ML20235M884

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Final Evaluation of Beaver Valley Power Station Unit 2 Tech Specs, Informal Rept
ML20235M884
Person / Time
Site: Beaver Valley
Issue date: 06/30/1987
From: Baxter D, Branson G
EG&G IDAHO, INC., IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:
NRC
Shared Package
ML20235M874 List:
References
CON-FIN-A-6824 EGG-NTA-7616, NUDOCS 8707170365
Download: ML20235M884 (34)
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                                                 .3    EGG-NTA-7616 ic                           ,             June 1987 1

INFORMAL REPORT l

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                     . idaho.               b National:                                EVALUATION OF BEAVER VALLEY POWER STATION Engineering .                               ,    UNIT 2 TECHNICAL SPECIFICATIONS Laboratory;                                 ,

Managea by the U.S. Y, g, g, ggxter f Department: G. L. Branson of Energy . 6 

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Prepared for the 

            "'  '*"87c%",,"lll                      U.S. NUCLEAR REGULATORY COMMISSION
       ^ No. D&AC07-76/D01570 L
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l 5 I i OlSCLAIMER , f This book was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, , nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefufaess of any j information, apparatus, product or process disclosed, or represents that its use would not infrvige pnvately owned nghts. References herein to any specific commercial uroduct, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessanly constitute or irroly its endorsement, recommendation, or favonng by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or any agency thereof. p

i EGG-NTA-7616 1 l l l EVALUATION OF BEAVER VALLEY POWER STATION UNIT 2 TECHNICAL SPECIFICATIONS D. E. Baxter 

   .                                                         G. L. Branson                       ,

i l Published June 1987 Idano National Engineering Laboratory EG&G Idaho, Inc. Idaho Falls, Idaho 83415 l ) Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C. 20555 . Under DOE Contract No. DE-AC07-76ID01570 l l 

  ,                                                           FIN No. A6824                      l l

1 I I l: L___----___---__----------._.___----------------- -- - -

ABSTRACT 

                                                                                           ~

This document was prepared for the Nuclear Regulatory Commission (NRC) to assist them in determining whether the Beaver Valley Power Station . Unit 2 Technical Specifications (T/S), which govern plant systems configurations and operations, are in confornance with the assumptions of the Final Safety Analysis Report (FSAR) as amended, and the requirements of ) the Safety Evaluation Report (SER) as supplemented. A comparative audit of the FSAR as amended, and the SER as supplemented was performed with the , Beaver Valley Unit 2 T/S. Several discrepancies were identified and subsequently resolved by the NRC cognizant reviewer. Resolutions to the ] discrepancies noted in this report were achieved through conversations between the NRC Reviewer and the Utility. Resolutions were received for I all discrepancies that required resolution. , I 1 4 I 1 l 

                                                                                           . l l

L

FOREWORD This report is supplied as part of the Power Reactor Technical Specifications Evaluations being conducted for the U.S. Nuclear Regulatory Comission, Of fice of Nuclear Reactor Regulation, Division of Licensing by EG&G Idaho, Inc., NRR and I&E Support Unit. The U.S. Nuclear Regulatory Comission funded the work under the authorization S&R 20 19 40 41 1, FIN No. A6824 - Power Reactor Technica? Specification Evaluation. e e O l ____ _ ______ _-__._____-__________-____a

CONTENTS ABSTRACT .............................................................. 11 FOREWORD .............................................................. Lili .

1. INTRODUCTION ..................................................... 1
2. REVIEW CRITERIA .................................................. 1
3.

SUMMARY

..........................................................      2
4. BEAVER VALLEY POWER STATION UNIT 2 TECHNICAL SPECIFICATIONS, FSAR, SER CONSISTENCY COMPARISON ................................. 3 l

Section I. Safety Limits ~...................................... 3 ] 1 Section II. Reactor Protection System Setpoints ................ 3  ! l Section III. Engineered Safety Features Actuation System 3 Setpoints .............................. ........... 3 

                                                                                        )

1 Section IV. Pressure Boundary Isolation Valves ................. 3' H 1 l Section V. Containment Isolation Valves ....................... 4 - i i a Section VI. Containment Depressurization and Cooling System ' Limiting Conditions for Operation (LCO) ............ 5 Section VII. Combustible Gas Control Systam Limiting l Conditions for Operation ........................... 5 Section VIII. Technical Specifications Requirements Documented in the Safety Evaluation Report .................... 6 iv

EVALUATION OF BEAVER VALLEY POWER STATION UNIT NO. 2 TECHNICAL SPECIFICATIONS

1. INTRODUCTION The Beaver Valley Power Station Unit 2 is a Westinghouse Pressurized Water Reactor (PWR) plant. It has been selected for an audit to determine if the Beaver Valley Technical Specifications (T/S) are consistent with the Beaver Valley Final Safety Analysis Report (FSAR) up to and. including i Amendment 15 and:the Beaver Valley Safety Evaluation Report (SER) up to and including Supplement 3. The specific sections of the T/S which were audited are listed in Part 2. Differences between these sections of the T/S and the FSAR and SER along with the resolutions are identified in Part 4 of this report.
2. REVIEW CRITERIA l

The following T/S sections were reviewed for this evaluation.

1. Safety Limits
2. Reactor Protection System (RPG) Setpoints
3. Engineered Safety features Actuation System (ESFAS) Setpoints j
4. Pressure Boundary Isolation Valves (PIVs) l S. Containment Isolation Valves (CIVs)
6. Containment Depressurization and Cooling System Limiting Conditions for Operation (LCO)
7. Combustible Gas Control System LCOs
8. Technical Specification Requirements Contained in the Safety Evaluation Report (SER) i The sections of the T/S listed in Part 4~were compared to'the FSAR and SER to determine if the 1/S are CONSISTENT, CONSERVATIVE or DIFFERENT than the FSAR and SER. Setpoints and lists of valves and instruments in the T/S were checked against tables in the FSAR and SER.

1

The SER was reviewed to ensure that T/S requirements in the SER were addressed in the T/S. - A description of each difference between the T/S and the FSAR and SER- . is included in this report.

3.

