ML20246A397

Safety Evaluation Report Related to the Operation of Limerick Generating Station,Units 1 and 2.Docket Nos. 50-352 and 50-353.(Philadelphia Electric Company)

ML20246A397
Person / Time
Site:	Limerick
Issue date:	06/30/1989
From:	Office of Nuclear Reactor Regulation
To:
References
NUREG-0991, NUREG-0991-S08, NUREG-991, NUREG-991-S8, NUDOCS 8907060264
Download: ML20246A397 (76)

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.i AVAILABILITY NOTICE -

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1 I NUREG-0991 Supplement No. 8 Safety Evaluation Report L related to the operation of Limerick Generating Station, Units 1 and 2 Docket Nos. 50-352 and 50-353 Philadelphia Electric Company U.S. Nuclear Regulatory Commission Office of Nuclear Reactor Regulation June 1989

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ABSTRACT In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the licensee) for licenses to operate the Limerick Generating Station, Units 1 and 2, located on a site in Montgomery i and Chester Counties, Pennsylvania.

Supplement I to NUREG-0991 was issued in December 1983. Supplements 2 and 3 were issued in October 1984. License NPF-27 for the low-power operation of Limerick Unit I was issued on October 26, 1984. Supplement 4 was issued in May 1985, Supplement 5 was issued in July 1985, and Supplement 6 was issued in August 1985. These supplements addressed further issues that required resolution before Unit I proceeded beyond the 5-percent power level. The full-power operating license for Limerick Unit 1 (NPF-39) was issued August 8,1985, and the unit has completed two cycles of operation.

Supplement 7 was issued April 1989 to address some of the few significant design differences between Units 1 and 2, the resolution of issues that remained open when the Unit I full-power license was issued and an assessment of some of the issues that required resolution before issuance of an operating license for Unit 2.

This supplement addresses the remaining issues that required resolution before issuance of an operating license for Unit 2.

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l Limerick SSER 8 111

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TABLE OF CONTENTS Page ABSTRACT.............................................................. iii 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT................... 1-1 4

1.1 I n t ro du c ti o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 -1 1.12 Modification to the Facility Subsequent to Unit 1 L i ce n s i n g . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1-2 3 DESIGN CRITERIA.FOR STRUCTURES, SYSTEMS, AND COMPONENTS......... 3-1 3.5 Missile Protection......................................... 3-1 3.5.1 Missile Selection and Description................... 3-1 3.5.1.3 Turbine Missiles........................... 3-1 3.9 Mechani cal Sys tems and Components . . . . . . . . . . . . . . . ... . . . . . . . . . 3-2 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures............. 3-2 3.9.3.5 Bulletin 88-05............................. 3-2 4 REACT 0R.........................................................4-1 4.2 Fuel System Des 1gn...................... .................. 4-1 4.2.3 Design Evaluation................................... 4-1 4.2.3.1 Fuel System Damage Evaluation.............. 4-1 4.2.4 Testing, Inspection, and Surveillance Plans......... 4-1 4.2.4.3 Post-Irradiation Surveillance.............. 4-1 6 E NGINE ERED SAFETY FEATURES . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-1 l 6.2 Containment Systems........................................ 6-1 6.2.3 Secondary Containment Functional Design. . . . .. . . . . . . . 6-1 6.2.4 Containment I solation System. . . . . . . . . . . . . . . . . . . . . . . . 6-1 1

6.2.4.2 General Design Criteria 56................. 6-1 6.2.4.3 General Design Criteria 57................. 6-1 6.5 Engineered Safety Feature Atmospheric Cleanup System....... 6-2 6.5.3.3 Reactor Enclosure Recirculation System..... 6-2 v

TABLE OF CONTENTS (Continued)

Page 7 INSTRUMENTATION AND CONTR0LS.................................... 7-1 7.2 Rea cto r Tri p . Sy s tem. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.2.2.7 RTS Powe r Sou rce . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7-1 7.4 Systems Required for Safe Shutdown......................... 7-1 7.4.2.3 Remote Shutdown System............................. 7-1 7.7 Control Systems............................................ 7-1 7.7.2.2 Rod Sequence Control System and Rod Worth M i n i m i z e r . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 -1 8 ELECTRICAL P0WER................................................ 8-1 8.1 Introduction............................................... 8-1 8.1.5 Nonconforming Molded-Case Circuit Breakers -

Bulletin 88-10.................................... 8-1 9 AUXILIARY SYSTEMS............................................... 9-1 9.2 O ther Aux i l i a ry Sy stems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9-1 9.5.1 Fire Protection..................................... 9-1 i

9.5.1.4.3 Alternate Shutdown....................... 9-1 10 STEAM AND POWER-CONVERSION SYSTEM.............................. 10-1 10.4: Other Features........................................... 10-1 10.4.6 Condensate Filter Demineralized System........... 10-1 14 INITIAL TEST PR0 GRAM........................................... 14-1 15 ACCIDENT ANALYSES............... .............................. 15-1 1 15.8 Anticipated Transient Wi thout S cram. . . . . . . . . . . . . . . . . . . . . . 15-1 16 TECHNICAL SPECIFICATIONS....................................... 16-1 17 QUALITY ASSURANCE.............................................. 17-1 17.4 C o n c l u s i o n . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 -1 17.6 Readiness Verification Program........................... 17-1 vi

i-L TABLEOFCONTENTS(Continued)

Page 17.6.1 Independent Construction Assessment (ICA)........ 17-1 17.6.2 Independent Design Assessment (IDA).............. 17-2 18 HUMAN FACTORS ENGINEERING...................................... 18-1 f

18.1 ' Detailed Control Room Design Review...................... 18-1 18.1.1 Background....................................... 18-1 18.1.2 Evaluation of Detailed Control Room Design Review......................................... 18-1 18.1.2.1 System Functional and Task Analysis.... 18-2 18.1.2.2 Control Room Inventory. . . . . . . . . . . . . . . . . 18-2 18.1.2.3 Contro l Room Su rvey. . . . . . . . . . . . . . . . . . . . 18-2 18.1.2.4 Assessment of HEDs..................... 18-3 18.1.2.5 Coordination of Control Room Improvements with c+her Programs.....

. 18-3 18.1.2.6 Staff Conclusions on 9RDR............. 18-3 18.2. Safety Parameter Display System.......................... 18-3 18.2.I' Background........................................ 18-3 18.2.2 Evaluation........................................ 18-4 18.2.3 Staff Conclusions of SPDS......................... 18-4 19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS......... 19-1 22 FINANCI AL PROTECTION AND INDEMNITY REQUIREMENTS. . . . . . . . . . . . . . . . 22-1 22.1 Funds for Decommissioning. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22-1 APPENDICES A CHRONOLOGY S REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS T REPORT ON CONTROL ROOM DESIGN REVIEW U ERRATA TO THE SAFETY EVALUATION REPORT FOR THE LIMERICK GENERATING STATION vii

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l 1 INTRODUCTION AND GENERAL DESCRIPTION OF PLANT 1.1 Introduction I

In August 1983, the Nuclear Regulatory Commission (hereinafter referred to as  !

NUREG-0991, )

theNRCorthestaff)issueditsSafetyEvaluationReport(SER)[

recarding the application of the Philadelphia Electric Company hereinafter referred to as PEco or the licensee) for licensees to operate the Limerick i Generating Station, Units 1 and 2, Docket Nos. 50-352 and 50-353.

Supplement 1 to the SER was issued in December 1983, Supplements 2 and 3 were issued in October 1984, and Operating License NPF-27, authorizing power up to 5 percent, was issued on October 26, 1984. Supplement 4 to the SER was

. issued in May 1985, Supplement 5 was issued in July 1985, and Supplement 6 was issued in August 1985. These supplements addressed issues that required further resolution before Unit 1 proceeded beyond the 5-percent power level.

A full-power operating license (NPF-39) was issued for Limerick Unit 1 on August 8, 1985.

As noted above, the staff's SER assessed operation of both Limerick Units 1 and 2. Construction of Unit 2 was halted in January 1984 by Order of the Pennsylvania Public Utility Commission. At the time, construction was about 30 percent complete. Construction of Unit 2 resumed in February 1986 with PECo's agreement to accept a cost containment cap of about $3.1 billion for construction and certain operational incentive prcgrams. On May 3, 1988, the Commission modified Construction Permit CPPR-107 to extend the earliest and latest completion dates to May 1,1989, and January 1,1992, respectively.

Supplement 7 to the SER was issued April 1989 and primarily related to Unit 2. SSER 7 addressed some of the few significant design differences between Units 1 and 2, the resolution of issues that remained open when the Unit I full-power license was issued and an assessment of some of the issues that required resolution before issuance of a low-power operating license for fuel loading, initial criticality and power ascension up to 5 percent for Limerick Unit 2.

This document, the eighth supplement to the SER (SSER 8), also primarily relates to Unit 2. This supplement addresses the remaining issues that required resolution before issuance of an operating license for Unit 2.

Each of the sections and appendices of this supplement is designated the same as the related portion of the SER. Each section is supplementary to and not in lieu of the discussion in the SER and Supplements 1 through 7, unless otherwise noted. Appendix A is a continuation of the chronology of this safety review. Appendix H lists the principal contributors. Appendix S is the ACRS report of May 11, 1989 on Unit 2. Appendix T is the contractor's report on review of the DCRDR and SPDS.

Copies of this supplement are available for inspection at the NRC Public Document Room, 2120 L Street, N.W., Washington, D.C. and at the local Public Document Room at the Pottstown Public Library, 500 High Street, Pottstown, Pennsylvania 19464.

Limerick SSER 8 1-1

The NRC Project Manager for Limerick Units 1 and 2 is Richard J. Clark. He may be contacted by telephone at (301) 492-3041 or by mail at the following address:

Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 1.12 . Modification to the Facility Subsequent to Unit 1 Licensing There have been a number of design changes and modifications made to Limerick Unit I since its license was issued on August 8, 1985. Some of these changes are the licensee's initiative, some are required by the NRC, and some required as a condition of the license. All design changes and modifications approved by the staff for Unit 1 also have been implemented for Unit 2. Some of the more significant modifications that required staff approval were discussed in SSER 7. The other modifications, listed in SSER 7 but not addressed, are repeated below:

Required / SSER-8 Modification Initiated By Section Install Isolation Valve in Hydrogen Recombiner Line L.C.2.C.11 6.2.4.2 Remote Shutdown System Redundancy L.C.2.C.12 7.4.2.3 Reactor Enclosure Cooling Water and Drywell Cooling Water Isolation Valves L.C.2.C.10 6.2.4.2 Add Stairway L.C.2.C.3 9.5.1.6.3 Elimination of Reactor Enclosure Recirculation System Cooldown Mode Licensee 6.5.3.3 Eliminate Reactor Sequence Control System / Lower Rod Worth Minimizer Set Point Licensee 7.7.2 i

Limerick SSER 8 1-2

3 DESIGN CRITERIA FOR STRUCTURES, SYSTEMS, AND COMPONENTS 3.5 Missile Protection 3.5.1 Missile Selection and Description 3.5.1.3 Turbine Missiles By letter dated August 19, 1903 Philadelphia Electric Company submitted the Limerick Unit 2 low pressure turbine maintenance program to satisfy the requirement identified in section 3.5.1.3.3 of the Limerick Unit 2 Safety Evaluation Report, NUREG-0991.

Although large steam turbines and their auxiliaries are not safety related systems as defined by NRC regulations, failures that occur in these turbines can produce large, high energy missiles. General Design Criterion 4,

" Environmental and Missile Design Bases," of Appendix A to 10 CFR Part 50 requires, in part, that structures, systems, and components important to safety be appropriately protected against the effects of missiles that might result from such failures. In the past, evaluation of the effects of turbine failure on the public health and safety followed Regulatory Guide 1.115, Turbine Missiles," and three essentially "independent ProtectionStandard Against Low-Trajectory Review Plan (SRP) sections,10.2, "Turbir.e Generator,"

10.2.3 " Turbine Disk Integrity," and 3.5.1.3 "Turt'ce Missiles." According to NRC guidelines stated in section 2.2.3 of the SRP and Regulatory Guide 1.115, the probability of acceptable damage from turbine missiles 4P4 ) should be less than or equal to about I chance in 10 million per year for an individual plant. The probability of unacceptable damage resulting from turbine missiles is generally expressed as the product of (1) the probability of turbine failure resulting in the ejection of turbine disc (or internal structure) fragments through the turbine casing (P ); (2) the probability of 3

ejected missiles perforating intervening bar,iers and striking safety-related structures, systems, or components (P ); and (3) the probability of struck structures, systems,orcomponentsfa$lingtoperformtheirsafetyfunction

'(P3 )*

In recent years, the staff has shifted review emphasis from P,3 and P to P .

Applicantsandlicenseesarerequiredtoshowthattheturbinkmissile 3 generation probability, P , satisfies turbine reliability requirements criteria. such as turbines, P Foris an unfavorably required oriented to be less than 10- turbing,perThis year.the is the general,Limerick Un minimumrella,bilityrequirement(CriterionA)forloadingtheturbineand Criterion 8 states that if the probability bringingthgsystemonline.per exceeds 10- year but less that 10-4 per year, the turbine may b j serviceuntilthenextscheduledoutage,atwhichtimgthelicenseeisto l take action to reduce the probability to meet the 10- per year limit before returning the turbine to service. In order to assure tlat the turbine missile probability follows the turbine reliability criteria, the staff requires applicants and licensees to submit, for approval, a turbine inspection Limerick SSER 8 3-1

l program including inspection intervals that are based on turbine missile I generation analysis.

l l Limerick Unit 2 has three General Electric low pressure turbines - LPA, LPB, and LPC. The licensee calculated the turbine missile generation probability of each low pressure turbine using the General Electric turbine reliability methodology. The staff has approved the GE methodology as documented in Safety Evaluation Report related to the operation of Hope Creek Generating Station Supplement Ne. 6, NUREG-1048, July 1986.

