion 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, Jam s G. Partlow Ass ciate 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'

crosstle

3.7.4 At least two independent service water loops per unit and the be between the service water systems of each unit (as applicable) shall

[from the operable. In addition, the crosstle shall be capable of being opened 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.

two

1. With one service water loop per unit OPERABLE, restore at leastunit loops per unit to OPERABLE status within 72 hours, or for the with the inoperable service water loop, be in at least HOT STANDBY

within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> 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 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> 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 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

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, restoretheat least 'one pump to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, or place in operating unit'in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />'and COLD SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />..

2. With one'service water loop in the operating unit INOPERABLE, restore two loops in the operating unit to OPERABLE status within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> or be in at least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD

SHUTDOWN within the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.::

3.' With one [or both] of the crosstie valve(s) INOPERABLE and not capable of being opened [from the control room], within 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> restore the valve(s) to OPERABLE status or open the affected in at valve(s), and maintain the affected valve(s) open; otherwise be within least HOT STANDBY within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> and in COLD SHUTDOWN

the following 30 hours3.472222e-4 days <br />0.00833 hours <br />4.960317e-5 weeks <br />1.1415e-5 months <br />.

- 2- 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 power-operated, or automatic) servicing safety-related valve (manual, is not locked, sealed, or otherwise secured in position equipment that correct position. is in its 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 to its correct position on a equipment actuates test signal;

2. Each service water system pump starts automatically test signal; and on a

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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 capacity isavailable for continued operation of safety-relatedsufficient cooling during normal and accident conditions. The redundant cooling equipment system, assuming a single failure, is consistent with the capacity of this the accident conditions within acceptable limits. assumptions used in In the event of a total loss of service water where backup cooling capacity is available via in a one unit of a two-unit site units, the OPERABILITY of the unit crosstie along crosstie with between the two the shut down unit ensures the availability of sufficient a service water pump in capacity for the operating unit. These limiting conditions redundant cooling significant risk reduction, as indicated will ensure a water system accident. The surveillance by the analyses of a loss-of-service long-term operability of the service waterrequirements system and ensure the short-term and two units.' The service water system crosstle between the crosstie between the the appropriate piping, valves, and instrumentation cross-connectingtwo units consists of of the service water pumps of the two units. By operating the discharge supply of additional redundant cooling capacity from one the crosstie, the the service water'system of the other unit. unit isavailable to

ENCLOSURE 2 BACKFIT ANALYSIS (REFERENCE 10 CFR 50.109)

FOR GENERIC ISSUE 130

A.1 INTRODUCTION

for Generic Issue 130 (GI-130),

This enclosure presents'the backfit analysis at Multi-Unit Sites." The technical

"Essential Service Water System Failures and the regulatory analysis findings for GI-130 are presented in NUREG/CR-5526, apply to 14 reactor units at seven is presented in NUREG-1421. The studies water system (ESWS) failures at these sites and indicate that essential service the overall plant risk. As a plants are a significant'contributor to and based on the cost/benefit analyses consequence of these technical findings, these 14 plants may need to modify performed, the staff has determined that the availability of the ESWS and to technical specifications (TS) to enhance of the high-pressure injection institute procedures to assure the integrity as a result of loss of essential (HPI) pump in the event of RCP seal failure to test and manipulate the ESWS

service water (LOSW), as well as procedures a LOSW accident.

crosstie between the two units during safety enhancements is a reduction in The estimated benefit from the identified in the associated risk of offsite the core damage frequency and a reductionfailure. The reduction of risk to the radioactive releases as a result of ESW to be 4141 person-rem (best estimate public (per plant lifetime) is estimated that these safety enhancements numbers used) and supports the conclusion overall protection of the public health provide a substantial increase in the costs of implementation are and safety. Also, the direct and indirect justified in view of this increased protection.

individually, most of the As discussed in NUREG-1421, when considered be risk associated with this issue would the alternativeS analyzed for reducing the guideline. The objective of cost-effective in meeting the $1000/person-rem loss of the ESWS be reduced consistent GI-130 resolution is that the risk frombackfit rule that the corrective with the two basic requirements of thecost-effective.

alternatives be both substantial and of improvements in TS and One of the potential improvements consisting of reducing the core damage be capable emergency procedures was shown to (1.5E-04/RY)

GCDF) from loss of ESW by 17 percent (or by frequency manner. The staff recognizes the approximately 3.OE-05/RY) in a cost-effective in recognition of the potentially uncertainties in these estimates, and person-rem per plant lifetime), the substantial risk reductions (over 4000 improvements can be achieved by low cost staff believes that significant safety deemed to be consistent with the changes in TS and procedures. This is provisions of the backfit rule.

proposed resolution considered both the The overall approach to arriving at the and the spectrum and type of numerical results of the cost-benefit analysis risk reduction for potential improvements available for potential