SUMMARY

During the performance of this audit, several differences between the T/S, SER and FSAR were noted. The items'are listed below and have been assigned a status code which indicates the status of the item. The7e items are discussed in detail in Part 4 of this report. All other sections were evaluated and found to be consistent or conservative. Section Item Title Page . Status * 

                                                                                                                       ~

Section IV Pressure Boundary Isolation Valves 3 5 Section V Containment Isolation Valves 4 4 Section VIII 1 Ohio River Floods 6 3 - Section VIII 5. Power Distribution 10 3,5 Section VIII 9. l.oose Parts Monitoring System 12 3 , Section VIII 14. Engineered Safety Features Materials 15 5 l Section VIII 16. Containment Structure 16 3 Section VIII 17. Control Room Habitability Systems 17 6 Section VIII 20. Trip Setpoints and Margins 18 5 Section VIII 21. Proposed Anticipatory Trip Modification 19 6 Section VIII 22. Undetectable Failure in Online Testing 20 5 Circuitry fur Engineered. Safeguards Relays Section VIII 26. Emergency Diesel Engine Fuel Oil Storage 22 5 l and Transfer System Section VIII 28. Turbine Generator 23 3,5 Section VIII 29. Auxiliary feedwater System 24 2 Section VIII 31. Past LOCA Leakage from ESF Systems 25 5 Section VIII 32. Inservice Testing of Pumps and Valves 26 .5

  • Status Codes
1. Unresolved, awaiting NRC/ Utility action
2. Resolved pending issuance of T/S revision -
3. Resolved pending issuance of SER Supplement
4. Resolved pending issuance of FSAR Amendment ~
5. Resolved NRC accepts as-is
6. Resolved, item clarified and accepted l b l

2 E_--_-________________________________.__________ i_ _ __ _ _ -- _ _ _ _ . _ _ _ _

i

4. DEAVER VALLEY POWER STATION UNIT 2 TECHNICAL SPECIFICATION, FSAR, SER CONSISTENCY COMPARISON
 ,                                           Section I. Safety Limits                           ;

1 This section covers the review of the safety limits as defined in Section 2.1 of the Technical Specifications. It includes reactor core limits and RCS pressure. FSAR SER Technical Specification Section Section Evaluation l 2.1.1 Reettor Core Limits 4.4.1.1 & 4.4.1 CONSISTENT 3.2.5 DNB Parameters 15.0 2.1.2 Reactor Coolant 5.2.2 5.2.2 CONSISTENT l System Pressure Section II. Reactor Protection System Setpoints This section covers the review of the Reactor Protection System Setpoints to ensure the T/S values agree with or are conservative to the values assumed in the safety analysis or defined in the SER. FSAR SER Technical Specification Section Section Evaluation 2.2 Reactor Trip System 7.2.2 7.2 CONSISTENT Instrumentation Setpoints 15.0 Section III. Engineered Safety Features Actuation System (ESFAS) Setpoints This section covers the review of the ESFAS setpoints to ensure the T/S values agree with or are conservative to the values identified in the l FSAR sections or as defined in the SER as required values. l FSAR SER Technical Specification Section Section Evaluation 3/4.3.2, Table 3.3-4 7.3, 7.3 CONSISTENT 15.0 & 15.1 Section IV. Pressure Boundary Isolation Valves (PIVs) This review determines if all the PIVs identified in the FSAR and SER are included in the T/S. i l 3 i

I FSAR SER Technical Specification Section Section Evaluation 3.4.6.3 Table 4.4-3 3.98.6.2 3.9.6 NOT EVALUATED - RESOLUTION The Staff has reviewed and accepts the PIV list of Table 4.4-3 This item is CONSISTENT There was no PIV list in the SER or the FSAR. Also, SSER #3 Section 3.9.6 states " Table 4.4-3 of the staff-approved Beaver Valley Unit 2 Technical Specifications will constitute the acceptable list of valves designated as Pressure Isolation Valves." Section V. Containment Isolation Valves (CIVs)  ! This review determines if all the CIVs identified in the FSAR and SER are included in the T/S. j FSAR SER Section Section Evaluation Technical Specification 6.2.4 3/4.6.3 pg. 3/4 6-15 Table 6.2-60 6.2.4 DIFFERENT l I Table 3.6-1 l l 1. Several maximum stroke times do not agree between T/S Table 3.6-1 & l FSAR Table 6.2.4. Penetration Valve ID Table 3.6-1 Table 6.2-60 7 2 SIS-MOV869A T/S N/A FSAR 10 sec 17 2 SIS-MOV869B T/S N/A FSAR 10 sec l 19 2CHS-MOV378 T/S <60 FSAR 15 sec j 19 2CHS-MOV381 T/S <60 FSAR 10 i 34 2 SIS-MOV836 T/S N/A FSAR 10 1 2 SIS-MOV840 T/S N/A FSAR 15 35 2CHS-MOV308A T/S N/A FSAR 10 36 2CHS-MOV308B T/S N/A FSAR 10 , 37 2CHS-MOV308C T/S N/A FSAR 10

Penetration Table 3.6-1 Table 6.2-60'

  • 60 2 SIS-MOV8888B T/S N/A FSAR 15
 ~

j 61 2 SIS MOV8889 T/S N/A FSAR 15 62 2 SIS-MOV8888A T/S N/A FSAR 15 106 2 SIS-MOV842 T/S <60 FSAR 10 ' l 2. T/S Table 3.6-1 penetration 77 lists 2FWS-HYV1578. FSAR Table 6.2-60 penetration 77 lists *2FWS-HVY157B. RESOLUTION With the issuance of FSAR Amendment 17 all of tne closure times will be in agreement with the Technical Specifications. With the issuance of FSAR Amendment 18.the typographical error for penetration 77.will be corrected. This item is CONSISTENT Section VI. Containment Depressurization and Coolina System (CDCS) Limitina Conditions for Operation (LCO) This section reviews the LCOs for the CDCS to ensure they adequately cover the operation of the CDCS during all reqc-ired modes of plant operation. FSAR SER Technical Specification Section Section Evaluation 3/4 6.2 pg. 3/4 6-10 6.2.2 6.2.2 CONSISTENT LCO 3.6.2.1 S/R 4.6.2.1 LC0 3.6.2.2 S/R 4.6.2.2 LCO 3.6.2.3 S/R 4.6.2.3 The LCOs and S/Rs for this system are applicable during Modes 1, 2, 3, and 4 and require all systems be operational. Section VII. Combustible Gas Control System (CGCS1 Limitino Conditions for Operation (LCOs) This section reviews the LCOs for the CGCS to ensure they adequately cover the operation of the CGCS during all required modes of plant operation. 5

4 FSAR SER Technical Specification Section .Section Evaluation 3/4.6.4 Pg. 3/4 6-31 6.2.5 6.2.5 CONSISTENT - LC0 3.6.4.1 S/R 4.6.'4.1 LC0 3.6.4.2 - S/R 4.6.4.2 i The LCOs and S/Rs for this system are applicable during Modes 1 and 2 and require all systems be operational. 

      -Section VIII. Technical Specification Reautrements Documented in the Safety Evaluation Report This section covers the review of all the items identified in the             -

Safety Evaluation Report (SER) and Supplements to the Safety ~ Evaluation Report (SSER) as T/S required items and whether they have or have not been adequately addressed in the T/S.