I The GE methodology considers turbine wheel operating conditions such as temperature, pressure, steam quality, material properties, inspection intervals and other factors. The results of the licensee's calculation show that the turbine missile generation probability of each low pressure turbine is less than 10-5 per year if each low pressure turbine is inspected every six operating years. The licensee has also included the turbine system maintenance program and the revision to the methodology of projecting turbine i missile generation in FSAR sections 10.2.3.6 and 3.5.1.3, respectively.

The staff concludes that the turbine maintenance program at Limerick Units I and 2 is acceptable because the calculated turbine missile generation ,

probability satisfies the staff's turbine reliability requirement criteria '

specified in section 3.5.1.3.3 of the SER. The staff's evaluation and approval of the Unit 1 turbine maintenance program was transmitted to PECo by our letter of November 3,1987. The staff's evaluation and approval of a similar program for Unit 2 was transmitted by our letter of May 9, 1989.

3.9 Mechanical Systems and Components 3.9.3 ASME Code Class 1, 2 and 3 Components, Component Supports, and Core Support Structures 3.9.3.5 Bulletin 88-05 NRC 8ulletin 88-05 (May 6, 1988) and Supplements 1 and 2 to that bulletin (June 15 and August 3,1988, respectively) were issued requiring holders of construction permits and operating licenses at less than full power to submit information regarding materials supplied by Piping Supplies, Incorporated (pS Inc.) at Folsom, New Jersey; West Jersey Manufacturing Company (WJMC) at Williamstown, New Jersey; and Chews Landing Metal Manufacturers (CLMM). The bulletin further requests that applicants and licensees (1) take action to ensure that materials comply with the ASME Code and design specification requirements or are suitable for their intended service or (2) replace such materials. The NRC action was the discovery that certified material test reports (CMTRs) precipitated for material by supplied by PS Inc., WJMC, and CLMM contained false information about material supplied to the nuclear industry. A number of CMTRs were apparently used to certify that commercial-grade steel meets the requirements of ASME Code Section III, Subarticle NCA-3800.

Limerick SSER 8 3-2

Philadelphia Electric Company (PECo) responded to Bulletin 88-05 for Limerick t

Unit 2 in a report dated March 31, 1989. In response to comments and questions from the staff, PECo responded with a revised report dated June 2, 1989 which enabled the staff to complete _its review of PECo's submittals.

The response contained sections which described: the methodology used to identify, test and evaluate WJM/ PSI /CLM material; the document and procurement review and testing programs; and an engineering evaluation and analysis of I

the nonconforming items.

The staff has reviewed the licensee's response to NRC Bulletin 88-05 which defines specific action and reporting requirements with respect to identifying, locating and testing nonconforming flanges and fittings supplied by PSI /WJM/CLM and evaluating their adequacy and suitability for their intended service.

The licensee's response consisted of 2 reports. The reports describe the methodology used to identify and test the nonconforming parts; contain a summary of the test results; and present the engineering evaluations and analyses.

The licensee conducted a multi-faceted program to identify and locate materials supplied by PSI /WJM/CLM. Initially, PECo conducted an in depth document review and field inspection. In addition, PECo conducted confirmatory reviews using data provided by GE. PECo determined that Limerick Unit 2 had 312 items of installed safety-related carbon steel material and 13 items of installed safety-related stainless steel material from WJM or PSI. PECo found no items supplied by CLM.

Stress analyses were performed for each of the 52 installed safety-related items which were found to have tensile strengths below 66 ksi (396 L D converted to 137 BHN). The 52 items consisted of:

46 carbon steel flanges (45 flanges + 1 blind flange) 6 stainless steel flanges (4 flanges + 2 blind flanges)

Structural evaluation of the nonconforming flanges was based on the assumption that the reduced flange capacity is linearly dependent on the yield strength of the material.

On the basis of its review of the licensee submittals, the staff finds that PEco conducted adeouate material property tests and structural analyses of the nonconforming flanges and fittings using acceptable and conservative analytical methods and evaluation criteria. The staff also finds that PECo was responsive to the action and reporting requirements of Bulletin 88-05, Supplements I and 2, and has qualified all nonconforming parts as being suitable for the intended service. I I

The staff concludes that the analytical procedures used by PECo to qualify the nonconforming parts and the results of the analyses provide an adequate basis for resolving the concerns with respect to demonstrating adequacy for service. The staff does not consider the nonconforming parts to be ASME Code material. However, the staff finds the use of this material is an Limerick SSER 8 3-3

acceptable alternative in accordance with 10 CFR 50.55a(a)(3)(ii) because full compliance with all specified requirements would result in hardship or-unusual difficulties without a compensating increase in the level of quality.

or safety.

.i'hestaff'sdetailedsafetyevaluationwassenttothelicenseebyletter dated June 20, 1989.

Limerick SSER 8 3-4

4 REACTOR 4.2 Fuel System Design i

4.2.3 Design Evaluation 4.2.3.1 Fuel System Damage Evaluation In SSER-7, the staff discussed an occurrence of " crud-induced localized corrosion" (CILC) failures identified during the Limerick Unit I second cycle. In that discussion, the fuel for Unit 2 was identified as entirely composed of the new GE 8X8 barrier fuel. This is correct except for 92 low duty natural uranium peripheral bundles which are not barrier fuel.

SSER-7 described the results of the inspections of the fuel assemblies discharged during the first Limerick Unit I refueling outage as a number of fuel pins with externally corroded cladding. In the licensee's submittal of April 3, 1989, these inspections were described as having no unusual findings. Other numerical revisions to the SSER-7 discussion are identified in the errata.

l The licensee also notified the staff via letter dated June 6, 1989, that the initial core of Limerick Unit 2 will contain four (4) fuel assemblies manufactured with an alternate clad surface process. The assemblies will be loaded into Unit 2 as part of a test to demonstrate the corrosion resistance of this cladding material manufactured with an alternate surface finish process.

The assemblies satisfy the fuel rod thermal-mechanical criteria described in the Limerick 2 Safety Analysis Report and in GESTAR-II when operated within the Maximum Average Planar Linear Heat Generation Rate (MAPLHGR) limits established for the standard 2.48 wt % bundle design. No new materials are introduced to the reactor environment. In addition, the clad surface finishing processes used for the assemblies have been previously used in manufacturing fuel assemblies for other reactors. As previously mentioned, the process changes being tested are such that comparable corrosion resistance is expected relative to the production fuel. Therefore, assumptions regarding corrosion rates in the cladding mechanical analyses conservatively bound the expected corrosion rates of then assemblies.

4.2.4 Testing, Inspection and Surveillance Plans 4.2.4.3 Post-Irradiation Surveillance The SER indicated that PECo would provide a post-irradiation surveillance program that would satisfy SRP 4.2. By letter dated September 12, 1986 from M.J. Cooney, Philadelphia Electric Company (PECo), to Robert Bernero, NRC PEco informed NRC that Director routine visual of Boiling Waterof inspection Reactor Licensing,(usua,lly discharged) fuel will be representative performed to comply with a provision reported in Section 4.2.4.3 of the NRC's Safety Evaluation Report (SER) of the Limerick Unit 1 Station by using the i l

Limerick SSER 8 4-1 1

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General Electric (GE) Post Irradiation Surveillance Program. The NRC staff has reviewed this letter to confirm that the licensee's proposal satisfies j

'the condition of approval reported in Section 4.2.4.3.

The General Electric (GE) Post Irradiation Surveillance Program has been previously reviewed by the staff and approved based on the use of design methodologies and testing procedures described as in accordance with the Post-Irradiation Surveillance,Section II.D.3 of Standard Review Plan (SRP) 4.2 (Ref. 3). Therefore, we conclude that the Post-Irradiation Surveillance Program of the Spent Fuel Assembly of Limerick Generating Station Unit 1 is acceptable.

The licensee proposed that the routine visual inspection of representative (usually discharged) fuel be performed consistent with the GE Post-Irradiation Surveillance Program. The staff finds that the licensee proposal is acceptable and appropriate. This information was previously transmitted to the licensee by letters dated May 4,1987 and June 1,1987.

On August 16, 1988, the applicant informed the staff that the same program would be implemented on Limerick Unit 2. The staff has concluded, based on the considerations discussed above, that the Post-Irradiation Fuel Surveillance has been correctly based on the results of the pre-approved staff analyses and are acceptable.

Limerick SSER 8 4-2

6 ENGINEERED SAFETY FEATURES 6.2 Containment Systems 6.2.3 Secondary Containment Functional Design SGRS Connection to Refueling Floor ,

SSER-7 described the licensee's January 13, 1987 proposal as (1) adding i prefilters; (2) adding fans to decrease the drawdown time; and (3) adding the necessary Technical Specification changes. A more appropriate description i would be to (1) directly connect the SGTS to the refueling area, (2) add two j new SGTS fans with higher capacity (8400 cfm) to replace the existing fans  !

(3000 cfm) to enable the system to meet the present drawdown time limitations with the simultaneous connection of all three secondary containment zones, and (3) revise the Technical Specifications to add new isolation signals and .

surveillance requirements. The proposal was previously approved on July 8, 1987.

6.2.4 Containment Isolation System 6.2.4.2 General Design Criteria 56 SSER-3 identified two areas where the Limerick Containment Isolation System was not completely in compliance with General Design Criteria (GDC) 56.

Corresponding License Conditions were imposed on Limerick Unit 1. Appropriate isolation valves have been added in the hydrogen recombiner lines and appropriate isolation signals were added to the reactor enclosure cooling water and the drywell cooling water isolation valves, as discussed in the staff's SER supporting Amendment No. 13 to the Limerick Unit 1 Technical Specifications, dated January 10, 1989. These modifications have also been made to Limerick Unit 2 (See Section 16), These modifications, including location and power supplies, have been reviewed and found to meet the provisions of Standard Review Plan 6.2.4, including GDC-56; and 10 CFR 50.44 and the License Conditions were deleted.

6.2.4.3 General Design Criteria 57 A generic safety evaluation, dated March 11, 1986, indicated the omission of the containment radiation isolation signals from two inch vent and purge lines could be acceptable on a plant specific basis assuming that operator action to close these valves in the event of failure to isolate from the other two signal sources would be taken within 30 minutes of initiation of accidents. To assure this assumption is valid, the generic evaluation called for individual licensees to verify the 30 minute operating time. With their submittal dated June 20, 1986, the licensee provided information to demonstrate compliance with the 30 minute operating action.

Safety related remote manual switches for the four two inch purge and vent isolation valve operators are located in the control room to assure accessibility to the operator, Instrumentation available to determine the need for manual closure of the purge and vent lines includes instrumentation Limerick SSER 8 6-1

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j which would detect small leaks inside the primary containment as well as increases in radioactive effluent. This instrumentation includes the south stack effluent radiation monitors, containment high range area radiation monitors, containment reactor coolant pressure boundary leakage detection  ;

radiation monitors, and containment high temperature monitors. These latter l three monitors alarm in the control room and serve to provide the operators with indications that an abnormal condition exists inside the containment.

Since the operator will have taken conscious actions to open these valves, receipt of any of these alarms is indicative that the valves should be closed. Furthermore, the plant operator's training prepares the operators to handle situations which may require manual containment isolation. In order to provide additional support, the licensee is revising the appropriate operating procedures to include statements regarding closure of these valves within 30 minutes of such an event. We concur with the licensee that the operator has enough time due to sufficient instrumentation, plant procedures and operator training to isolate the two inch purge and vent valves within 30 minutes of accident initiation in the event that other isolation signals (containment pressure and reactor water level) fail to cause automatic isolation of the valves.

Based on our review of the licensee's submittal, we conclude that the licensee has verified that 30 minutes is sufficient time to allow the operator to close the two inch vent and purge lines following an accident in the event that other isolation signals (containment pressure and reactor water level) fail to cause automatic isolation of the valves. This information was previously provided to the licensee on August 22, 1986.

6.5 Engineered Safety Feature Atmospheric Cleanup System 6.5.3.3 Reactor Enclosure Recirculation System With their letter dated December 29, 1986, Philadelphia Electric Company submitted a proposal to modify the filter train configuration within the Reactor Enclosure Recirculation System (RERS) for Limerick Unit 2.

Specifically, the licensee proposed to delete the cooldown air flow path to each of'the charcoal adsorbers in the RERS filter trains. The cooldown air is provided upstream of each charcoal adsorber to limit excessive charcoal temperature rises due to potential radioactive decay heat buildup in the adsorbers during and following an accident.

It should be noted that the separate water deluge system, which is provided within the charcoal adsorbers for fire protection, is not affected by this proposal.

The RERS is a redundant engineered safety feature system. Each system is provided with a filtration train that has a design air flow capacity of 60,000 cfm. Each filtration train consists of a prefilter, an upstream HEPA, a charcoal adsorber and a downstream HEPA. The RERS is designed to provide a filtration of contaminated air (iodines and particulate) in the reactor

. enclosure building following a postulated accident or an abnormal release of l

high airborne radioactivity into the reactor enclosure. A cooldown air path, equipped with an isolation valve, is provided upstream of each charcoal adsorber. The cooldown air is supplied only to a charcoal adsorber that (1) l is not in operation and (2) is indicating an excessive charcoal temperature Limerick SSER 8 6-2

a rise due to radioactive decay heat. In addition, a separate water deluge system is provided within each charcoal adsorber for fire protection.