'_I

-2 loss-of-service-water sequences. Those alternatives number of occurrences of the LOSW initiators would that could reduce the prevention perspective. Those alternatives that be desirable from the 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 improved procedures would provide a blend of both side, while the capabilities. prevention and mitigation The conclusion of this backfit analysis is that a protection of the public health and safety will be substantial increase in the derived from backfitting of the ESWS improvements and that the backfit is justified favorable cost/benefit ratios. In the following in view of the sections analysis, the nine factors stipulated by 10 CFR 50.109(c) of this backfit determination of backfitting are addressed. to be used in the 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)

performance of the ESW system by providing a blend is to improve the of both mitigation capabilities. This backfit will be applicable prevention and pressurized-water reactor (PWR) plants (14 units) to all the 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 operating and maintaining HPI pump integrity and procedures for 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 in the incidence of public risk from the accidental will yield a reduction offsite radioactive materials of 4141 person-rem (best-estimate) release of average remaining life of 30 years. This backfit per plant with an will frequency from an LOSW by 17 percent (or by approximately reduce the core damage

3.OE-05/RY).

As detailed in Chapter 6 of NUREG-1421., the staff recognizes the uncertainties in these estimates and has considered both the numerical cost-benefit analysis as well as the spectrum and results of the improvements for risk reductions associated with type of potential LOSW sequences.

A.2.4 Occupational Exposure The radiological operational exposure is negligible implementation of Alternative 5 will not result in and, therefore, the radiological exposure to facility employees. any increase in the

N I

-3- A.2.5 Installation Costs with Alternative 5 is The best estimate total cost per reactor associatedinto account, this

$83,000. When the onsite averted costs are taken alternative results in a net savings.

A.2.6 Potential Safety Impact been in various stages A number of generic safety issues related to GI-130 have resolved. The relation of of resolution, including some that have already been these issues to GI-130 is as follows:

generic safety o GI-23, "Reactor Coolant Pump Seal Failures" -- ThisAlternative 6 and, issue addresses the same possible improvements as 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. staff On the basis of the the has identified staff's current knowledge and perspective, is contained an approach for the resolution of GI-23. This approach in Draft Regulatory Guide DG-1008.

of GI-23 An objective of the identified approach for the resolution with RCP seal is to reduce the risk of severe accidents associated seal failure, or to failure by reducing the probability of that it demonstrate that the risk is not significant, thus assuring frequency.

core damage is a relatively small contributor to total of a separate The proposed means of doing so entails the installation RCP seals. Hence, and independent cooling system for the provide a implementation of the proposed GI-23 resolution could As such, the substantial portion of the proposed GI-130 resolution.

with the resolution of GI-23 by resolution of GI-130 is coordinated system to be allowing the installation of a backup RCP seal cooling and review of deferred to the resolution of GI-23 pending the receipt developed as a public comments. It is expected that informationbe helpful in our result of the submittal of public comments will RCP seals under efforts to better understand the performance of the loss of seal cooling conditions.

o GI-51, "Improving the Reliability of Open-Cycle Service-Water was reported Systems" -- The resolution of this generic safety issue of Generic in August 1989 and its imposition began with the issuance 6I-51 entails Letter 89-13 and Supplement 1. Implementation of the and test the implementation of a series of surveillance, control, power plants are requirements to ensure that the ESWS of all nuclear 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.

- 4.-

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 th'ree 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-,150 study is a GI-130 plant; however, it is worthwhile to note that one of the PWR§s(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 andNuclear 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 'isx 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 's 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."

-5- A.2.9 Term of Requirements No. 6 entailing This represents the final resolution of GI-130. Alternativehas been subsumed the installation of an independent RCP seal cooling system 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)

for the Resolution of Generic As identified in NUREG-1421, "Regulatory Analysis at Multi-Unit Sites," a Issue 130: Essential Service Water System Failures the installation of a of combination of potential improvements consisting improvements in technical backup, dedicated RCP seal cooling system, and of substantial risk specifications (TS) and procedures are shown to be capable RCP seal cooling reduction. The specific features of such a backup, dedicated system would be as follows:

o Single high pressure pump, 50-100 gpm capacity at least 8-10

o Dedicated water storage tank with capacity to last hours o AC-independent (non-seismic) pump o No support system cooling required o Once-through RCP seal heat removal the existing literature Limited plant-specific information obtained through with licensees have (FSARs, and so forth), site visits, or discussions already have indicated that a number of the units covered by GI-130 generic safety this plant-unique features that could be responsive to series of PRAs tailored to enhancement. Rather than attempting to perform a or applicant to review each of the 14 units, the NRC encourages each licensee be credited with departing from the plant-specific features (if any) that could modelled in the generic (representative) base case plant configuration may also be considered NUREG/CR-5526. In addition, other design alternatives high-pressure pump seal utilizing arrangements different from that of the injection.