1. SER Section: .2.4.3.1 Ohio River Floods- Pg. 2-17 states:

A Technical Specification will require that a plant flood alert be issued for an Ohio River water level of 690 feet ms1. The plant will ) be shutdown immediately when the river water level reaches an j elevation of 695 feet ms1 and the water level is' rising upstream. - 

                                                                                    ]

T/S Section: 3/4.7.6 Pg. 3/4 7-14 , No T/S was found to address a plant flood alert for an Ohio River water level of 690 feet ms) or immediate shutdown at 695 feet ms1. This item is DIFFERENT I RESOLUTION SSER 3 Section 2.4.3.1 deletes these requirements as follows: By letter dated August 12, 1986, the applicant submitted a letter to state that the current Beaver Valley Unit 1 Technical Specifications on flood protection do not require a " flood alert" be issued; however, a Beaver Valley Site Emergency Preparedness Plan (EPP) procedure does contain such a requirement. Therefore, on the basis of a comparable, i requirement already in existence, and on the basis that it is not the staff's practice to impose additional " administrative type" specifications on a second unit at the same site, the staff deletes the proposed " flood alert" requirement in the Technical Specifications. n 6

In the same letter, the applicant pointed out that the Beaver Valley Unit i Technical Specification on flood protection does not require immediate shutdown when the water level reached 

    .                  695 feet msi; it does, however, require that the reactor "be in at least HOT STANDBY within 6 hours and in COLD SHUTDOWN within the following 30 hours." The basis of that specification states that "the limit of 695 msl was selected on an arbitrary basis as an appropriate flood level at which to terminate further operation and initiate flood protection measures for safety-related equipment." Therefore, the Unit 2 specification will be written to be the same as that of Unit 1 in regard to reactor shutdown due to an Ohio River flood.      This i

corrects the statement in the SER. This item is CONSISTENT

2. SER Section: 2.4.11.2 Emergency Water Supply Pg. 2-24 states:

As a result of the staff's request, a Technical Specification will limit the operation of the station to periods when the water level in the river is at or above elevation 654 feet msl. The service water pumps are designed to supply water at river levels as low as 648.6 feet msl. Therefore, the staff concludes that a river level limit of 654 feet ms) is adequate to ensure that the plant can be safely shutdown during low flow conditions in the Ohio River. Because the UHS will handle heat loads at a maximum river water temperature of 86*f, the Technical Specifications also limit station operation to periods when the river temperature is less than or equal to 86*f. Average river water temperature will be determined once every 24 hours. T/S Section: 3/4.7.5 Pg. 3/4 7-13 T/S 3.7.5.1 specifies a minimum water level and a maximum average water temperature consistent with the above requirements and requires that it be so determined once per 24 hours. This item is CONSISTENT

3. SER Section: 3.4.1 flood Protection Pg. 3-6 states:

A Technical Specification will require flood protection measures be taken for all safety-related systems, components, and structures when the water level of the Ohio River at the intake structure exceeds 695 feet msl, which is 10 feet below the PMF level. This Technical Specification, which is the same as that for Unit 1, is applicable at all times and requires that the following protective actions be taken, as a minimum: achieve hot standby within 6 hours, achieve cold shutdown within the following 30 hours, and close and seal the six cubicle flood doors in the four intake structure cubicles. 7

T/S Section: 3/4.7.6 Pg. 3/4 7-14  ; T/S 3.7.6.1 requires the above actions when the Ohio River water level exceeds 695 msl. . This item is CONSISTENT

4. SER Section: 3.9.6 Inservice Testing of Pumps and Valves Pg. 3-39 states:

There are several safety systems connected to the reactor coolant r pressure boundary that have design pressure below the rated reactor coolant system (RCS) pressure. There are also some systems that are rated at full reactor pressure on the discharge side of pumps but have suction below RCS pressure. To protect these systems from RCS pressure, two or more isolation valves are placed in series to form the interface between the high-pressure RCS and the low-pressure system. The leaktight integrity of these valves must be ensured by periodic leak testing to prevent exceeding the design pressure of the low-pressure systems. Limiting conditions for operation must be added to the Technical Specifications that will require corrective action (shutdown or system isolation) when the final approved leakage limits are not met. Also, the Technical Specifications must provide surveillance requirements that will state the acceptable leak rate testing frequency. Periodic leak testing of each pressure isolation valve must be performed at least once each refueling outage, after valve - maintenance, before return to service, and for systems rated at less than 50% of RCS design pressure each time the valve has moved from its fully closed position unless justification is given. The testing interval should average about 1 year. Leak testing should also be performed after all disturbances to the valves are complete, before power operation following a refueling outage, maintenance, and so forth. The staff's position on leak rate limiting conditions for operation is that leak rates must be equal to or less than 1 gpm for each valve to ensure the integrity of the valve, demonstrate the adequacy of the redundant pressure isolation function, and give an indication of valve degradation over a finite period of time. Significant increases over this limiting value would indicate valve degradation from one test to another. The Class 1 to Class 2 boundary will be considered the isolation point that must be protected by redundant isolation valves. In cases where pressure isolation is provided by two valves, both will be independently leak tested. When three or more valves provide ' isolation, only two of the valves must be leak tested. 8

In a letter dated November 7, 1984, the applicant submitted a list of pressure isolation valves to be included in the leak rate testing program along with four sets of piping end instrument diagrams. On 

                  ,           the basis of its review of that submittal, the staff determined that-the applicant's response was incomplete with respect to the above staff position on the pressure isolatici valves leak testing
                 .            requirement. The specific concerns have been transmitted to the applicant. The staff will report its final evaluation in a supplement to this SER.

SSER 1 Section: 3.9.6 Inservice Inspection of Pumps and Valves Pg. 3-2 states: In the SER, the staff indicated that the allowable leak rate limit for pressure isolation valves (PIVs) was to be no more than 1 gpm for each valve. Since that time, the NRC has adopted a revised and more realistic leak rate criterion for PIVs. The new acceptable leak rate is 1/2 gpm for each nominal inch of valve size up to a maximum of 5 gpm. In addition, the requirements of paragraph IWV-3427(b) of Section XI of the ASME Code are to be applied in order to determine if the leak rates are acceptable. The applicant may submit revised proposed technical specifications to comply with this new criterion. SSER 3 Section: 3.9.6 Inservice Inspection of Pumps and Valves Pg. 3-2 states: 

                ~

In the SER, the staff stated that the applicant's proposed list of pressure isolation valves (PIVs) to be listed in the Technical Specification and associated leak testing requirements were unacceptable. In SSER 1, the staff provided revised limits for allowable leak rates. By letter dated July 21, 1986, the applicant committed to submit revised proposed Technical Specifications to adopt these limits. T/S Section: 3/4.4.6 Pg. 3/4 4-21 T/S 3.4.6.3 specifies maximum allowable leakage limits in accordance with Table 4.4-3 which is in agreement with the revised limits of SSER 1. This item is CONSISTENT S/R 4.4.6.3.1 and 4.4.6.3.2 specify the leakage testing frequency of the PIVs. This item is CONSISTENT 4 1 9

l I

5. SER Section: 4.3.2.1 Power Distribution Pg. 4-15 states:  !

The analysis. performed by W indicated that the peaking factor limit could not be met at the beginning of life (BOL) of cycle 1 because of - 

                                                                                                         ' i the wide Al band. This resulted in limiting the width of the band-                                  j for the first.20% of the cycle typically, and until 3000 MWD /MTU burn
  • up to the value of 5% AI. This 15% AI is the value. ll previously justified by the CAOC analysis. These features will be  !

incorporated in the Technical Specifications. T/S Section: 3/4.2 Pg. 3/4 2-1 H T/S 3.2.1 does not appear to have incorporated the 5% AI limit , but rather a 7% AI. This item is DIFFERENT i RESOLUTION The Staff is presently reviewing the applicants newfanalysis'. With' the issuance of.SSER #5 the. Staff will determine which value is I acceptable. Any required changes to the Technical Specifications will be made at that time. This item is accepted as-is. This item is CONSISTENT