The licensee stated in the referenced letter that the maximum calculated temperature rise due to radioactive decay heat buildup following a design basis loss of coolant accident (LOCA) is 3.2"F. This value was estimated on the basis of the conservative assumption that the Reactor Enclosure Secondary Containment post-LOCA atmosphere temperature is 150 F.

We have reviewed the following licensee assumptions which were used to calculate the temperature rise:

(1) 25% of the core iodine inventory is immediately available for leakage from the primary reactor containment. This is consistent with the Regulatory Position in Regulatory Guide 1.3.

(2) 50% of the core iodine inventory is in the suppression pool water. This is consistent with the Regulatory Position in Regulatory Guida 1.7.

(3) The primary containment atmosphere leaks to the reactor building at a rate of 0.5 percent per day, with additional steam leakage of 11.5 standard cubic feet per hour through each main steam isolation valve.

Five gpm of suppression pool water is also assumed to be leaking into the Reactor Building through equipment leakage. These assumed values are consistent with those used in recently issued BWR technical specifications.

(4) An air mixing efficiency of 50 percent within the secondary containment.

This assumption is consistent with the acceptance criteria in SRP Section 6.5.3.

(5) A decontamination factor of 10 was used for iodine which becomes airborne after partitioning from the suppression pool water leakage.

This assumption is consistent with the assumption used in NUREG-0016.

(6) A secondary containment post-LOCA environmental temperature of 150 F.

The staff's Regulatory Position is 180*F, as delineated in Regulatory Guide 1.52. This is addressed in the paragraph at the end of this section.

(7) A 100 percent removal efficiency for radioactive iodines by tha charcoal adsorber. The staff considers t.is conservative for purposes of decay heat calculations.

(8) All of the iodine beta energies and 50 percent of gama energies for each isotope are adsorbed in the charcoal. The staff considers this assumption conservative.

(9) No iodine removal mechanism other than the charcoal adsorption. The staff considers this assumption conservative.

Limerick SSER 8 6-3

(10) No unfiltered leakage to the environment during the initial system 2 minute 15 second drawdown time after an accident. The staff considers this assumption conservative.

(11) No decay heat load from noble gases. Because of the transient time for such isotopes, the staff finds this assumption acceptable.

(12) The maximum heat load due to the heat of iodine oxidation is approximately 42 watts (3 percent of the decay heat load). The staff finds this assuxption acceptable.

(13) A 1,520 fte filter front surface area for natural convective heat loss.

The staff finds this assumption acceptable.

(14) A 820 ft2 outside steel surface area fo- radioactive heat loss. The staff finds this assumption acceptable.

(15) The minimum charcoal ignition temperature of 627"F. The staff's regulatory position in Regulatory Guide 1.52 stipulates that the minimum charcoal ignition temperature is 627'F (330 C) at a 100 feet per minute air flow rate through the adsorber. With the adsorber isolated (no air flow), the staff estimates that the minimum charcoal iqnition temperature could be as low as 450 F in a 2 inch adsorber bed.

Using the above assumptions, the licensee has calculated a cumulative charcoal temperature of 153.2*F at the time of maximum iodine loading during and following an accident. The staff has reviewed the licensee's assumptions and, with two exceptions, finds them to be acceptable. Our acceptance is based on either (1) conformance with appitcable Regulatory Guides and other staff guidance, or (2) sufficient conservatism to account for uncertainties in the analysis. The two exceptions are with respect to the licensee's assumptions regarding the secondary containment post-LOCA environment temperature (150*F) and the minimum charcoal ignition temperature (627'F).

Using a 180*F secondary containment post LOCA environment temperature, the staff estimates that the charcoal temperature may reach up to 250 F with gradual radioactive iodine loading into the charcoal adsorber during and following an accident. Also, the charcoal ignition temperatuia may be as low as 450 F with no air flow through the 2 inch thick adsorber. In view of these considerations, the staff concludes that the maximum expected charcoal temperature rise in the adsorber still be be well below the minimum charcoal ignition temperature. Therefs e, the staff accepts the licensee's request to delete the cooldown air paths to the charcoal adsorbers.

On the basis of the above evaluation, the staff concludes the licensee's request to delete the cooldown air flow paths to the charcoal adsorbers is acceptable. The bases for acceptance are that (1) the expected maximum charcoal temperature rise in the adsorber due to potential radioactive decay heat buildep during and following an accident is well below the minimum charcoal ignition temperature, ard (2) a separate water deluge system within each adsorber is provided with temperature alarm set points. Hence, the requested change does not affect significantly the risk of a charcoal fire following an accident, and the safety function of the filters will be preserved.

Limerick SSER 8 6-4

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'The staff's evaluation and approval to remove the cooldown' air flow' path to  !

each.of the charcoal adsorbers in the Unit 2 RERS filter trains was  :

transmitted to;PECo by our letter of March 18, 1987. Similarly, our~ letter I of December 7, 1988 approved modification of the Unit 1 RERS and the common  !

plant Standby ~ Gas Treatment System to remove the cooldown mode of operation  !

for the' charcoal beds.

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' Limerick SSER 8 6-5

t 7 INSTRUMENTATION AND CONTROLS 7.2 Reactor Trip System 7.2.2.7 RTS Power Source The SER discussed the RTS power sources in Sections 7.2.27 and 8.3.1. In both  !

discussions, the preferred source was identified as a non-Class 1E plant l auxiliary power source, and the alternate was identified as an invertor  :

supplied by 250Vdc Class IE batteries. The FSAR was revised in Amendnment 45 to reflect the addition of a second alternate power source. During this review, the staff identified that the preferred source is actually the invertor supplied by the 250Vdc Class IE batteries and the alternate'is a transformer, which can now be supplied through a transfer switch either from a non-Class 1E 440V MCC or from a non-Class IE 440V MCC energized by a ,

non-Class 1E UPS. The original staff review focused on the quality of.the.

power source and the Class IE.to non-Class IE separation.

The use of the Class IE batteries supplied invertor as the preferred source and adding a non-Class IE UPS source are considered improvements in the power source quality and do not affect the separation aspects and are, therefore, acceptable.

7.4 Systems Required for Safe Shutdown 7.4.2.3 Remote Shutdown System In SSER-3 and SSER-5, the staff found that the Limerick remote shutdown system did not fully comply with GDC-19. The license was conditioned to L require modifications that provide a redundant safety-related method of achieving safe shutdown conditions without liftin- leads or adding jumpers as described in the licensee's letters of April 18 and 22, 1985. In FSAR Amendment 52, the licensee provided information which incorporated these modifications as made during the Limerick Unit I first refueling outage. The licensee has incorporated similar capabilities into the Limerick Unit 2 design to allow remote operation of the D RHR pump, the D RHRSW pump and the D ESW pump. Therefore, the associated license condition will not be required for Limerick Unit 2.

7.7 Control Systems 7.7.2.2 Rod Sequence Control System and Rod Worth Minimizer By letter dated November 9, 1988, Philadelphia Electric company (PECo) requested approval to remove the Rod Sequence Control System (RSCS) and to lower the Rod Worth Minimizer (RWM) Low Power Setpoint from 20 percent to 10 percent at Limerick Unit 2. This modification for Unit 2 was approved by our letter of February 7, 1989. By letter dated December 14, 1988, PECo also submitted an application to amend the Technical Specifications for Limerick Unit 1 to permit the same modifications to be made at Unit 1. The Unit I amendment was approved on March 22, 1989. (Amendment No. 17)

Limerick SSER 8 7-1

The Rod Sequence Control System restricts rod movement to minimize the individual worth of control rods to lessen the consequences of a Rod Drop Accident (RDA). Control rod movement is restricted through the use of rod select, insert, and withdrawal blocks. The Rod Sequence Control System is a hardwired (as opposed to a computer controlled), redundant backup to the Rod Worth Minimizer. It is independent of the Rod Worth Minimizer in terms of i inputs and outputs but the two systems are compatible. The RSCS is designed l

to monitor and block when necessary, operator control rod selection,

withdrawal and insertion actions, and thus assist in preventing signifcant control rod pattern errors which could lead to a control rod with a high reactivity worth (if dropped). A significant pattern error is one of several abnormal events which must occur to have an RDA which might exceed the fuel energy density limit criteria for the event. It was designed only for possible mitigation of the RDA and is active only during low power operation (currently generally less than 20 percent power) when an RDA might be significant. It provides rod blocks on detection of a significant pattern error. It does not prevent an RDA. A similar pattern control function is also performed by the RWM, a computer controlled system. All reactors having an RSCS also have an RWM.

In response to a topical report submitted by the BWR Owners' Group on December 27, 1987, the NRC staff issued a letter with a supporting safety evaluation approving (1) elimination of the RSCS while retaining the RWM to provide backup to the operator for control rod pattern control and (2) the lowering of the setpoint for turnoff of RWM to 10% of rated thermal power from its current 20% level. (Letter, A. C. Thadani, NRC to J. S. Charnley, GE,

Subject:

Acceptance for Referencing of Licensing Topical Report NEDE-24011-P A. " General Electric Standard Application for Reactor Fuel,#

Revision 8, Amendment 17).

The letter of December 27, 1987 and supporting safety evaluation approving the topical report concluded that the modifications proposed by PECo were acceptable provided:

_(1) The Technical Specifications (TS) should require provisions for minimizing operations without the RWM system operable.

(2) The occasional necessary use of a second operator replacement should be strengthened by a utility review of relevant procedures, related forms and quality control to assure that the second operator provides an effective and truly independent monitoring process. A discussion of this review should accompany the request for RSCS removal.

(3) Rod patterns used should be at least equivalent to Banked Position Withdrawal Sequence (BPWS) patterns.

Withrespecttoitem(1)above,intheletterofNovember9,1988,PECo stated that "TSs for Limerick 2 are currently under review and revisions will be proposed to minimize operations without the active use of the RWM." The first draft of the proposed Limerick Unit 2 TSs was submitted by PECo's letter of June 3, 1988. The Unit 2 TSs are going to be identical to the Unit 1 TSs unless there is something that requires a change (e.g., difference in design). PEco's application of December 14, 1988 submitted proposed TSs for Limerick SSER 8 7-2

Unit I which we have reviewed and find acceptable. We conclude that item (1) will be adequately addressed.

With regard to item (2) above, PECo described the programs and procedures that would be provided during instances when the RWM is not available to independently verify the correctness of the first operator's actions during rod movements. The procedure for " bypassing the Rod Worth Minimizer,"

procedure S73.0.D., Rev. 7 dated October 31, 1988 has been reviewed by the resident inspectors and the NP,R Project Manager; we conclude that it provides acceptable controls when used in conjunction with the specific procedural restrictions listed in the November 9, 1988 and December 14, 1988 letter.

During the January 11, 1989 shutdown of Unit I for the second refueling outage, the resident inspectors observed management's attention to the procedures.

We conclude that the procedural controls are acceptable.

The Rt'M at Limerick Unit 2 is a NUMAC system that utilizes the BPWS patterns recommended in the staff December 27, 1987 letter. This satisfies item (3) above.

PECo's proposal to remove the RSCS and to lower the RWM low power setpoint from 20 to 10 percent at Limerick Unit 2 meets the requirements detailed in the staff's letter of December 27, 1987 approving the topical report on these modifications. Accordingly, the modifications proposed in PECo's letter of November 9, 1988 are found to be acceptable and are approved. As previously, noted, this information was previously transmitted to the licensee on February 7, 1989.

Limerick SSER 8 7-3

8 ELEC7RICAL POWER 8.1 introduction 8.1.5 Nonconforming Molded-Case Circuit Breakers - Bulletin 88-10 NRC Bulletin 83-10, Nonconforming Molded-Case Circuit Breakers issued November 22, 1988, requires Construction Permit holders to complete the Bulletin actions before fuel load.

In its response to Bulletin 88-10 dated April 3, 1989, the applicant informed the staff of actions taken in complying with the information request of Bulletin 88-10. The staff has evaluated the applicant's response and notes the following:

• A total of 122 molded-case circuit breakers have been identified by the ifcensee and are being maintained as stored spares for safety-related (class IE) application in LGS Unit 2.

• The licensee has verified the traceability of these circuit breakers to their manufacturer. (The breakers were manufactured b Electric Corporation and were supplied by Eaton, Inc.)y West %ghouse .

• Additional safety related circuit breakers will be purchased from the original equipment manufacturer. Purchase orders will invoke the requirements of Appendix B to 10 CFR Part 50, and 10 CFR Part 21.

The licensee shall maintain all information obtained from this review and evaluation for five years.

The staff found this response acceptable and consistent with the information request in Bulletin 88-10 for Limerick, Unit 2. The licensee was notified by letter dated May 9, 1989.

Limerick SSER 8 8-1

9 AUXILIARY SYSTEMS 9.2 Other Auxiliary Systems

,9.5.1 Fire Protection 9.5.1.4.3 Alternate Shutdown SSE'R-2 reports that the design objective of the remote shutdown panel is to achieve and maintain cold shutdown in the event of an evacuation as a result of a fire that disables the control room. Section 9.5.1.4.2 and 9.5.1.4.3 identify fires in the control room and in the auxiliary equipment room as having the possibility of disabling the control room. SSER-2 was not clear, however, in identifying that a fire in the cable spreading room may also disable the control room and require use of the remote shutdown panel.