RCP thermal barrier heat One such alternative would provide flow through the the component cooling water exchangers by connecting the fire water system into fire water pump, (CCW) lines. Most fire water systems have one diesel-driven which usually is independent of the ESWS.

deals with this Generic Issue 23, "Reactor Coolant Pump Seal Failures,"

resolving that generic issue is recommendation also, and specific guidance for completion of given in proposed Regulatory Guide DG-1008. While awaiting resolution of this Guide DG-1008, public review and comment on draft Regulatory The reason for this GI-130 item has been deferred until GI-23 is resolved. of 10 CFR 50.63 deferral relates to the earlier development and promulgation an assumption regarding the (station blackout rule), which was based on event. While it was magnitude of RCP seal leakage during a station blackout

-

-2-

-left to GI-23 to validate that assumption, the resolution based on a RCP seal failure LOCA model very similar to of GI-130 is also different from the leakage assumption in 10 CFR 50.63. that of GI-23, but

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 Require- ments (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 o' Potential Safety Benetit 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.

.

-2- 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.OE-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.

alternatives that would help to reduce the consequences of a LOSWThose 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 Response to 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 costmay

-3- 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.

'

ENCLOSURE 7 LIST OF RECENTLY ISSUED GENERIC LETTERS

Generic Date of Letter No. Sub.[e ct Issuance Issued To

91-12 OPERATOR LICENSING NAT. 08/27/91 ALL PWR REACTOR

EXAMINATION SCHEDULE AND APPLICANTS FOR

AN OPERATING LICENSE

91-11 RESOLUTION OF GENERIC 07/18/91 ALL HOLDERS OF

ISSUES 48, "LCOs FOR CLASS OPERATING LICENSES

1E VITAL INSTRUMENT BUSES,"

and 49, "INTERLOCKS AND LCOs FOR CLASS 1E TIE BREAKERS"

PURSUANT TO 1OCFR50.54(f)

91-10 EXPLOSIVES SEARCHES AT 07/08/91 TO ALL FUEL CYCLE

PROTECTED AREA PORTALS FACILITY LICENSEES

WHO POSSESS, USE,

IMPORT OR EXPORT

FORMULA QUANTITIES

OF STRATEGIC SPECIAL

NUCLEAR MATERIAL

88-20 INDIVIDUAL PLANT EXAMINATION 06/28/91 ALL HOLDERS OF

SUPP. 4 OF EXTERNAL EVENTS (IPEEE) OLs AND CPs FOR

FOR SEVERE ACCIDENT VULNERA- NUCLEAR POWER

BILITIES - 10 CFR 50.54 (f) REACTORS

9 1-09 MODIFICATION OF SURVEILLANCE 06/27/91 ALL HOLDERS OF

INTERVAL FOR THE ELECTRICAL OLs FOR BWRs PROTECTIVE ASSEMBLIES IN

POWER SUPPLIES FOR THE

REACTOR PROTECTION SYSTEM

91-08 REMOVAL OF COMPONENT LISTS 05/06/91 ALL HOLDERS OF OLs FROM TECHNICAL SPECIFICA- OR CPs FOR NUCLEAR

TIONS POWER REACTORS

91-07 GI-23 "REACTOR COOLANT 05/02/ 91 ALL POWER REACTOR

PUMP SEAL FAILURES" AND LICENSEES AND

ITS POTENTIAL IMPACT ON HOLDERS OF CPs STATION BLACKOUT

91-06 RESOLUTION OF GENERIC ISSUE 04/29/91 ALL HOLDERS OF OLs A-30, "ADEQUACY OF SAFETY-

RELATED DC POWER SUPPLIED,"

PURSUANT TO 10 CFR 50.54(f)

91-05 LICENSEE COMMERCIAL-GRADE 04/09/91 ALL HOLDERS OF OLs PROCUREMENT AND DEDICATION AND CPs FOR NUCLEAR

PROGRAMS POWER REACTORS

Generic Letter 91-13 -4 September 19, 1991 your Project Manager.

If you have any questions on this matter, please contact Sincerely, Original signed:

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-

3. Background Discussion of a Deferred Safety Enhanc from GI-130 to GI-23

4. Justification Analysis

[10 CFR 50.54(f)] for Ge Letter on GI-130

5. NUREG-1421

6. NUREG/CR-5526

7. List of Recently Issued Generic Letters Reviewed by Barbara Calure, Technical Editor, on 7/12/91.

DISTRIBUTION

Central Files NRC PDR

PDIII-1 r/f MGamberoni

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NAME :MGAMBERONIA/1N :TMARSH :ELEEDs~

IC/ U/91 1/1 /91 DATE :8/6/DT

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CBERLINGER :JPARTLOW

NAME  : ADAANI

DATE  : lygi/9 I . /l6/91 * : /91  : 1 /1>/91 Document Name: GI 130