6. SER Section: 4.3.2.4 Control Rod Patterns and Reactivity Worths Pg. 4-17 states: ,

i The control rods are divided into two categories--shutdown rods and I regulating rods. The shutdown rods are always completely out of the core when the reactor is at operating conditions. Core power changes are made with regulating rods that are nearly out of the core when it is operating at full power. Regulating rod insertion will be controlled by power-dependent insertion limits required in the Technical Specifications to ensure that (1) there is sufficient negative reactivity available to permit rapid shutdown of.the reactor with adequate margin (2) the worth of a control rod that might be ejected is not greater , than that which has been shown to have acceptable consequences in 1 the safety analyses i T/S Section: 3/4.1.3 Pg. 3/4 1-23 and 3/4 1-24 i T/S 3.1.3.5 requires all shutdown rods be fully withdrawn. T/S 3.1.3.6 regulates control banks by power-dependent insertion limits. This item is CONSISTENT s e i 10 1

7. SER Section: 4.4.3.1 Fuel Rod Bowing Pg. 4-22 states:

A significant parameter that affects the thermal-hydraulic design of 

  ,                                  the core is rod-to-rod bowing within fuel assemblies.       The W method for predicting the effects of rod bowing on DNB are in WCAP-8691, Revision 1, " Fuel Rod Bow Evaluation,"~which has been approved by the staff.

For plants designed by W, the staff has approved the generic margins given in Table 4.1, which may be used to offset the reduction in DNBR as a result of rod bowing. Plant-specific margins that could be available are: (1) The Technical Specification minimum flow rate is greater than the design flow rate. (2) The Technical Specification maximum T ave is less than the design Tave-(3) The trip setpoints are more limiting than the thermal-hydraulic analysis indicates. In a letter dated July 12, 1984, responding to the staff's concerns the applicant stated that a 9.1% margin is maintained at Beaver Valley Unit 2 to accommodate full- and low-flow DNBR penalties. This is consistent with WCAP-8691, which has been approved by the staff, and 

 .                                   thus is acceptable. However, the applicant should insert into the basis of the Technical Specification any of the generic or plant-specific margins that may be used to offset the reduction in DNBR as a result of rod bowing.

T/S Section: 8 3/4.2.2 and B 3/4.2.3 Pg. B 3/4 2-4 T/S Bases 3/4.2.2 and 3/4.2.3 include the required DNBR penalties and generic margins used to offset the rod bow penalties. This item is CONSISTENT

8. SER Section: 4.4.3.2 Crud Deposition Pg. 4-23 states:

Crud deposits in the core and an associated change in core pressure drop and flow have been observed in some PWRs not of W design. The staff requested that the applicant describe the procedures to detect flow degradation as a result of crud buildup. The applicant responded that, except for steam generatos tube plugging, there have been no reports of significant flow reduction in a relatively short period of time at any W plant. The staff will ensure that the Technical Specifications contain the requirement that the actual reactor coolant system flow rate be verified to be greater than or equal to the minimum design flow rate plus uncertainties at least once every 12 hours. In addition, the i staff will ensure that the applicant performs a channel calibration at least once every 18 months. 11

T/S Section: 3/4.2 Pg. 3/4 2-13 T/S 3.2.5 and S/R 4.2.5.1.1 and 4.2.5.2 specify the required flow rate monitoring and channel calibration.

  • This item is CONSISTENT ,
9. SER Section: 4.4.4 Loose Parts Monitoring System Pg. 4-23 states: i The applicant has provided a description of the loose parts monitoring system (LPMS) that will be used at Beaver Valley Unit 2. The design will consist of 10 active instrumentation channels, each comprising a piezoelectric accelerometer (sensor) and signal conditioning equipment. Sensors are fastened mechanically to the reactor coolant system (RCS) at each of the following potential loose parts collection regions:

(1) reactor pressure vessel, upper head region  ; (2) reactor pressure vessel, lower head region l (3) each steam generator, reactor coolant inlet region The system will be capable of detecting a metallic loose part that weighs from 0.25 to 0.30 pound impacting within 3 feet of a sensor and having a kinetic energy of 0.5 foot-pound on the inside surface of the RCS pressure boundary. The staff was concerned about compliance with Paragraph C.4.K of RG 1.133, " Loose-Part Detection Program for the Primary System of , Light Water Cooled Reactors," which states that the portion of the system within containment will be designed and installed to function following all seismic events up to and including the operating basis earthquake (08E). The applicant, in FSAR Table 1.8, took exception to this as requiring the system to be seismically qualified. After discussions with the staff, the applicant, in a letter dated March 4, 1985, stated that despite this, the system was qualified to loads greater than those expected at the Beaver Valley Unit 2 site for an 08E. The staff, therefore, considers this issue acceptably resolved. In a letter dated October 12, 1984, the applicant responded acceptably to the staff's concerns. In Attachment 2 to the letter dated August 7, 1985, the applicant has committed to provide the alert level for startup and power operation to the NRC staff by September 30, 1987 following completion of the startup test program. Thus, this issue is now closed. However, the Technical Specifications should have a section on the LPMS addressing operability and surveillance requirements similar to the Westinghouse Standard Technical Specifications. T/S Section: Pg. None of these requirements were found in the T/S. 12

_m This item is DIFFEREN,T RESOLUTION The T/S requirements for the LPMS will be; deleted with the issuance of SSER #5. This item is CONSISTENT

10. SER Section: 4.4.6 N-1 Loop Operation Pg. 4-24 states:

N-1 loop operation refers to operation of the' reactor with one of the reactor's coolant loops out of service. Thus, in the case of Beaver Valley Unit 2 only two coolant loops are available to supply coolant to the reactor core. To exercise the option.to operate'in'the N-1 mode, the applicant must provide core thermal-hydraulic analyses taking into account the effect of partial loop operation on core inlet flow distribution, minimum DNBR (MDNBR), and the effect of N-1 loop operation on postulated transients and accidents. If the applicant chooses to not use the N-1 loop operation, the staff will require that the Technical Specifications include appropriate provisions to ensure that this type of operation is prohibited. 