However, the conclusions of SSER-2 remain valid.

Limerick SSER 8 9-1

10 STEAM AND POWER-CONVERSION SYSTEM 10.4 Other Features 10.4.6 Condensate Filter Demineralized System SSER-7 described a review by the licensee of " crud-induced localized corrosion" (CILC) failures identified during the Limerick Unit 1 second cycle of operation.

The staff's SSER-7 also discussed several commitments by the licensee discussed in the letter of April 3, 1989 and in a meeting held March 15, 1989 regarding water chemistry and corrosion control. After further review of the applicant's April 3, 1989 letter, these commitments are identified as follows:

For Limerick Unit 1, PECo will:

employ continuous in-line analyzers to monitor water chemistry by using ion chromatography to monitor feedwater and reactor water ionic concentrations during startup using a gas chromatograph to measure oxygen and nitrogen concentrations in the reactor water during startup using a sensitive TOC analyzer to measure TOC concentrations in the feedwater install CAV and ECP to characterize the chemistry environment limit chemical transients by establishing goals to limit TOC in radwaste sample tank transfers to the CST to less than 200 ppb limit condenser air inleakage to less than 30 scfm control feedwater dissolved oxygen to between 20 to 50 ppb

- evaluate procedures and train operators in condensate / demineralized operation minimize copper in feedwater by

- changing filter / demineralized elements

- continuing tests with alternate resin mixes l -

install full-flow, deep-bed demineralizers no later than the next refueling outage setting a goal for copper concentration in feedwater to less than or equal to 0.2 ppb at various hold points during power ascension Limerick SSER 8 10-1

For Limerick Unit 2, PECo will control water chemistry in accordance with the vendor defined startup test program.

Following the startup test program, a chemistry action plan similar to that for Unit 1 will be implemented. Feedwater copper concentration will be controlled during plant operation, chemical transients will be limited, on-line feedwater monitoring will be utilized and deep bed demineralizers will be installed at the first refueling outage.

The improved initial condition of the cladding material and condenser combined with the above chemistry action plan assures that the Unit 2 fuel reliability expectation is good.

l The staff endorses the licensee's action plan to prevent future CILC failures. The licensee has put forth a comprehensive effort to evaluate the l factors that may have caused the fuel failures in Limerick Unit 1. These investigations have contributed significantly to the body of knowledge on the . . ,

CILC phenomena in BWRs. On the basis of the information about CILC today, the staff concludes that, in combination, the actions being taken by the fuel manufacturer to improve fuel cladding metallurgy and the actions being taken by the licensee to control water chemistry will together prevent future CILC fuel assembly failures at Limerick Units 1 and 2.

Limerick SSER 8 10-2 -

14 INITIAL TEST PROGRAM The initial test program for Limerick Units 1 and 2 was approved in the staff's SER of August 1983. By letter dated June 21, 1988, the applicant submitted the proposed Power Ascension Test Program for Limerick Unit 2. The revised program was reflected in Revision 52 to the FSAR dated June 1988.

Supporting information was provided in a letter of May 12, 1989.

Based on our review of the modified program, the staff concluded that the proposed Limerick Unit 2 Power Ascension Test Program follows the guidance provided by Regulatory Guide 1.68, Revision 2, August 1978, " Initial Test Programs for Water-Cooled Nuclear Power Plants." The staff's letter of June 20, 1989 approved the proposed Power Ascension Test Program for Limerick Unit 2.

Limerick SSER 8 14-1

15 ACCIDENT ANALYSES 15.8 Anticipated Transients Without Scram The staff stated in the SER that Generic Letter 83-28 " Required Actions Based on Generic Implications at Salem ATWS Events" had been issued on July 8,1983 and that the staff would report the results of its review of the applicant's re.sponse in a supplement to the SER.

The applicant's response to Action Item 1.1 was provided in letters dated November 10, 1983 and supplemented on August 31, 1984 and June 7,1985.

This response was subsequently found acceptable as reported in SSER 5 for both Units I and 2.

The applicant's response to Action Item 1:2 was provided in leu.ers dated November 10, 1983 and supplemented on August 31, 1984 and June 7, 1985. The staff provided a summary of its review on November 7,1985 and issued a final SER on April 23, 1986 for both Units 1 and 2.

The applicant provided information on Action Items 2.1 and 2.2 on September 6, 1983, November 10, 1983 and May 8, 1984. The staff requested additional information on March 19, 1985. The applicant provided aciditieral information on April 1, 1985, May 29, 1985 and June 7, 1985. The staff again requested information on Action Items 2.1.1. and 2.2.1 on March 4, 1988.

PECo responded on April 13, 1989; however, the staff has not yet completed its review.

The applicant's response to Action Items 3.1.1, 3.1.2, 3.2.1 and 3.2.2 were provided on November 10, 1983. The staff reviewed this submittal and accepted it for Limerick Unit 1 in an SER on March 26, 1987. The information provided by the applicant was provided as applicable to both Units 1 and 2 and, therefore, the conclusions reached for Unit 1 in the March 26, 1987 SER are also applicable to Limerick Unit 2.

The applicant provided information in response to Action Items 3.1.3 and 3.2.3 on November 10, 1983. The staff requested additional information en March 19, 1985 which was provided by the applicant on May 29, 1985. On January 6,1986, the staff accepted the applicant's responses in an SER for both Units 1 and 2.

The applicant's responses to Action Items 4.5.1 and 4.5.2 were provided on September 6, 1983, November 10, 1983 and May 8, 1984. The staff reviewed these responses and accepted the positions regarding Action Item 4.5.1 for l Limerick Unit 1 in an SER dated March 26, 1987. However, the information provided by the applicant also included Action Item 4.5.2 which the staff has reviewed in conjunction with Action Item 4.5.1 and found acceptable.

Further, the information provided by the applicant was applicable to both Units 1 and 2 and therefore, the March 26, 1987 conclusion regarding Action Item 4.5.1 and the May 9, 1989 conclusion regarding Action Item 4.5.2 are applicable to both Limerick units.

Limerick SSER 8 15-1

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The applicant addressed Action Item 4.5.3 in its submittals of September 6, 1983, November- 10, 1983, and May 8, 1984. The staff reviewed these and a BWR Owners Group . submittal of January 31, 1985, but required additional-information as requested on March 19, 1985. The applicant responded on May 29 1985 and on June 7, 1985 endorsing the BWR Owners Group effort. The staff issued a generic SER.on July 15, 1987 and the applicant provided a plant specific response on November 5, 1987, in accordance with that SER. I The staff's letter of' June 21, 1989 advised PEco that the above responses  ;

were acceptable and that all 4.5 Action Items were considered completed. )

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Limerick SSER 8 15-2 l

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16 TECHNICAL SPECIFICATIONS Proposed Technical Specifications for Limerick Generating Station, Unit 2 were submitted by the licensee on June 3,1988. The proposed draft was based on the Limerick Unit 1 Technical Specifications as issued with the full power license NPF-39, on August 8, 1985 and subsequently amended. The licensee identified items in the proposed draft where Limerick Unit 2 was different from Unit I and modified the proposed draft to reflect these differences, as necessary.

The staff has completed its review of the applicant's proposed draft TechnicalSpecificationsandfoundthemtobeappropriately(basedupontheNUREG-0991),it design of the plant and in accordance with the staff's Si?

supplements and the safety evaluations prepared for the amendments to the Limerick Unit 1 Technical Specifications listed in Table 16.1.

The staff has also completed its review of the Off-site Dose Calculation Manual (ODCM) to be used in conjunction with the radiological effluent and radiological environmental monitoring Technical Specifications for Limerick Generating Station - Unit 2. Revision 5 of the ODCM for both Units 1 and 2 was submitted by Philadelphia Electric Company on March 30, 1988 and approved by the staff in our letter of December 9, 1988.

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Limerick SSER 8 16-1

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Table 16.1 dmerick Unit 1-Technics 1 Specification Amendments and Safety Evaluations Applicable to Limerick Unit 2  !

Original Amendment  :

Amend Submittal Issue  !

No. Date Date Subject 3 11/17/86 2/17/87 ICF and FFWF (L.C.2.C(13))

4 2/11/87 5/11/87 Source' Range Monitor Minimum )

Count Rate I

-5 1/'0/87 6/2/87 Increased Control Room Air Flow Rate 6 1/13/87 7/8/87 SGTS Service to the Refueling Area 7 4/3/87 8/14/87 Core Reload 8 3/23/87 3/10/88 Secondary Containment Integrity (WindSpeed) 10 11/18/87 10/31/88 Corporate and Station Staff  ;

Organizational Structure 11 7/19/88 11/7/88 Operational Conditions in Response to Generic Letter 87-09 12 7/7/88 12/7/88 Charcoal Adsorber Cooldown Mode 13 5/11/88 1/10/89 Containment Isolation Valves

& Deletes License Conditions 2.C.10 and 2.C.11 14 9/29/88 1/11/89 Reactor Yessel Head Spray Piping 15 11/1/88 1/18/89 Modification and Testing of Containment Penetrations 16 11/5/87 2/9/89 LPCI Injection Valve Diffe-rential Pressure Instrument Loops ,

17 12/14/88 3/22/89 Deletion of RSCS and Lowering of RWM Low Power Setpoint to 10 percent Limerick SSER 8 16-2

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V Table 16.1(continued) l Limerick Unit'1 Technical Specification Amendments and l Safety Evaluations Applicable to Limerick Unit 2 Original Amendment ,

Amend Submittal Issue  :

N o.. Date Date Subject l 18 3/23/89 -4/14/89 Revisions to the Degraded Grid Undervoltage Relay  ;

L Setpoints l 20- 8/29/86 5/19/89 Reporting Requirements on Primary Coolant Iodine Spikes 21 1/23/89 5/31/89 Pore Size of Filters Used During Testing of DG Fuel Oil 22 2/22/89 6/8/89 Standby Liquid Control System 23 1/27/89 6/14/89 Tie-in of Unit 2 SGTS and HVAC Systems 24 4/10/89 6/15/89 Incorporation of Unit 2 Power Supplies to Support Common Equipment in Unit 1 25 9/14/88 6/16/89 Increase in Spray Pond Level 26 1/27/89 6/19/89 Effluent Dose Limits 27 4/10/89 6/20/89 Tie-in of RHRSW and ESW Systems 28 4/10/89 6/20/89 T/S Cleanup 2/14/86 Clarification of T/S 9/9/88 Diesel Generator Testing 11/4/88 Single Loop Operation l Limerick SSER 8 16-3

7 17 QUALITY ASSURANCE 17.4 Conclusion In SSER-7, we stated that the staff was pursuing resolution of 17 questions regarding the detailed implementation of the QA program. The licensee's letter of May 5, 1989 provided acceptable responses to all of the staff's questions. The staff concludes that the quality assurance program for Limerick Units 1 and 2 complies with Appendix B to 10 CFR Part 50 and with the acceptance criteria in Section 17.2 of the Standard Review Plan (NUREG-0800). The licensee's QA program is acceptable.

17.6 Readiness Verification Program In SSER-7, the staff described the extensive Readiness Assessment and Readiness Verification Programs (RVP) being conducted y PECo to assess the design, construction and operational aspects of Limerici Unit 2. A major feature of the RVP was an independent design and construction assessment (IDCA). The IDCA consisted of two major programs - an independent design assessment (IDA) and an independent construction assessment (ICA). PECo's performance of the IDCA is essentially completed. The NRC's inspection of the programs is nearing completion. The following is a status report on both programs.

17.6.1 IndependentConstructionAssessment(ICA)

By letter dated February 10, 1989 PEco submitted the Construction Assessment Report for Limerick Unit I documenting the results of the l

Independent Construction Assessment (ICA) portion of the IDCA performed by SWEC. The assessment concludes "that the construction of safety-related systems and structures at Limerick 2 is satisfactory and generally in accordance with the drawings and specifications. Although SWEC identified deficiencies and documented them on 56 construction observation reports (CORs), the deficiencies were determined to have no impact on the components' ability to perform their intended safety function An NRC team inspected the results presented in the ICA report during an on-site inspection the week of March 27, 1989. The results of the inspection are documenteu in Inspection Report 50-353/89-200 issued May 17, 1989.

The NRC team reviewed the ICA report including the construction action items (CAls) and con n ruction observation reports issued during SWEC's ICA effort.

The team also evaluated a sample of Bechtel's responses and corrective actions

> concerning SWEC's CAls and CORs, as well as discrepancies identified by the NRC's independent inspection documented in Inspection Report 50-353/88-202 issued January 23, 1989.

Overall, the team found that SWEC properly followed the IDCA program for

reporting deficiencies and discrepancies on observation reports and, with l two exceptions, for accepting only appropriate responses to the CORs. The l team also found that, with only several exceptions, the responses provided I to SWEC were appropriate and corrective actions were appropriate and properly ,

implemented. PECo's responses and corrective actions for the NRC-identified I

I Limerick SSER 8 17-1 l

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. deficiencies also were, with several exceptions, appropriate and properly i implemented. In fact, the team was favorably impressed with the licensee's

-efforts to determine the scope of deficiencies identified by SWEC and the NRC.