   .                                                                     T/S Section: 3/4.4.1                        Pg. 3/4 4-1 T/S 3.4.1.1 requires that all reactor coolant loops be in cperation
   .                                                                     for modes 1 and 2.

i This item is CONSISTENT 1 1 11. SER Section: 5.2.2.2 Overpressure Protection During Low-Temperature l Operation Pg. 5-4 states: l l Low-temperature overpressure protection is primarily provided by two 

                                                                         .. the three pressurizer PORVs. These two have automatically edjusted opening setpoints, adjusted as a function of reactor coolant temperature. The reactor coolant temperature measurements will be auctioneered to obtain the lowest value. This temperature will be translated into a PORV setpoint curve that will adequately account for the lag in the temperature change of the reactor vessel and for possible single failures in the auctioneering system, so'the system pressure will always be below the maximum allowable pressure. This          l PORV setpoint curve and the requirement for its periodic updating shall be in the Technical Specifications to ensure that the stress          l intensity factors for the reactor vessel at any time in the life of        I the plant are lower than the reference stress intensity factars            I specified in 10 CfR 50, Appendix G.
   ~

1 The system logic will first annunciate at a predetermined low RCS temperature to alert the operator to arm the system. The staff will require a Technical Specification on this temperature. 13

i l-j T/S Section: 3/4.4.9 and B 3/4.4.9 Pg. 3/4 4-35 and B 3/4 4-14 j T/S 3.4.9.3 and Figure 3.4-4 along with B 3/4.4.9 incorporate the above T/S requirements. . l This' item is CONSISTENT ' j i j 12. SER Section: 5.2.2.2 Overpressure Protection During Low-Temperature j 1 Operation Pg. 5-4 states: ) l l With'a single failure of one of the two PORVs and no credit for the RHR system relief valves, the low temperature overpressure protection system can relieve the capacity of only one high head safety injection i (HHSI)/ charging pump and maintain pressure.below the Appendix G limits. Thus, operating procedures will require the removal of power from all HHSI/ charging pumps that are not required to be operable. To j prevent an accidental overpressurization by an accumulator discharge, j operating procedures will stipulate that (1) the accumulator isolation -l valves shall be closed when the RCS pressure is below the safety j injection (SI) unblock setpoint and (2) after they are closed, their j operating power shall be removed. To prevent overpressurization as a J result of an excessive temperature differential between the RCS and an i isolated steam generator, there will also be restrictions on the conditions under which a reactor coolant pump may be started. The staff will require Technical Specifications on these three items. T/S Section: 3/4.5 and 3/4.4.1 Pg. 3/4 5-2 and 3/4 4-7 S/R 4.5.1.3 requires the accumulator discharge isolation valves be closed and deenergized when Reactor Coolant System Pressure is below  ; 1000 1 100 psig. ) i T/S 3.4.1.6 places restrictions on reactor coolant pump startup with  ! an isolated steam generator and an inoperable PORV. j i These items are CONSISTENT  !

13. SER Section: 5.2.5 Reactor Coolant Pressure Boundary Leakage Detection Pg. 5-11 states: '

l The applicant has stated that the plant Technical Specifications would provide limiting conditions for identified and unidentified leakage, thus satisfying RG. 1.45, Position C.9. T/S Section: 3/4.4.6 Pg. 3/4 4-19 T/S 3.4.6.2 provides limiting conditions for identified and unidentified leakage. l This item is CONSISTENT l l 14 l 1

1 ! 14. SER Section: 6.1.1 Engineered Safety Features Materials Pg. 6-2 i states l l , The staff evaluated the pH of the water (mixture of refueling water l l storage tank and sodium hydroxide solution) in the containment sump and verified, by independent calculations, that sufficient sodium , l hydroxide is available to raise the containment sump water pH above ' the minimum 7.0 level to reduce the probability of stress-corrosion j cracking of austenitic stainless steel components. The removal effectiveness of the chemical additive for fission products in containment is reviewed in Section 6.5.2. The staff will review the i surveillance requirements in the plant Technical Specifications to d verify that sufficient sodium hydroxide is maintained in the containment spray additive tank. T/S Section: 3/4.6.2 Pg. 3/4 6-14 l T/S 3.6.2.3 specifies a 23-25% by weight HaOH solution. 1 This item requires staff review and approval. This item is NOT EVALUATED , i RESOLUTION ] 

  .                                   The Staff has reviewed and approves this concentration level.

This item is CONSISTENT

15. SER Section: 6.2.1.1 Containment Structure Pg. 6-4 states:

With respect to the peak containment pressure analysis, the LOCAs analyzed by the applicant (RCS pipe breaks) include a spectrum of hot leg and cold leg (pump suction and pump discharge) breaks, up to and including the double-ended rupture of the largest reactor coolant line. The spectrum of secondary system pipe breaks analyzed by the applicant includes double-ended and split breaks of the main steamline at different reactor power levels (102%, 70%, and 30% of full power, and the hot shutdown condition). A single failure analysis is not necessary for the peak containment pressure evaluation because the peak pressure for each case analyzed occurs before active ESF systems can influence the results. The DBA for peak containment pressure (containment integrity DBA) was determined to be the double-ended guillotine break in the hot leg (hot leg double-ended rupture, HLDER). The peak containment pressure calculated by the applicant (using the Stone and Webster LOCTIC computer code) was 44.7 psig, which is below the containment design pressure of 45 psig. The applicant also performed a sensitivity study and found that the initial conditions that result in the highest peak calculated pressure are the maximum initial containment pressure (11.6' psia), maximum initial containment temperature (10$*F), and maximum initial-containment dew point (105"F) (relative humidity). These are the limiting values that will be allowed by the Technical Specifications. 15 E__ - - _ - - - - - - - - - _ - - - - - - - - - - - - - - - _ - - _

r ] l SSER 3 Section: 6.2.1.1 Containment Structure' Pg. 6-1 states: In the SER, the staff stated that "The applicant also performed a' i sensitivity study and found that initial conditions that result in-the- . i highest peak calculated pressure are the maximum initial containment pressure (11.6 psia), maximum initial containment temperature (105'F), and maximum containment dew point (105'F) (relative humidity).- These ~ are the limiting values that will be allowed by the Technical Specifications." j i The staff does not intend to impose a limiting value on containment j atmospheric humidity, nor does the staff intend to require by 3 Technical Specifications that humidity be monitored. Since the l i highest peak calculated post-LOCA pressure would be attained from the L above initial conditions, among them 100% relative humidity, any relative humidity less than 100% would not result in the highest peak ' calculated post-LOCA pressure. It would not make sense to impose a limiting value of 100% for relative humidity;.the staff will, however, impose limits on containment pressure and temperature. This clarifies- j the subject statement.  ; i T/S Section: 3/4.6.1 Pg. 3/4 6-6 & 3/4 6-8 T/S 3.6.1.4 limits containment internal air partial pressure to the-acceptable range shown on Figure 3.6-1. T/S 3.6.1.5 restricts containment average air temperature to between 85'F and 105*f. , These items are CONSISTENT 1

16. SER Section: 6.2.1.1 Containment Structure Pg. 6-5 states:

i With respect to the containment depressurization analysis, only pump  ! suction ruptures were determined to be of concern because they produce the highest energy flow rates during the post-blowdown period. The DBA for maximum depressurization time and subatmospheric peak pressure I (containment depressurization DBA) was found to be the double-ended I rupture of the pump suction line (PSDER), with minimum ESF (loss of offsite power and emergency diesel generator failure resulting in the  ; loss of one ESF train--i.e., one charging pump, one safety injection l pump, one quench spray pump, and two containment recirculation pumps l with associated coolers). The applicant also performed a sensitivity l' study and found that the initial conditions that result in the maximum depressurization time are initial containment pressure of 9.05 psia, l initial contair. ment temperature of 85'F, initial containment dew point  ; of 85'f, service water temperature of 86*F, and RWST temperature of 50*F. .These are the limiting values that will be allowed by the Tet nnical Specifications. The applicant calculated a maximum ,  ; containment depressurization time of 3480 seconds, which is within the design limit of 3600 seconds, and a subatmospneric peak pressure of 