Several issues remain open following the NRC inspection. The issues which require additional information from the licensee or additional review by the NRC are 1) verification by the licensee that the wire size used for motor leads on the operator for valve HV-52-2F001C is adequate for its application, '3

2) a clarification program as it relates bytothethelicensee licensee'sofresponse its construction quality) to COR-056, 3 NRC assurance review of additional information regarding resolution of grouted-in anchors that did not meet minimum embedment depths, and 4) an NRC-identified weakness associated with improperly performed quality control inspections. None of the issues which have not yet been resolved are of sufficient safety significance to impact the upcoming licensing decisions dealing with Limerick Unit 2. However, the issues should be resolved before the full power license is issued.

17.6.2 IndependentDesignAssessment(IDA)

The IDA report from SWEC was submitted to the NRC and PECo concurrently on April 12, 1989. As a result of their review, SWEC concluded that:

"the engineering and design of these safety-related systems and structures at Limerick 2 is satisfactory as reflected in the specifications, calculations, drawings, or other documents which provide the final Limerick 2 design. Although a few deficiencies were identified which have been or will be corrected, the design work reviewed satisfied Project licensing commitments and was determined to be technically adequate. Based on the results of this assessment and contingent upon satisfactory completion of remaining committed Project actions and resolution of_ six open items for which further Project action is necessary, SWEC has a reasonable basis to conclude that the Limerick 2 design of safety-related systems and structures complies with licensing commitments and is technically adequate. SWEC further concluded that the design process employed by the Project is an acceptableprocess,withtheexceptionoftheBalanceofPlant(B0P) setpoint calculations program which the Project is addressing appropriately. These conclusions are based upon the specific ' vertical slice' detailed review conducted by IDA and also on additional reviews by the Project which extended beyond these specific systems and structures. These additional reviews were conducted to bound the extent of concerns identified during the IDA review and provide #

confidence that affected designs outside the sample reviewed were technically adequate."

The IDA resulted in only one hardware change associated with replacing the Class 1E undervoltage relays as a result of inadequate degraded grid relay setpoints. Also the IDA identified a trend regarding calculational discrepancies which included numerical errors, inconsistencies between calculation and as-built design, undocumented engineering judgment, etc.

Many calculations were redone or new analyses were performed, but in all cases, the design margin was sufficient to avoid plant hardware changes.

Limerick SSER 8 17-2

To verify that SWEC had properly resolved the 118 design observation reports (DORS) that resulted from their review and to ensure that PEco completed or scheduled the associated corrective actions, the NRC conducted an inspection at the architect-engineer offices during the week of April 24, 1989. The inspection team reviewed approximately one-half of the design observation reports, all six of the SWEC identified open items, as well as the issue regarding the B0P setpoint calculations which is described below in item (6). The inspection team concluded that the IDA provided the needed additional design assurance that Limerick Unit 2 has met its licensing commtments. The inspection team's conclusion is based on its overview of the IDA program as well as PECo's providing an acceptable written response to the following items identified during the inspection conducted the week of April 24, 1989.

(1) Confirm that the action identified in the SWEC IDA report including the supplemtnal Hazards IDA Report have been completed. Also, for any action that will not be completed prior to fuel load, PECo was requested to provide a scheduled completion date and justification why its resolution does not impact granting of a full power operating license.

(2) Explanation of how the existing design basis documents will be used when making modifications to the facility, considering the calculational discrepancies identified as a trend in D0R-118.

(3) Verify the adequacy of bus voltage (i.e., verify that enough margin is present to prevent spurious separation of the onsite safety-related buses from the grid for the worst case voltage condition. The condition to be analyzed assumes a single source of offsite power supplying both units, the load tap changer is in the most unfavorable position prior to the event, a LOCA in one unit and safe shutdown in the other unit, and a dip in the grid voltage resulting from the loss of another offsite unit when the voltage on the grid was at a normal minimum value. This action was requested as a result of D0R 103.

(4) To avoid spurious tripping of the thermal overload relays associated with continuous running non-MOV 480-V motors, PEco is requested to evaluate the sizing of the thermal overload relay heaters for all safe-shutdown applications considering the effects of low voltage, high ambient temperatures and negative tolerances on the heaters. This action is requested as a result of DOR 039.

(5) Verify the vital battery end-of-life capacity considering the effect of a nondetectable high impedance fault on the ac side of the inverter.

This action is requested as a result of 00R 87 and design action item 261.

(6) In regard to D0R 021 concerning B0P Q-functional setpoints, PECo is requested to:

(a) confirm that all Bechtel established Q-functional B0P instrument and channel tolerances that support the operating basis surveillance / calibration frequencies have been reviewed and properly taken into account.

Limerick SSER 8 17-3

N (b) confirm that the weaknesses identified in the setpoint program regarding instrument tolerances, did not unduly affect the critical instrument tolerances required to support preoperational acceptance testing of safety-related B0P equipment.

PECo's letter of May 16, 1989, appropriately addressed items (1) and (2),

except for the supplemental hazards report and committed, in general terminology, to perform the required actions to resolve items (3), (4), (5) and (6) in a reasonable time frame.

The inspection team did identify one issue regarding the diesel-generator (DG) ground fault protection (ref. DDR-113) where PECo utilized a poor design. However, the design of the DG ground fault protection is not in violation of any licensing commitment or regulatory requirements, but is inconsistent with acceptable industry practice. Because the existing ground fault protection resistor is too small, the heat produced is greater and the risk of a fire is increased. {

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I Limerick SSER 8 17-4

18 HUMAN FACTORS ENGINEERING 18.1 Detailed Control Room Design Review j 18.1.1 Background The licensee submitted the Limerick Detailed Control Room Design Review Summary Report for the Limerick Nuclear Generating Station, Units 1 and 2, on 4 August 16, 1984 and Supplement No. I to the Summary Report on November 2, 1984. The staff's evaluation of the DCRDR had previously been reported in SSER-3 and SSER-5. The results of our previous evaluations concluded that the Limerick Unit 1 DCRDR met NUREG-0737, Supplement I requirements for work completed, but the DCRDR was incomplete in the following areas:

the use of function and task analysis to identify control room operator tasks and information and control requirements during emergency operations, a comparison of display and control requirements with a control room inventory to identify missing displays.

In the DCRDR Summary Rc art, the licensee requested a delay in the completion of the task analysis until June 1985. The staff's review and acceptance of the request was reported in SSER-3 and was reflected in Condition 2.C(8)(a) to License NPF-27 for Unit No. 1.

By letter dated October 27, 1988, the licensee indicated that the DCRDR, though it was performed on Unit 1, was applicable to Unit 2 because there are no physical differences between the Unit I and Unit 2 Control Rooms.

Consequently, the DCRDR Summary Report and supplements to the summary report that were submitted for Unit I are applicable to Unit 2. Also, the licensee indicated that the Human Engineering Discrepancies (HEDs) that were identified during the DCRDR, were applicable to Unit 2 and that those scheduled to be corrected prior to the end of the first refueling outage for Unit I would be corrected prior to fuel load for Unit 2.

By letter dated June 28, 1985, PEco provided Supplement No. 2 of the Summary Report for the Limerick Control Room Design Review. Supplement No. 2 documents the completion of the Limerick plant-specific task analysis and presents the results from the comparison of the display and control requirements with the control room inventory. Supplement No. 2 also contains information on the reassessment of certain HEDs and the final assessment of four HEDs that were held out for further review in previous supplements to the Summary Report.

( 18.1.2 Evaluation of Detailed Control Room Design Review Our evaluation of Limerick's DCRDR activities described ir Supplement 2 to the DCRDR Summary Report follows. The Science Applications International Corporation (SAIC) assisted the staff in performing this evaluation. Their Technical Evaluation Report is included in Appendix T.

Limerick SSER 8 18-1

18.1.2.1 System Function and Task Analysis l In Supplement 2 to the Sumary Report, the licensee describes the bases and the methods used to conduct the task analysis. The licensee's bases for the task analysis are the Limerick plant-specific Transient Response Implementation Plan (TRIP) procedures, which were developed from the Boiling Water Reactor Owners Group generic emergency operating procedure guidelines.

The TRIP procedures were prepared in accordance with a Procedure Generation Package which was reviewed and approved by the NRC. On this basis, the staff concludes that the TRIP procedures serve as an appropriate basis for conducting the task analysis.

The staff reviewed the task analysis method used by the licensee to identify information and control requirements for the operator to execute the TRIP precedures. The licensee used a task analysis work sheet to structure and record the results of the analysis. Attributes of each task considered the requirements for type of action, parameter, state, dynamic, indication, position, range, scale resolution, and response. Based on a comparison of the task analysis methodology used by the licensee with the NUREG-0800, Section 2.2 description of an acceptable methodology, the staff concludes that the licensee's methodology is acceptable.

18.1.2.2 Control Room Inventory The licensee conducted a comparison of the display and control requirements identified in the task analysis with the existing displays and controls in I the control room. The purpose of the comparison was to evaluate the availability and suitability of Limerick's control room instrumentation and controls. Missing instrumentation was identified in the verification of availability of required instruments. The suitability of required controls and displays was determined by comparing the required characteristics (identified from the task analysis) and NUREG-0700 based Task Analysis Guidelines to the actual controls an displays in the control room. Based on these comparisons, the licensee identified 36 HEDs.

Based on our review, we find the licensee's method for determining the adequacy of control room instrumentation to be acceptable.

18.1.2.3 Control Room Survey All DCRDR surveys have been completed, including a survey of the computer-based Safety Parameter Display System (SPDS) which had not been completed at the time the staff issued SSER-5. By letter dated February 25, 1986 the licensee indicated that they had performed a human factors survey of the SPDS and reported the results of the survey to the staff. Their report indicated that they did not identify any safety significant human engineering discrepancies related to SPDS. Based on the staff's evaluations in SSER-3 and SSER-5 and our review of the above submittal, we conclude that the licensee has satisfactorily completed the Control Room Survey portion of the DCRDR.

Limerick SSEP, 8 18-2

18.1.2.4 Assessment of HEDs The staff reviewed the assessment of the 36 HEDc identified from the task analysis and co'ntrol room inventory. This review focused on the safety significance assigned by the licensee to the HEDs. Of the 36 HEDs, 15 were assessed as Priority 2 (low safety significance); there were no HEDs identified with higher safety significance. He reviewed all of the Priority 2 HEDs and concluded that they had been properly assessed by the licensee.

We also reviewed HEDs on tasks needed to support rapid operator action. We concluded that these HEDs also had been properly assessed by the licensee.

The staff also reviewed the results of the licensee's assessment and resolution of four outstanding HEDs (Al-11, 15-04, 503-15, and S12-04). Our review focused on evaluating the priority code assigned by the licensee to the HED, and on the solution proposed to resolve the HED or the rationale to not fix the HED. In each case, the results from the staff's review agreed with the licensee's results.

18.1.2.5 Coordination of Control Room Improvements with Other Programs With successful completion of the system function and task analyses by the licensee, as evaluated by the staff in a previous section of this report, the staff has no further concerns regarding the licensee's coordination of control room improvements with other programs.

18.1.2.6 Staff Conclusions on DCRDR The staff concludes that the Limerick Generating Station, Units 1 and 2 DCRDR meets the NUREG-0737, Supplement I requirements for DCRDR. The TMI Action Item 1.D.I, Detailed Control Room Design Review, has been satisfied.

Condition 2.C(8)(a) of License NPF-27 and condition 2.C(5) of License NPF-39 have been met and have been satisfied for Limerick Nuclear Generating Station. The staff, at its discretion, may conduct a post-implementation audit of both Units 1 and 2 DCRDR.

18.2 Safety Parameter Display System 18.2.1 Background The Safety Parameter Display System (SPDS) for the Limerick Generating Station was reviewed by the staff in SSER-3 and SSER-5. All staff concerns were resolved, except for information from the licensee necessary to close Unit 1 license condition 2.C(5). This license condition required the SPDS to be operational within 30 days after completion of the 100-Hour Warranty Run.

Also, the licensee was to provide the staff with a summary of the problems encountered and resolutions implemented to make the SPDS operational and submit a report of the results from the field verification tests of the SPDS.

I Limerick SSER 8 18-3

i 18.2.2 Evaluation By letter dated February 25, 1986 the licensee indicated that the SPDS for Unit I was fully operational and that the 100-Hour Warranty Run had been completed January 28, 1986. The letter also summarized the problems i encountered during power ascension and the resolutions implemented to make the SPDS operational, and reported results of field verification tests. The staff considers Unit 1 license condition 2.C(5) to be closed.

By istter dated December 5, 1988 the licensee requested that Unit 2 also be allowed to com System (ERFDS)plete SPDS verification andthe during thePower relatedAscension Emergency Response Test Program Facility and Display initially declare the Unit 2 SPDS and ERFDS operational within 30 days after completion of the Unit 2 100-Hour Warranty Run. The licensee committed to provide the NRC with a summary of problems encountered, if any, during verification, the solutions implemented to make the SPDS operational, and to report the results from the field verification tests to the NRC within 30 days of declaring the SPDS Operational. We find this commitment to be acceptable. In addition, the licensee will submit a summary of the problems encountered during verification tests, solutions implemented, and report test results to the NRC within 30 days af ter completing the 100-Hour Warranty Run.

The licensee also is required to certify to the NRC the status of the SPDS for Limerick Generating Station Units 1 and 2 in accordance with Generic Letter 10 CFR 89-06(f)",datedApril"Tasi 50.54 12,Action 1989. Plan Item may, The staff 1.D.2at

- Safety its discretion, Parameter Display S verify full operation of the SPDS by conducting a post-implementation audit.

18.2.3 Staff Conclusions on SPDS Based on the staff's review of the information provided by the licensee, the staff concludes that the Limerick Generating Station SPDS for Units 1 and 2, may continue to be operated. The licensee will be required to operate the SPDS and ERFDS within 30 days after completion of the 100-Hour Warranty Run.

In addit. ion, the licensee is required to respond to the requirements of Generic Letter 89-06, for both Units 1 and 2.

Limerick SSER 8 18-4 1

19 REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS During 1983 the Advisory Committee on Reactor Safeguards (ACRS) performed a review of the application for an operating license for the Limerick Generating Station Units I and 2. The results of that review were included in the ACRS interim report to the NRC Chairman dated October 18, 1983. The interim report and a discussion of it by the NRC staff are included in '

Supplement No. I to the Limerick Safety Evaluation Report. The interim report indicated that the ACRS believed that fuel loading and operation up to five percent power could be carried out without undue risk to the health and safety of the public.

During 1984 the ACRS continued its review of the Limerick application. The ACRS held a combined meeting of its Subcommittee on the Limerick plant and its Subcommittee on Reliability and Probabilistic Assessment on October 9-10, 1984, in Washington, D.C. A further meeting of the Subcommittee on Reliability and Probabilistic Assessment was held on October 20, 1984, in Los Angeles, California.

On November 1-3, 1984, the full Committee considered the Limerick application at its 295th meeting in Washington, D.C. A copy of the Committee's report, dated November 6, 1984, to the NRC Chairman was included in SSER-4. Because of the uncertain schedule for Unit 2, the Comittee did not believe it appropriate to report on Unit 2 at that time.

Early in 1989 the ACRS began its additional review of Limerick Unit 2 material. On April 25, 1989, the Subcommittee toured the site and held an open public session nearby to consider the application. On May 4, 1989, the full committee considered the Limerick application at its 349th meeting in Bethesda, MD. A copy of the Comittee's report, dated May 11, 1989, to the NRC Chairman is included in Appendix S to this Supplement to the SER. The Committee's report indicated that the Comittee believed that, subject to the resolution of certain identified issues, there is reasonable assurance that the Limerick Generating Station Unit 2 can be operated at power levels up to 3293 MWt (100%) without undue risk to the health and safety of the public.

In the report of May 11, 1989, the ACRS Comittee commented that:

"For the past several years, it has been standard NRC practice to require extended periods of plant operation at very low power before approving operating at full power. Presumably, this has been done in the belief that it is safer than going more directly to higher power operation. It appears that if Limerick 2 is approved for full power operation, the Comission will require several months of operation at less than full power and that probably two months of this will be at about five percent of full power. However, we have yet to find anyone on the staff who has done or who knows of any systematic attempt to investigate whether there are any negative effects associated with this practice. Certainly, the units are not designed for extended operation at, for example, five percent of full power. And at least one licensee representative recently referred to operation at five percent of full Limerick SSER 8 19-1

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power as being ' uneasy,' although he did not believe there was anything unsafe about it."

The applicant estimates that it will take about 24 to 30 days after issuance of a low power license to load fuel and complete the precriticality testing.

After achieving criticality, it will take another 3 weeks to conduct reactor startup, complete low power testing and be~ ready to exceed 5% power. During this 3 week period, augmented attention will be given to all equipment in the plant. Barring any unforeseen developments or restrictions by the Commission, Limerick 2 will not be operated at very low power levels (less than 5%) for more than approximately 3 weeks. The staff considers this to be about the minimum time necessary to complete the required testing. The augefnted monitoring program should detect any equipment problems that might devt'op during this period.

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Limerick SSER 8 19-2

l' 22 FINANCIAL PROTECTION AND INDEMNITY REQUIREMENTS 22.1 Funds for Decommissioning On June 27, 1988, the NRC published in the Federal Register the final rule for 10 CFR 50.33(k) and 10 CFR 50.75, 10 CFR 50.33(k)(1) states, "Each application shall state for an operating license, for a production or utilization facility, information in the form of a report, as described in 50.75 of this part, indicating how reasonable assurance will be provided that funds will be available to decommission the facility." Section50.33(k)(a) requires each holder of an operating license to submit a decommissioning plan on or before July 26, 1990.

By letter dated February 7, 1989, tne licensee requested a schedular exemption from the applicable requirements of 10 CFR 50.33(k) and 10 CFR 50.75. Granting the exemption would authorize delay in submittal of the decomisioning report for Unit 2.

The licensee noted that the requested exenption is authorized by 10 CFR 50.12(a)(2)(v), which provides for issuance of an exemption when it would provide only temporary relief from the applicable regulation and the applicant has made good faith efforts to comply with the regulation. The requested exemption would provide only' temporary relief until July 26, 1990.

When promulgating the decommissioning fund rule, the NRC staff stated in NUREG-1221, "Sumary Analysis, and Response for Public Coments on Proposed amendments to 10 CFR Parts 30, 40, 50, 51, 70 and 72: Decommissioning Criteria for Nuclear Facilities," that "it is not expected that the rule amendment will affect pending operating licenses. Application of this rule to Limerick Unit 2, which is scheduled to receive its low-power operating license in the very near future, could result in undue hardship by unnecessarily delaying comercial operation.

The decommissioning funding report called for by 10 CFR 50.33(k) and 50.75(b) require a certification that decommissioning funds will be accumulated by an identified and acceptable method; and as part of the certification, a copy of the financial instrument obtained to satisfy the requirements is to be included.

In the request of February 7,1989, the applicant stated that:

" Financial planning responsive to the rule has been initiated and PEco intends to submit funding reports for both Limerick Generating Station ,

Units 1 and 2 on or before July 26, 1990. PEco's effort to comply is 4 apparent as the Company is already receiving current recovery of the LGS Unit I decommissioning cost and has established an external trust to accrue the necessary decommissioning funds. Furthermore, it should be noted that PEco already has a decommissioning cost estimate for LGS Unit 2 and intends to seek current recovery of this cost using the same methodology approved by Pennsylvania Public Utility Comission for LGS i Unit 1, upon comercial operation of the unit. Thus, Philadelphia Electric Company is making good faith efforts to comply with the rule's substantive requirements."

Limerick SSER 8 22-1

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The staff has reviewed-the licensee's request and its basis and finds them e acceptab.le. Therefore, the Comission has determined in accordance with L 10 CFR 50.12(a) that (1) a temporary exemption is authorized by law, will not present an undue risk to the public health and safety, and is consistent with the comon defense and security and (20 in this case, special circumstances are present. Therefore, the Comission hereby grants the following exemption:

Therefore,pursuantto10CFR50.12(a)(1),and50.12(a)(v), Limerick, Unit 2 is hereby granted a temporary exemption from the schedular requirements of 10 CFR 50.33(k)(1) and is required to submit the decommissioning plan for Limerick, Units 1 and 2 on or before July 26, 1990.

The Comission has determined that the granting of this exe.nption will not result in any significant environmental impact. The environmental assessment and finding of no significant impact was published April 19, 1989 (54 FR 15851).

Limerick SSER 8 22-2

l APPENDIX A CHRON0 LOGY LIMERICK GENERATING STATION, UNIT 2 February 24, 1989 Letter from applicant transmitting changes to Final Safety Analyses Report (FSAR) Since Unit I licensing.

March 1, 1989 Letter from applicant transmitting Final Safety Analysis Report (FSAR), Rev. 57, Fire Protection Evaluation Report, Rev. 11, Design Assessment Report, Rev. 11.

March 7, 1989 Letter f' rom applicant regarding Equipment Qualification Program.

March 7, 1989 Letter from applicant, response to Nuclear Regulatory Commission (NRC) Bulletin No. 88-07, Supplement 1:

" Power Oscillations in Boiling Water Reactors (BWRs)."

March-15, 1989 Letter to applicant transmitting Systematic Assessment of Licensee performance (SALP) Report Number 50-353/87-99.

March 21, 1989 Generic Letter 89-02 to all holders of Operating Licenses (OLs) and Construction Permits (cps) for nuclear power reactors regarding actions to improve detection of counterfeit and fraudulently marketed products.

March 21, 1989 Letter from applicant forwarding items with potential impact on NRC Safety Evaluation Report (SER) from FSAR, Appendix B, per February 24, 1989 letter.

March 21, 1989 Letter from applicant informing that utility will conduct self-assessment of performance of facility during power ascension program starting with initial fuel loading.

March 24, 1989 Generic Letter 89-03 to all power reactor licensees and applicants for OL regarding operator licensing national exam schedule.

March 24, 1989 Letter to applicant discussing closecut of multi-plant issues regarding NRC Bulletins 88-005 and 88-010, specifically requiring certain utility actions scheduled for week of April 24, 1989.

Limerick SSER 8 Appendix A

March 28, 1989 Letter to applicant forwarding proof and review Technical Specifications (TSs), for comment by April 17, 1989.

March 28, 1989 Letter from licensee forwarding revision to physical security plan. Revisionwithheld(Reference 10CFR 2.790 and 10 CFR 73.21).

March 29, 1989 Letter from applicant forwarding proprietary GE NEDC-31629P, " Single-Loop Operation Analysis for Limerick Generating Station Unit 2." Utility will continue to work with NRC to develop appropriate TS to reflect analysis, Re (Reference IP CFR 2.790(b)(port 1)). withheld March 31, 1989 Letter to applicant forwarding environmental assessment and 1989finding requestof forno significant exemption impact from regarding) 10 CFR February 7, 50.33(k (1),

extending time required for submittal of report regarding availability of funds for decommissioning.

Maru. 31,1989 Letter from applicant forwarding results of fracture toughness analysis completed for Unit 2 using methods approved in Regulatory Guide (RG) 1.99, Revision 2. ,

Utility authorized revision of pressure / temp curves per '

Revision 2. FSAR will be amended following NRC approval of analysis.

March 31, 1989 Letter from applicant forwarding additional information regarding preservice inspection program Specification 8031-P-504, per NRC March 8, 1989 telephoneconversation(telecon). Revision 1 Code Case N307 approved by NRC per RG 1.147, Revision 6.

March 31, 1989 Letter from applicant forwarding response to NRC Bulletin 88-007, " Power Oscillation in BWRs," and revised response to Supplement I for plant. Licensed reactor operators and shift technical advisors thoroughly briefed on LaSalle event described in bulletin.

March 31, 1989 Letter from applicant forwarding response to NRC Bulletin 88-005, " Nonconforming Material Supplied by Piping Supplies, Inc. at Folsom, NJ and West Jersey Manufacturing Co. at Williamstown, NJ." Materials installed in facility meet ASME Code requireinents.

Linerick SSER 8 Appendix A

April 3, 1989 Generic Letter 89-04 to all holders of Light Water Reactor (LWR) Operating Licenses (OLs) and Construction Permits (cps) regarding guidance on developing acceptable inservice testing programs.

April 3, 1989 Letter from licensee forwarding proprietary summary of evaluation of Unit I fuel failures which occurred during Cycle 2 operation, fuel failures due to crud induced localized corrosion. Pro withheld (Reference 10 CFR 2.790(prietary a)(4)). report April 3, 1989 Letter from licensee responding to NRC Bulletin 88-010, " Nonconforming Molded-Case Circuit Breakers."

All molded-case circuit breakers being maintained as stored spares for use have been verified traceable to circuit breaker manufacturing.

April 4, 1989 Generic Letter 89-05 to licensees of all power reactors and applicants for reactor operator license under 19 CFR Part 55 regarding pilot testing of fundamentals exam.

April 6, 1989 Letter to applicant granting NRC approval for use of ASME Code Cases N307, N356, N435 and N460 for preservice inspection program, per March 31, 1989 request.

April 10, 1989 Letter from applicant confirming that mechanical equipment qualification program for environmental qualification of active safety-related mechanical equipment at Unit 2 consistent with program utilized at Unit 1.

April 13, 1989 Letter from licensee responding to Generic Letter 83-28, Items 2.1 and 2.2 regarding NSSS vendor interface program. Maintenance of controlled vendor manuals will be controlled by issuance of nuclear group administrative procedure.

April 18, 1989 Letter from applicant responding to request for information on Probabilistic Risk Assessment.

April 18,1989 Letter from applicant referencing November 3, 1987 letter regarding compliance with A1YS Rule 10 CFR 50.62.

l April 21, 1989 Letter to applicant transmitting Systematic Assessment of Licensee Performance (SALP) Report Number 50-353/87-99.

Limerick SSER 8 Appendix A

April 27, 1989 Letter from licensee responding to Generic Letter 89-05.

April 27, 1989 Letter to applicant reemphasizing licensees responsibility to assure that contractors and subcontractors permit employees to contact NRC regarding concerns about potential safety issues, without restrictions.

April 28, 1989 erating Generic letter 89-07 to all licensees of op(0Ls) and plants, holders applicants for Operating of construction (Licenses permits cps) regarding power reactor safeguards contingency planning for surface vehicle bombs.

April 28, 1989 Letter from licensee revising utility's August 2, 1988 response to Generic Letter 88-01, "NRC Position on IGSCC in BWR Austenitic Stainless Steel Piping." NRC positions on IGSCC will be incorporated into inservice inspection program.

April 23, 1989 Letter from licensee responding to request for additional information regarding 4 kV bus undervoltage relay setpoint calculations, per March 23, 1989 Technical Specification Change Request 89-03.

May 2, 1989 Letter to licensee forwarding safety evaluation accepting N2H nozzle-to-safety end weld for continued service during Cycle 3. Concurs with analysis ..

estimating that growth of flaw will be minimal with improved water chemistry and should not exceed allowable limits.