                                   -0,03 psig. A barometric pressure of 14.36 psia was used in the                                     . i l

l 16 

                                                                                                                'l 1

analysis. This value is based on climatological data for Pittsburgh '] (U.S. Dept -of Commerce, Weather Bureau, 1963-64 local climatological  ; data, Pittsburgh, Pennsylvania,-Greater Pittsburgh Airport)' adjusted l to plant grade. The staff also performed a confirmatory conta'.nment depressurization analysis based on' initial conditions and analytic parameters given by the applicant and finds the applicant's calculated 

 -      maximum containment depressurization time is conservative.                                                 ;

T/S Section: 3/4.6.1 Pg. 3/4 6-5 & 3/4 6-8 T/S 3.6.1.4 and 3.6.1.5 limit the containment pressure, containment , temperature, service water temperature, and RWST temperature to the i values required by the'SER. j i There was no T/S identified on the containment dew point as required ) by the SER.- l This item is DIFFERENT 1 RESOLUTION SSER #5 will delete the requirement for the containment dew point as  ; specified. ) i This item is CONSISTENT I .

17. SER Section: 6.4 Control Room Habitability Systems Pg. 6-28 states: j l
 .      The Beaver Valley Unit 2 FSAR states that chlorine will be stored in eight 1-ton containers located 500 feet from the nearest control room intake. To ensure compliance with NRC guidelines on the protection of_                                      ,

the control room operator following a chlorine release, the staff will j review the Beaver Valley Unit 2 Technical Specifications to ensure that control room isolation and pressurization response-times and the pressurization test flow rates are consistent with Table 1 of RG 1.95. l T/S Section: 3/4.7.7 Pg. 3/4 7-15 This item requires staff review and approval. l 1 This item is NOT EVALUATED RESOLUTION ) The Staff has reviewed the system and has drafted a new Technical- I Specifications which is acceptable.- 1 This item is CONSISIENT l r I 1 i 17 1 

                                                                    --__2_. . _ _ _ _ _ _ _ _ - _ _ _ _ _ - - -

l i

18. SER Section: 7.2.2.1 Lead, Lag, and Rate Time Constant Setpoints Used in Safety System Channels Pg. 7-7 states:

Several safety. system channels make.use of lead, lag, or rate signal

  • compensation to provide signal time responses consistent with ~i assumptions in the analyses in FSAR Chapter 15. The time constants ,

for these signal compensations are adjustable setpoints within the analog portion of the safety system. The time. constant setpoints will be incorporated into the plant Technical Specifications. T/S Section: 2.2 Pg. 2-7 ) Table 2.2-1 Notes 1, 2 and 3 specify the time constant setpoints as required. This item is CONSISTENT

19. SER Section: 7.2.2.2 Turbine Trip Following a Reactor Trip Pg. 7-7 states.

Credit is taken in the accident analysis for turbine trip on a reactor l trip. The protection system trips the turbine following a reactor I trip using the turbine emergency trip system. Redundant circuits used I to trip the turbine are independently routed to and processeo within the emergency trip system to provide two independent means.of tripping . the turbine. The circuits.that traverse structures that are not i seismically qualified are isolated from the solid-state protection system. The circuits are fully testable during full-power operation. , l The staff finds this design consistent with the function's importance to safety and, therefore, acceptable. The staff will include in the plant Technical Specifications a requirement to periodically test these circuits. j T/S Section: 3/4.3.2 Pg. 3/4 3-15 , l S/R 4.3.2.1.1 requires a channel functional test be performed on the reactor trip circuits which in turn trip the turbine at the frequency specified in Table 4.3-2. This item is CONSISTENT

20. SER Section: 7.2.2.4 Trip Setpoint and Margins Pg. 7-8 states:

l The setpoints for the various functions in the reactor trip system are-determined on the basis of the accident analysis requirements. As such, during any anticipated operational occurrence or accident, the reactor trip maintains system phran:eters with the following limits: a (1) minimum DNBR of 1.30 (2) maximum system pressure of 2750 psi (absciute) (3) fuel rod maximum linear power of 18.0 kW per foot 18 i I j

' 1 The staff requested detailed information on the methodology used to  ! establish the Technica1' Specification trip setpoints and allowable values for the reactor protection system (RPS)-(including reactor trip  ! and engineered safety feature channels) assumed to operate in the FSAR  ! accident and transient analyses. This includes the following information:  ; (1)- The trip setpoint and allowable value for.the Technical '! Specifications. (2) The safety limits necessary to protect the integrity of the  ; physical barriers that guard against uncontrolled release of l radioactivity. (3) The values assigned to each component of the combined channel error allowance (modeling uncertainties,. analytical l uncertainties, transient overshoot, response time, trip unit

setting accuracy, test equipment accuracy, primary element l accuracy, sensor drift, nominal and harsh environmental allowances, trip unit drift), the basis for these values .and the method used to sum the individual errors. Where zero is assumed for an error, a justification that the error is negligible should be provided. ,

(4) The margin (the difference between the safety limit and the i setpoint less the combined channel error allowance). l The detailed trip setpoint review will be done as part of the staff's review of the plant Technical Specifications and will be completed before the operating license is issued. 

  .      T/S Section:                                Pg.

This item requires staff review and approval. This item is NOT EVALUATED RESOLUTION j The Staff has reviewed and accepts the trip setpoints in question. This item is CONSISTENT j

21. SER Section: 7.2.2.5 NUREG-0737 Item II.K.3.10, Proposed Anticipatory _l Trip Modification Pg. 7-9 states: )

The design includes an anticipatory reactor trip upon turbine trip. Provisions are included to automatically block the reactor trip upon  ; turbine trip at power levels below approximately 70% (P-9 interlock) I where the condenser steam dump is capable of mitigatir.g the reactor coolant system temperature and pressure transient without actuating , pressurizer power-operated relief valves. A decision to trip the I l 19 - l  ! l

reactor following turbine trip at the 50% power level, noted in the TMI Action Plan requirements, would involve only bistable setpoint changes and not instrument hardware changes. -The staff _ finds that the design is, therefore, acceptable. The specific power level setpoint . below which a reactor trip following a turbine trip is blocked will be reviewed and specified in the plant Technical Specifications. . T/S Section: 2.2.1 Table 2.2-1 Pg. 2-4 This item requires staff review and approval. This item is NOT EVALUATED RESOLUTION The Staff is presently reviewing the applicants analysis. Until such time that they_ approve the analysis, the Technical Specification trip point will remain at 50%. l This item is CONSISTENT

22. SER Section: 7.3.3.3 Undetectable Failure in Online Testing Circuitry for Engineered Safeguards Relays Pg. 7-19 states:

On August 26, 1982 (letter from Rahe), Westinghouse notified the staff of a potential undetectable failure in online test circuitry for the . master relays in the engineered safeguards systems. The undetectable failure involves the output (slave) relay continuity proving lamps and their associated shunts provided by test pushbuttons. If, after

  • testing, a shunt is not provided for any proving lamp because of a switch contact failure, any subsequent safeguards actuation could cause the lamp to burn open before its associated slave relay is energized. This would then prevent actuation on any associated safeguards devices on that slave relay. Westinghouse has provided test procedures that ensure that the slave relay circuits operate normally when testing of the master relays is completed.