May 2, 1989 Generic Letter 89-08 to all holders of Operating Licenses (OLs) or construction permits (cps) for nuclear power plants regarding erosion / corrosion-induced pipe wall thinning.

May 5, 1989 Letter from licensee forwarding utility responses to informal request for additional information on plant Quality Assurance (LjA) program, for review.

May 5, 1989 Letter from licensee forwarding response to request for additional information regarding plant Quality Assurance (QA) program. FSAR will be revised in future amendment.

Limerick SSER 8 Appendix A

May 8, 1989 Generic Letter 89-09 to all holders of Light Water Reactor (LWR) Operating Licenses (OLs) regarding ASME Section III Component Replacements.

May 8, 1989 Letter to licensee responding to September 22, 1988 request for exemption from requirements of 10 CFR 50, Appendix J, Paragraphs II.H.4 and III.C for 54 valves. Proposed leak testing program for valves under GDC 54 acceptable.

May 9, 1989 Letter to applicant advising that Revision 1 to plant preservice inspection program for reactor coolant pressure boundary and for ASME Code Class 2 and 3 system and components acceptable and in compliance with 10 CFR 50.55a(g)(2). Evaluation included in NUREG-0991, Supplement 7 May 9, 1989 Letter to applicant advising that April 3, 1989 l response to NRC Bulletin 88-010, " Nonconforming Molded-Case Circuit Breakers," acceptable.

May 9, 1989 Letter to applicant forwarding safety evaluation accepting utility's August 19, 1988 submittal of turbine system maintenance program.

May 9, 1989 Letter to applicant advising that utility response to Generic Letter 83-28, " Reactor Trip System Reliability on Line Testing," acceptable.

May 9, 1989 Letter from applicant confirming information provided during April 11, 1989 telephone conference regatfing revisions to Section 9.4.5.2.1.a and 9.4.2.2.2.c of plant FSAR. Revisions cover environmental qualification of mechanical equipn.ent.

May 12, 1989 Letter from applicant advising that test simplification and elimination 2 in power ascension testing program based on information provided by NSSS General Electric (GE).

May 12, 1989 Letter from applicant providing confirmation of information discussed in April 25, 1989 telephone conference regarding SPDS for facility. SPDS contained within plant monitoring system and based on General Electric (GE) response information system described in Topical Report NEDE-30284-P.

Limerick SSER 8 Appendix A

May 15, 1989 Letter to applicant forwarding notice of utility antitrust Operating License (0L) review and no significant change finding.

l May 16, 1989 Letter from applicant forwarding original of May 16, l

1989 response to NRC request regarding review of independent design and construction assessment program performed during week of April 24-28, 1989.

May 19, 1989 Letter to applicant forwarding final draft Technical Specifications for review. Certification that Technical Specifications reflect as-built construction and design described in FSAR requested by June 2, 1989.

May 19, 1989 Letter to licensee forwarding Amendment 20 to License NPF-39 and safety evaluation. Amendment revises Technical Specification by changing reporting requirements for iodine spiking from short term report to item to be included in annual report.

May 23, 1989 Letter to licensee forwarding request for additional information regarding severe accident mitigation design alternatives, per NEPA. Response requested within 30 days of letter receipt.

May 23, 1989 Letter from licensee requesting exemption from 10 CFR 50.44(c)(3)(ii)(B) to power hydrogen recombiner containment isolation valves as stated.

May 25, 1989 Letter from applicant forwarding Volume II, Book I to

" Limerick Generating Station, Unit 2 Independent Design and Construction Assessment" and supplemental information regarding instrument setpoints discussed in previous status of Design Observation Report 021.

May 26, 1989 Letter to licensee forwarding director's decision regarding Ohio Citizens For Responsible Energy 10 CFR 2.206 petition regarding March 9, 1988 power oscillation event at LaSalle, Unit 2.

Limerick SSER fi Appendix A

May 26, 1989 Letter from licensee application for amendment to License NPF-39, consisting of Technical Specifications Change Request 89-01, correcting administrative errors in Technical Specifications.

May 31, 1989 Letter to applicant forwarding environmental assessment and finding of no significant impact regarding exemption from 10 CFR 50 requirements for containment airlock testing.

Limerick SSER 8 Appendix A

APPENDIX S REPORT OF THE ADVISORY COMMITTEE ON REACTOR SAFEGUARDS l

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Limerick SSER 8 1 Appendix S

[ \ UNITED STATES NUCLEAR REGULATORY COMMISSION

[J ADVl80RY cotamTTat ON REACTOR SAFa0UARDS wAa m otou,s,c.asene May 11, 1989 The Honorable Lando W. Zech, Jr.

Chairman U.S. Nuclear Regulatory Connission Washington, D.C. 20555

SUBJECT:

OPERATING LICENSE APPLICATION FOR LIMERICK GENERATING STATION, UNIT 2 During the 349th meeting of the Advisory Connittee on Reactor Safeguards.

May 3-6.1989 we reviewed the application of the Philadelphia Electric Company, the Applicant, for a Itcense to operate the Limerick Generating Station. Unit 2. Our Subcommittee on Limerick 2 toured the facility on the morning of April 25, 1989 and met in the afternoon, in PhiladelpF.ia, to consider this application. During our review, we had the benefit of discussions with representatives of the Applicant and the NRC staff. We also had the benefit of the documents referenced.

In the ACRS report, dated November 6,1984, to then Chairsan Nunzio J.

Palladino, the Committee commented on the application for an operating license for Limerick Unit 1. In that report the Committee noted that, although the Applicant had requested an operating license for both Units 1 and 2. the Committee felt that it was not #;propriate to comment on Unit 2 at that time because of the uncertain schedule for construction and operation of Unit 2.

Although Unit 2 is being considered for an operating license some four-and-l a half years after the approval of an operating license for Unit 1. the two units have the same rated power level, use the same model nuclear steam supply system, and are generally very similar.

In the course of our review, we discussed management and staffing of Unit

2. Recent changes in the Applicar.t's management have resulted in the location on-site of a vice president responsible for the Limerick Station.

A number of those individuals responsible for testing and startup of Unit 2 have gained experience on Unit 1. This experisoved group appears to be conductinj a well crganized and effective test program and to be accomplis ting a smooth transition in the turnover of responsibilities to the crew tatt will be responsible for operation. The nest recent Systematic Asscssment of Licensee Performance ($ ALP) rating by the NRC staff gives the management of this group an unusually high rating for its

The Honorable Lando W. Zech, Jr. 2- May 11, 1989 l

perfora nce. We found no reason to question the experience, training, or capability of the personnel who will be responsible for operating Unit 2.

In the ACRS report, dated November 6.1984. the Cossnittee mentioned that a probabilistic risk assessment (PRA) had been performed for Unit 1. The Applicant.now has its own staff of PRA practitioners who, with some outside assistance, have not only revised the PRA for Unit 1 to reflect changes in the plant and in operat4ng practices, but have also performed a PRA for Unit 2. Among changes that have taken place since the earlier PRA was per-fonned are the installation of van:s for the containments of both units and the adoption of Revision 3 of the Boiling Water Reactor Emergency Procedure Guidelines. The PRA group, in a 1988 update, reports a calculated core damage frequency of 6.69E-6 per year for each unit. This is slightly less than half that calculated when the original FRA was perfonned. It should be noted that this does not inc1Lde any contribution from seismic events which have been a significant con ributor in other contemporary PRAs. It appears that the Appitcant's organization is using the insights from PRA in training and in planning their maintenance program. They intend to use these insights also in the formulation of their accident management pro-gram.

In the course of preparing the organization and plant for operation, the Applicant performed a Readiness Program Assessment "to assess the adequacy of existing licenses programs and processes." and retained a contractor to perform a Readiness Yorification Pro!; ram to provide "a comprehensive integrated process to assess the des < gn, construction, and operational aspects of Unit 2." The NRC staff then reviewed both of these assessments.

Both the Applicant and the NRC stsff reported that these reviews provided convincing evidence that the plant is ready for startup.

In the ACRS report. dated November 6,1984, the Committee recommended also that Unit I receive special attention in the NRC staff's resolution of the unresolvedsafetyissue(USI)onsystemsinteractions. We reconsnended also that special attention be given to the identification of sry risk outliers associated with satanic events. These issues are being dealt with generically.

Limerick Unit 1 has had some difficulty with corrosion of fuel claddings however, this does not appear to be a serious safety problem. The Appli-cant proposes some changes in p* ant equi 3rocedures which should make the corrosion less likely.pment and operatingAlthough some insigits app to have been developed that may make the problem less severe or pertaps even eliminate it, the results of applying these insights are not yet available.

For the past several years, it has been standard NRC practice to require extended periods of plant operation at very low power before approving operation at full power. Presumably, this has been done in the belief that it is safer than going more directly to higher power operation. It appears that if Limerick 2 is approved f,1r full power operation, the Commission

The Honorable Lando W. Zech, Jr. -3 May 11. 1989 will require several months of operation at less than full power and that probably two months of this will be at about five percent of full power.

However, we have yet to find anyone on the staff who has done or who knows of any systematic attempt to investigate whether there are any negative effects associated with this practice. Certainly, the units are not designed for extended operation at. for example, five percent of full power. And at least one licensee representative recently referred to operation at five percent of full powar as being ' uneasy." although he did not believe there was anything unsafe about it. We have no evidence that it is unsafe, but do know of instances in which operation at low flow has produced excessive wear in check valves and in which operation at low swer has produced axcessive vibration in a feedwater pump not designet for extended operation at low flow. Our principal concern stems from the memory that the operators of the Chernobyl plant. Unit 4, were unaware of the dangers of operation at low power, whereas a careful analysis would have convinced them that this was undesirable. It appears to us that if the practice of extended operation at low power is to be continued, some systematic search for possible harmful effects should be performed.

We believe that, subject to satisfactory completion of construction and preoperational tasting, there is reasonable assurance that the Limerick Generating Station, Unit 2. can be operated at power levels up to 3293 MWt without undue risk to the health and safety of the putlic.

Mr. James C. Carroll did not participate in the Comittee's review of this matter.

Sincerely.

4 Forrest J. Remick Chaiman Referencet

1. u.s. Nuclear Regulatory comission. Region I. Systematic Assessant of Licensee Performance Board Report. Philadelphia Electric Company, Limerick Generating Station. Unit 2. Inspection Report 50-353/87-99.

February 22, 1989

2. U.S. Nuclear Regulatory Comission. NUREG-0991. Supplement No. 7 "Safetv Evaluation Report Related to the Operation of Limarick Gener-ating Station. Units l, and 2.* April 1989

The Honorable Lando W. Zech, Jr. May 11, 1989

3. Letter dated January 23, 1989 from Gus C. Lainas. U.S. Nuclear Regula-tory Commission, to G. A. Hurger. Jr., Philadelphia Electric Company.

Subject:

Inspecticn of Independent Construction Assessment. Limerick Generating Station. Unit 2; Inspection Report Number 50-353/88202

4. Public Statements provided during the April 25, 1989 meeting of the ACR$' Limerick 2 Subcommittee from the following:

a. Marvin I. Lewis, Limerick Ecology Action

b. Richard %ers, Citizens' League for Energy Awareness and Resources

c. Ruth Miner, Citizens for Environmental Rights

d. Emanuel Mendelson, Citizens for Environmental Rights

e. Phyllis Gilbert, Sierra Club. Philadelphia, Pennsylvania

APPENDIX T REPORT ON CONTROL ROOM DESIGN REVIEW BY SCIENCE APPLICATIONS INTERNATIONAL CORPORATION FOR LIMERICK GENERATING STATION UNITS 1 AND 2 l

l l

Limerick SSER 8 1 Appendix T

TECHNICAL EVALLIATION REPORT FOR PHILADELPHIA ELECTRIC COMPANY'S LIMERICK GENERATING STATION r

CONTROL ROOM DESIGN RE,IEW SUPPLEMENTAL REPORT 2 0F THE FINAL REPORT OF JUNE 1984 IKTRODUCT!0N Tnis technical evaluation report documents the findings of Science Applications International Corporation (SAIC), consultants to the Nuclear Regulatory Commission for the review of Philadelphia Electric Company's Limerick Generating Station Control Roos Design Review Supplemental Report 2.

of the Final Report of June 1984 (Reference 1). The SAIC evaluation team members were familiar with nuclear power plant control rooms ar.d experienced in evaluating detailed control room design reviews (DCRDRs). The purpose of the evaluation was to:

1. To determine if a qualified multidisciplinary team performed tb Limerick function and task analysis activities at Limerick.

2. To determine if the function and task analysis methodology used in the Limerick DCRDR satisfied the requirements of Supplement 1 to NUREG-0737 (Reference 2).

3. To determine if methodology used to compare display and control requirements with the control room inventory satisfied the require-ments of the Supplement 1 to NUREG 0737.

l 4. To determine if the numan engineering discrepancies identified during the Limerick DCRDR task analysis activity reflect a thorough exercise of the function and task analysis methodology.

l l

1

5. Determine if the proposed solutions to the Limerick function and task analysis human engineering discrepancies. satisfy the require.

ments of Supplement 1 to NUREG-0737 in terms of content and sched-uling.

This evaluation was conducted relative to the recivirements of Supple-ment I to NUREG-0737. Additional guidance was providad by NUREG 0700 '

(Reference 3) and Section 18.1. revision 0, of NUREG-0800 (Reference 4).

This report provides the results of the SAIC evaluation.