l Until an acceptable circuit modification is installed, the staff will I require that the Technical Specifications include monthly (rather than quarterly) testing of slave relays. These tests should be performed immediately following the monthly testing of associated master relays. T/S Section: 3/4.3.2 Pg. 3/4 3-33 T/S Table 4.3-2 does not address testing of the slave relays on any scheduled frequency. This item is DIFFERENT 9 2C l

l 1 l 1 RESOLUTION j i The Staff has reviewed this problem and has agreed to allowing the applicant to incorporate this relay testing in the FSAR Chapter 16 j testing schedule. 1 

 -     This item is CONSISTENT
23. SER Section: 8.2.3.1 Capability to Test Transfer of Power Between Normal and Preferred Of f site Circuits Pg. 8-4 states:

In Amendments 3 and 9 to the FSAR and letter dated September 20, 1984, the applicant described the transfer circuitry, how it is tested , during normal plant operation, and its compliance with GDC 18. On the I basis of the description, the staff concludes that the design is testable, meets GDC 18, and is acceptable. In response to a staff concern that periodic testing of the transfer may create transients in the plant if done during power operation, the applicant indicated that testing would be performed during refueling. Testing during refueling or when the plant is shutdown resolves the staff concern and is, therefore, acceptable. Testing at 18-month , intervals when the plant is shutdown will be included in the plant l Technical Specifications. T/S Section: 3/4.8.1 Pg. 3/4 8-2 S/R 4.8.1.1.1.b requires the above testing at an 18-month interval, 

 . when the plant is shutdown.

This item is CONSISTENT

24. SER Section: 8.3.1.2 Bypass of Diesel Generator Protective Trips Pg. 8-6 states:

FSAR Section 8.3.1.1.15 indicates that a number of tripping devices have been provided for each diesel generator. The majority of these tripping devices are bypassed when the diesel generator receives an emergency start signal. Tripping devices that are not bypassed include generator current differential, generator overexcitation, and engine overspeed protection. This design meets Position 7 of RG 1.9 except for the generator overexcitation tripping device that is not bypassed. In Amendment 3 to the FSAR, the applicant indicated that the design for the generator overexcitation trip has two independent measurements with coincident logic for trip actuation. This design also meets Position 7 of RG 1.9 and therefore is acceptable. Surveillance requirements for the protective trips that are bypassed will be included in the Technical Specifications. 21 l _-________-____a

T/S Section: 3/4.8.1 Pg. 3/4 8-3 S/R 4.8.1.1.2b.4 provides the required surveillance. This item is CONSISTENT 9.2.1.1 

                                                                               ~
25. SER Section: Service Water System Pg. 9-13 states:

The SWS will be periodically tested and inspected in accordance with the Technical Specifications. Preoperational tests will also be performed. The major portions of the SWS, including the SWS pumps, are in continual use and therefore do not require periodic testing. However, the motor-operated valves in the lines to and from the recirculation spray coolers, which are not normally operated, will be , tested periodically to ensure satisfactory operation. A program for 1 detecting potential biological fouling problems also has been established, as discussed above. Therefore, the staff concludes that GDC 45 and 46 are satisfied. , T/S Section: 3/4.7.4 Pg. 3/4 7-12 S/R 4.7.4.1 specifies the required periodic testing. This item is CONSISTENT

26. SER Section: 9.5.4.2 Emergency Diesel Engine fuel Oil Storage and .

Transfer System Pg. 9-71 states: 1 The fuel oil quality and tests will conform with RG 1.137,  ; Positions C.2.a through C.2.f, and the requirements will be included I in the plant Technical Specifications. l T/S Section: 3/4.8.1 Pg. 3/4 8-1 S/R 4.8.1.1.2a.3 requires the diesel fuel meet the limits specified in  ! Table 1 of ASMT D975-68. The SER requires the fuel oil quality and tests conform with RG 1.137, Positions C.2.a through C.2.f. This item is DIFFERENT RESOLUTION The Staff has reviewed the fuel oil cuality requirements and accepts them. This item is CONSISTENT i 1 i l

27. SER Section: 9.5.7 Emergency Diesel Engine Lubricating Oil System Pg. 9-81 states:

In Amendment 7 of the FS*9, the applicant stated that assuming a high lube oil consumption rate., 504 gallons of lube oil would be required for each diesel for 7 days of operation. The applicant stated that 

    .                                                     1400 gallons of lube oil would be stored on site for the diesel generators.      The staff finds this acceptable; however, the staff is currently incorporating surveillance requirements to ensure 7 days' supply of lube oil on site at all times into the Standard Technical Specifications.       In the interim the staff will require that the following be included as plant-specific Technical Specifications:

(1) In Section 3.8.1.1 of the Technical Specifications: o Provide for lubricating oil storage containing a minimum total volume of 504 gallons of lubricating oil per engine. s o Demonstrate the capability to transfer lubricating oil from storage to the diesel generator unit for both standby and operating modes. (2) In Section 4.8.1.1.2.a of the Technical Specifications: o Verify the lubricating oil inventory in storage. T/S Section: 3/4.8.1 Pg. 3/4 8-1 and 3/4 8-3 The above SER requirements have been incorporated into the tech specs as required. This item is CONSISTENT

28. SER Section: 10.2 Turbine Generator Pg. 10-2 states:

The applicant had not provided (1) justification for the change from weekly valve testing, as specified in SRP 10.2 and the Standard Technical Specifications, to monthly valve exercising; (2) the frequency of valve inspection following the initial 39-month inspection interval, including a description of this program; (3) proposed changes to the plant Technical Specifications; and (4) the frequency of the mechanical and backup overspeed trip tests and a description of the special test provisions that allow testing while carrying load and without tripping the unit. Therefore, the staf f required the applicant to provide (1) justification for changing the periodic turbine valve testing program from weekly to monthly, (2) a copy of the plant Technical Specifications marked to show the intended changes, (3) confirmation from Westinghouse that the turbine generator valves (turbine stop, control, intercept, and extraction l steam valves) at Beaver Valley can be periodically tested on a monthly ) basis, and (4) a description of the proposed periodic dismantling and j 23

inspection program for all turbine valves following the initial 39-month' inspection-program to be included in the plant Technical Specifications. Items 1-3 will be pursued by the staff as part of-the-Technical Specifications review. In Amendment 8 of the FSAR, the . applicant committed to conform to the inservice inspection requirements (item 4 above) of SRP 10.2 and the Standard Technical 

     ' Specifications for the turbine valves.                                                '

T/S Section: 3/4.3.4 Pg. 3/4 3-74 T/S 3.3.4 and S/R 4.3.4.1 do not appear to be in compliance with the. SER requirements as specified. This item is DIFFERENT RESOLUTION The Staff is presently reviewing the applicants submittal. SSER #5 will provide final staff acceptance or rejection. Any required Technical Specification changes will be made at that time. l i This item is CONSISTENT 

                                                                                                ]
29. SER Section: 10.4.9 Auxiliary feedwater System Pg. 10-17 states:

Provisions for AFWS testing and inspection are included in the . 

                                                                                                 )

design. Each AFWS pump is equipped with a recirculation line to the PPDWST for periodic functional testing. Local manual realignment of valves is not required for this testing. Continuous recirculation during pump operation is provided through a fixed orifice. When one I AFWS pump train is being tested, the other two trains are available 1 for automatic operation. Periodic survetilance testing of the essential pumps and their associated flow trains is identified in the Standard Technical Specifications. The applicant has committed to incorporate in the proposed plant Technical Specifications a statement that one essential AFWS pump train may be inoperable for no more than 72 hours. If this time is exceeded, the unit affected must be in hot I shutdown within 12 hours. The AFWS is tested each month for pump capacity and valve position and each 18 months for automatic start i capability. Further, the applicant has committed to incorporate in the plant Technical Specifications an AFW flow path verification test - during which water is pumped from the primary water source to the steam generators before startup after any cold shutdown of 30 days or longer. On the basis of the applicant's commitments, the staff I concludes that the AFWS meets NUREG-0611 regarding functional testing and surveillance. T/S Section: 3/4.7.1 Pg. 3/4 7-4 A 24

1 1 I I T/S 3.7.1.2 and related surveillance requirements contain the above limiting conditions for operation and testing. However, the SER j requires an AFW flow path verification after any cold shutdon.n of J 30 days or longer. S/R 4.7.1.2b requires this test following an ) o extended plant outage of greater than 30 days. 

 . This item is DIFFERENT RESOLUTION This Technical Specification will be revised to say ...of 30 days or      l greater... with issuance of the full Power License.                       j This item is CONSISTENT
30. SER Section: 11.4.2 Evaluation findings Pg. 11-9 states: l The applicant has not yet submitted a process control program to ensure that waste solidification will meet the requirements for ,

packaging, handling, shipping, and disposal. Although such a program ) is not addressed in this report, a program of this type will be J j required by the Technical Specifications, as specified by BTP ETSB 11-3. As part of this process control program, the applicant must address the additional requirements of 10 CFR 61. The applicant should confirm that he will submit a solid waste process control program to the staf f for review before initial reactor hNtup. 4 T/S Section: 3/4.11.3 Pg. 3/4 11-18 l 7/S 3.11.3.1 specifies the appropriate references be followed as  ! stated in the SER. This item is CONSISTENT I

31. Sl'R Section: 15.6.5.2 Post-LOCA Leakage from ESF Systems Outside Containment Pg. 15-18 states:

As part of the LOCA analysis, the staff also evaluated the consequences of leakage of recirculated sump water. During the recirculation mode of operation, the sump water is circulated outside containment to the auxiliary building. If a leak, such as a pump seal failure, should develop, a fraction of the iodine in the water could become airborne in the auxiliary building and exit to the atmosphere. For Beaver Valley Unit 2, the ECCS area in the auxiliary building is served by the SLCRS. Therefore, doses from passive failures were not considered (as specified in Appendix B to SRP 15.6.5). FSAR Table 15.6-9 gives 9.4 x 10-3 gpm as the expected amount of leakage for the ECCS equipment following an accident. Using Appendix B to SRP 15.6.5, the staff evaluated the potential radiological consequences from this release pathway assuming a routine leakage rate of twice the applicant's value (1.9 x 10-2 gpm). The 25 l

staff will review the Beaver Valley Unit 2 Technical Specifications relative to the testing of ESF system's recirculating sump water outside contair. ment to ensure that the leakage outside containment for all these systems is less than 9.4 x 10-3 gpm. . T/S Section: Pg. This item requires staff review. This item is NOT EVALUATED l RESOLUTION i The Staff is presently reviewing this item and will provide final , recommendations at a later date. This item is accepted as-is. I i I This item is CONSISTENT

32. SSER 3 Section: 3.9.6 Inservice Testing of Pumps and Valves Pg. 3-2 l states: 1 The review and approval of the applicant's proposed list of pressure isolation valves is a subject of the Technical Specifications review 3 and is deferred until that review is completed. Table 4.4-3 of the I staff-approved Beaver Valley Unit 2 Technical Specifications will constitute the acceptable list of valves designated as pressure isolation valves. The Technical Specifications for Beaver Valley 2 are under review and will be issued in final form with the operating license. '

T/S Section: 3/4.4.6 Pg. 3/4 4-22 T/S Table 4.4-3 contains all of the valves identified by the utility as PIVs. i This item requires staff review and approval. This item is NOT EVALUATED RESOLUTION The Staff has reviewed and accepts Table 4.4-3 as a complete list of PIVs. This item is CONSISTENT A 26

i

33. 'SSER 3 Section: -14 Initial Test Program Pg. 14-3 states: )

(6) "fSAR Section 14.2.12 should be modified to ensure that the accumulator isolation valves can open under maximum differential pressure conditions." 

 -        RG 1.79 states, in part, that facilities.which provide a confirmatory' open signal to the accumulator isolation valves (as does Beaver Valley ;

Unit 2) should demonstrate that "the valve will open under the maximum j expected accumulator precharge pressures." The applicant has not  ; included a test to demonstrate this capability because of I administrative controls and Technical ~ Specifications which the applicant states would prevent inadvertent valve closure that might prevent operation of the core flooding system. The plant Technical Specifications will require verification every 18 months that each accumulator isolation valve will open automatically upon receipt of a safety injection test signal. The test should be performed during the preoperational test program to satisfy the guidelines of RG 1.79. Resolution of this issue is ' tracked as confirmatory issue 52(a). T/S Section: 3/4.5 Pg. 3/4 5-2 S/R 4.5.1.d specifies the required accumulator isolation valve test. This item is CONSISTENT ) l t e i 4 l 27 

                                                                                 )

mac om. 338 W 8. NUCL 51.:lI E GiWL&T oR T Comunessaow i AG OAT NUMB 4A ,.egaer ar r,#C. ese ret ,4. ef eaF, h3' ' , ' - BIBUOGRAPHIC DATA SHEET EGG-NTA-7616 

     $tt .N87muCT,ows on t>t pt./gmsg 2 T,TLE .No he f,TLS                                                                                              J kl.WS SL.h.

EVALUATION OF BEAVER VALLEY POWER STATION UNIT 2 TECHNICAL SPECIFICATIONS s o.it meront courtgito 

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     . .v T o.is,                                                                                                               June                             1987 D. E. Baxter, G. L. Branson                                                                                                       .o.fi.iro 7,uv.o
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NRR and I&E Support EG&G Idaho, Inc. . ,= oa ca.at aw .a P. O. Box 1625 Idaho Falls, ID 83415 A6824 

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Division of Licensing Final Technical Office of Nuclear Reactor Regulation Evaluation Report U.S. Nuclear Regulatory Commission '"*'**""'""-****"" Washington, DC 20555 17 SUPPL (MSNT.mv NOTES IJ t J3f R.cf ,100 weres or 'eu, Final technical evaluation report on the audit of the Beaver Valley Power i Station Unit 2 Technical Specifications performed for the NRC in connection with l

the issuance of Low Power and Full Power License for the applicant. All identified discrepancies have been resolved.

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