1. Establishment of a. qualified multidisciplinary review team.

The licensee's function and task analysis review team was composed of:

ces nuclear /ISC engineer, two human factors specialists, and two licensed operato'rs Limerick Unit 1. In addition. the team was supported by other Itcensed operators and engineering personnel. This description of the Limerick function and task analysis review team conforms to the team com-position and qualifications guidance provided in NUREG-0800. It is there-fore our conclusion that a qualified mult1 disciplinary team was used to perform the Limerick function and task analysis activities.

2. Use of function and task analysis.

The licensee's submittal provides a description of the function and task analysis methodology along with an example task analysis worksheet, a kay to the task analysis worksheet, a team briefing for task analysis activities and a tast analysis guide. The Limerick methodology uses their plant specific Transient Response Implementation plan (TR!p) procedures which were developed from the Boiling Water Reactor Owners Group (BWROG) generic emergency procedures guidelines as the functional and systems basis for conducting the task analysis. The licensee stated that the procedures generation package used to develop the TRIP procedures was found to be I acceptable to the NRC in the Limerick Safety Evaluation Report. Supplement

2. Section 133.2.3. NUREG-0800. Section 2.2 states the emergency operating procedures te6anical guidelines developed by the four owner's groups provide the functional and system bases for conducting the task analysis. There-fore, it is our judgment that the TRIP procedures represent appropriate functional and systems bases for conducting the task analysis.

2

L l The NUREG-0800. Section 2.2 description of an acceptable task analysis l' . methodology begins with an analysis of the functions to be performed by the systems in responding to transients and accidents in order to define and i describe the tasks the operators are espected to perform. In Limerick's I team briefing for task analysis, the task analysis team members are instruc-ted to " concentrate on the system requirements and actions. Think in terms of what the systems are doing and what is required to be done in the systems." The licensee's methodology as described in their submittal is used to identify operator information and control requirements independent of the existing control room.

According to licensee's methodology, the information and control requirements along with the required instrument and control characteristics needed to support the TRIP are entered on the left side of the Task Analysis

! Worksheet by the task analysis team before control room specific information is recorded on the form. This conforms to guidance provided in NUREG-0800 Section 2.2 regarding the a priori nature of the task analysis.

A description of the characteristics which were entered on the Task Analysis Worksheets is provided in the licensee's Key to Task Analysis Worksheet. The entries on the worksheets included tasks analysis of operator decision making, action, and feedback type tasks. For each task that was analyzed, the task analysis team analyzed the operator response time criticality. In addition, other task analysis worksheet entries include the requirements for type of action, parameter, state, dynamic.

indication, position, range, scale resolution, and response.

3. Comparison of display and control requirements with control room inventory.

The availability and suitability of Limerick's control room instrumen-tation were evaluated by comparing the display and control requirements documented on the Task Analysis Worksheet with the actual control room instrumentation. Missing instrumentation was identified in the verification of availability of required instruments. The suitability of required con-trols and displays was determined by comparing the required characteristics and MUREG-0700 based Task Analysis Guidelines to the actual controls and displays in the control room. The result was a list of 36 human engineering ,

3

discrepancies which fell into the main categories of instrument suitability and resolution characteristics.

It is our judgment that the methodology described by the licensee should result in a satisfactory comparison of display and control require-ments with the control room inventory in order to satisfy the requirements of Supplement 1 to NUREG-0737

4. Review of human engineering discrepancies identified during task analysis.

The purpose of our review of human engineering discrepancies (HEDs) was to determine if they reflected a thorough exsrcise of the functicn and task analysis methodology. The type of HEOs which should be identified in task analysis are HEDs which would otherwise not be identified in the WUREG-0700 checklist surveys. Typically, task analysis should be ned to identify missing displays (NUREG-0700 guideline 6.1.1.1.a), displays with scale resolution discrepancies (NUREG-0700, guideline 6.5.1.2), and displays which are unsuitable for the required operator task (NUREG-0700 guideline 6.5.1.1.b). . .

Appendix B. Tehm Brief for Task Analysis contains instructions to the task analysis team to aid them in recognizing the human engineering suita-bility considerations. The task analysis team was referred to Task Analysis Guidelines, which were ' intended for frequent reference by the team during the verification phase." These Task Analysis Guidelines are included as Appendix 0 to the licensee's submittal. The Task Analysis Guidelines pro- -

vide an appropriate list of MUREG-0700 based human engineering principles to evaluate during the task analysis activity.

Our review of the 36 HEOs identified during the task analysis activity indicatid that the task analysis team members followed the guidance provided in the Task Analysis guidelines by identifying displays which were missing, unsuittele displays and those displays which lacked the appropriate resolu-tion for the task. Of particular importance in the exercise of task analysis is the identification of missing controls and displays. The team identified a missing isolation alarm for group 11 isolation, suppression pool low level alarm, and drywell temperature alarm indication.

4

The task analysis team also identified 18 HEDs for tasks which the operator performs rapidly under time pressure. Of these HE.Ds.10 were assessed ~as priority 2 items which will be corrected during the first or second refueling outage., Of particular interest is the fact that the task analysis team determined that the automatic depressurization timer did not provide the cperator with sufficient feedback to indicate that the logic had been reset. A final design solution is scheduled for this HED during the first refueling outage.' This indicates that the task analysis and assessment methodologies were appropriately used to identify and correct safety significant HEDs.

It is SAIC's judgment that the HEDs identified during the task analysis activity reflect an appropriate and thorough exercise of the licensee's methodology.

5. Proposed control room changes and schedules for implementation.

SAIC review of the proposed control room changes concluded that the licensee does intend to correct discrepancies assessed as significant HEDs.

In addition, the licensee did provide an appropriate schedule for implemen-tation of each of the proposed changes.

CONCLUSION The licensee's submittal described a task analysis methodology which should produce results which meet the requirements of Supplement 1 to NURM-D737. The licensee-established an appropriate multidisciplinary team. &c.

appropriate methodology for the system function and task analysis, and an acceptable control room inventory. The human engineering discrepancies identified during the task analysis activity reflect a appropriate applica-tion of task analysis activities. Furthermore, the proposed control room changes described reflect an intent by the licenset to correct the human engineering discrepancies.

5

REFERENCES

1. Letter from J.S. Kemper. Philadelphia Electric Company, to W.R. 8ulter.

DL. MRC. dated June 28,105, subject: ' Limerick Generating Station.

Units 1 and 2. Limerick Control Room Design Review. Final Report.

Supplement 2."

2. NUREG-0737 Supplement 1 " Requirements for Emergency Response Capability.' USNRC Washington, D.C. December 1982, transmitted to reactor licensees via Generic Letter 82-33. December 17, 1982.

3. NUREG-0700. " Guidelines for Control Room Design Reviews.' USNRC.

Washington. D.C., September 1981.

4. NUREG-0800. " Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants." Section 18.1 Rev. O USNRC.

Washington. D.C., September 1984.

6

APPENDIX U ERRATA TO THE SAFETY EVALUATION REPORT FOR THE LIMERICK GENERATING STATION SSER-7 Location Current Wording Revision Pg 4-2, line 47 "296" "307" Pg 4-2, line 48 "5" "4" Pg 4-3, line 3 "26" "53" Pg 4-3, line 3 "12" "8" Pg 4-3, line 5 "two-thirds of" deleted Pg 4-3, line 25 " heat" "sublot" Pg 4-3, line 29 "1569" "1138" Pg 4 3, line 38 "doner" " donor" Pg 4!4, line 4 "42" "44" Pg 8-2, line 1 "in conformance with "in conformance with the Institute of" the recommendations of the Institute of" Pg 9-1, line 14 "100-percent capacity "100_ percent loop-pumps per loop" capacity pumps" Pg 10-2, line 2 " replace the filter" " replace or supplement the filter" l Pg 10-2, line 2 "with full-flow" "with downstream full-flow" Pg 20-3, line 2 " Cycle 1." " Cycle 2, relative to Cycle 1."

Pg 10-3, line 4 " average values," " average values over 3 days periods,"

Pg 10-3, line 6 " Iron" " Soluble iron" Pg 10-3, line 48 " deep-bed " deep-bed demineralized are" demineralizers in series with filter demineralizers are" Pg 10-3, line 49 "

. ineralizers." "demineralizers alone."

Limerick SSER 8 1 Appendix U

Appendix 0 (con't)

ERRATA to the Safety Evaluation Report for the Limerick Generating Station SSER-7, Appendix Q Location Current Wordina Revision Pg 4, line 23 "NIIB" "N11B" Pg 4, line 26 "NIIB" "N11B" Pg 6, line 13 "DLA-212-2 FWI" "DLA-212-2 FWi" -

Pg 6, line 14 "DLA-212-3 FWI" "DLA-212-3 FWi" Pg 6, line 15 "DLA-21Z-4 FWI" "DLA-212-4 FWi" Pg 6, line 16 "DLA-210-1 FWI" "DLA-210-1 FWi" Pg 6, line 23 "DCA-204-1 FWI" "DCA-204-1 FWi" Pg 6, line 24 "DCA-204-2 FWI" "DCA-204-2 FW1" Pg 6, line 34 "DCA-201-1 FWII" "DCA-201-1 FW11" Pg 6, line 36 "DCA-201-1-10 SWI" "DCA-201-1-10 SWi" Pg 6, line 38 "DCA-201-3 FWIO" "DCA-201-3 FW10" Pg 7, line 37 " Class 9" " Class 1" Pg 11, line 7 " EBB-235-1 FWI" " EBB-235-1 FWi" Pg 11, line 13 " EBB-201-1-10 SW1 LDRI" " EBB-201-1-10 SW1 LDR1" Pg 11, line 15 " EBB-2 02-1-10 SWI LD" " EBB-202-1-10 SW1 LD" Pg 11, line 17 " EBB-2 03-1-10 SWI LD" " EBB-203-1-10 SW1 LD" Pg 11, line 19 " EBB-2 04-1-11 SWI LD" " EBB-2 04-1-11 SW1 LD" Pg 14, line 19 "DLA-205-1 FWI" "DLA-205-1 FW1" Pg 14, line 23 "DBA-207-1-5A WBA" "DBA-207-1-5A WBA" Pg 14, line 26 "DCA-20I-1 FW7" "DCA-201-1 FW7" Pg 14, line 27 "DCA-201-2 FWI" "DCA-201-2 FW1" Pg 14, line 34 "DBB-203-1 FW10Rl" "DBB-203-1 FW10R1" Pg 14, line 37 "DBB-204-1-IA SW2" "DBB-204-1-1A SW2" Pg 14, line 38 "DBD-203-1 FWIRI" "DBD-203-1 FW1R1" Pg 14, line 40 "DBD-204-1 FWIRI" "DBD-204-1 FW1R1" Pg 14, line 41 "DBD-204-1 FW3RI" "DBD-204-1 FW3R1" Pg 14, line 42 "Reoust:" " Request:"

Pg 14, line 43 " complete Code" " complete augmented" Limerick SSER 8 2 Appendix U - - - -

1. P T ER U.S. NUCLEAR REGULATORY COMMisslON NR FonM 335 arid Adoendum Numtsrs,if any i NRCM 1102' no. 22n BIBLIOGRAPHIC DATA SHEET

,s,,i,,,,,,,,,,,,,,,,n,,,,,,,,,

NUREG-0991 Su lement No. 8

2. m LE AND SU6M LE Safety Evcluation Report related to the operation of Limerick Generating Station, Units 1 and 2. 3.

MONTH DATE REPOR1 PUBLISHED vtAR JUNE 1989

4. FIN OR GH ANT NUMBER 6 TYPE OF REPORT

b. AUTHOR (s)

Supplement No. 8 to the Safety Evaluation Rpt.

7. VE R COD COV E R E D tinclusore ooress B.,,P.E R R F O,,s

.. ,,,i.. , MING,,ORG ANIZ AT ION - N AME AND ADDRESS to' NRC. orovoae ouvwon. onose er Regron. Lt.S. Nuclear Regul sa,e.a Division of Reactor Projects 1/11 Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C. 20555

9. SPON50R ING,e,ORG ANIZ ATION - N AM E AND ADDR ESS in NRC type "Same as above"; ti cantveror orcerne NRC Divisoon. OHoce ar

,s ,,,,u,6, ,as a Same as 8 above.

10. SUPPLEME NT ARY NOTES Pertains to Docket Nos. 50-352 and 50-353.

11. ABST RACT 000 woras or seus In August 1983 the staff of the Nuclear Regulatory Commission issued its Safety Evaluation Report (NUREG-0991) regarding the application of the Philadelphia Electric Company (the licensee) for the licenses to operate the Limerick Generating Station, Units 1 and 2, located on a site in Montgomery and Chester Counties, Pennsylvania.

Supplement 1 was issued in December 1983. Supplement 2 was issued in October 1984. Supplement 3 was issued in October 1984. Supplement 4 was issued in May 1985. Supplement 5 was issued in July 1985. Supplement 6 was issued in August 1985 and Supplement 7 was issued in April 1989. Supplement 7 addresses the major design differences between Units 1 and 2, the resolution of all issues that remained open when the Unit 1 full-power license was issued, the staff's assessment regarding the application by the licensee to operate Unit 2 and issues that require resolution before issuance of an operating license for Unit 2.

This Supplement 8 addresses further issues that require resolution prior to issuance of an operating license.

i3. Av At L Absui V bI A1kMt N 3

12. KE Y WQRDS/DE SCFCPIORS tider worcs or phrases ther wait antsr researcaers in focersny ene report ;

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