ML050630062

RAI, - Extended Power Uprate

ML050630062
Person / Time
Site:	Beaver Valley
Issue date:	03/11/2005
From:	Colburn T NRC/NRR/DLPM/LPD1
To:	Pearce L FirstEnergy Nuclear Operating Co
Colburn T, NRR/DLPM, 415-1402
References
TAC MC4645, TAC MC4646
Download: ML050630062 (24)

Text

March 11, 2005 Mr. L. William Pearce Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Post Office Box 4 Shippingport, PA 15077

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 (BVPS-1 AND 2) -

REQUEST FOR ADDITIONAL INFORMATION (RAI) - EXTENDED POWER UPRATE (EPU) (TAC NOS. MC4645 AND MC4646)

Dear Mr. Pearce:

By letter to the U.S. Nuclear Regulatory Commission (NRC) dated October 4, 2004, FirstEnergy Nuclear Operating Company (the licensee) submitted a license amendment request for BVPS-1 and 2 to change the operating licenses to increase the maximum authorized power level from 2689 megawatts thermal (MWt) to 2900 MWt which represents an increase of approximately 8 percent above the current maximum authorized power level. The proposed amendment would also change the BVPS-1 and 2 Technical Specifications (TSs) to authorize operation with replacement Model 54F steam generators for BVPS-1 and full implementation of the Alternate Source Term for both BVPS-1 and 2, in accordance with Regulatory Guide 1.183, Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.

Various other miscellaneous and administrative TS changes were also proposed not related to the EPU. The NRC staff has determined that the additional information contained in the enclosure to this letter is needed to complete its review. As discussed with your staff, we request your response within 45 days of receipt of this letter, in order for the NRC staff to complete its scheduled review of your submittal.

If you have any questions, please contact me at 301-415-1402.

Sincerely,

/RA/

Timothy G. Colburn, Senior Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412

Enclosure:

RAI cc w/encl: See next page

Mr. L. William Pearce Vice President FirstEnergy Nuclear Operating Company Beaver Valley Power Station Post Office Box 4 Shippingport, PA 15077

SUBJECT:

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 (BVPS-1 AND 2) -

REQUEST FOR ADDITIONAL INFORMATION (RAI) - EXTENDED POWER UPRATE (EPU) (TAC NOS. MC4645 AND MC4646)

Dear Mr. Pearce:

By letter to the U.S. Nuclear Regulatory Commission (NRC) dated October 4, 2004, FirstEnergy Nuclear Operating Company (the licensee) submitted a license amendment request for BVPS-1 and 2 to change the operating licenses to increase the maximum authorized power level from 2689 megawatts thermal (MWt) to 2900 MWt which represents an increase of approximately 8 percent above the current maximum authorized power level. The proposed amendment would also change the BVPS-1 and 2 Technical Specifications (TSs) to authorize operation with replacement Model 54F steam generators for BVPS-1 and full implementation of the Alternate Source Term for both BVPS-1 and 2, in accordance with Regulatory Guide 1.183, Alternate Radiological Source Terms for Evaluating Design Basis Accidents at Nuclear Power Reactors.

Various other miscellaneous and administrative TS changes were also proposed not related to the EPU. The NRC staff has determined that the additional information contained in the enclosure to this letter is needed to complete its review. As discussed with your staff, we request your response within 45 days of receipt of this letter, in order for the NRC staff to complete its scheduled review of your submittal.

If you have any questions, please contact me at 301-415-1402.

Sincerely,

/RA/

Timothy G. Colburn, Senior Project Manager, Section 1 Project Directorate I Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-334 and 50-412

Enclosure:

RAI cc w/encl: See next page DISTRIBUTION:

PUBLIC MO'Brien ACRS PDI-1 R/F TColburn DLPM/DPR OGC RLaufer GMatakas, RGN-I JStang ALund SWeerakkody RJenkins TChan MMitchell DCoe LRaghavan KManoly PPatnaik LLois JMedoff EMarinos DTrimble KMartin GMakur LMiller IAhmed RPelton RWolfgang NPatel ACCESSION NO. ML050630062

• Input received. No substantive changes made.

OFFICE PDI-1/PM PDI-2/LA EMCB/SC SPLB/SC EEIB/SC EEIB/SC NAME TColburn MOBrien ALund*

SWeerakkody*

EMarinos*

RJenkins*

DATE 3/10/05 3/10/05 01/10/05 12/17/04 02/24/05 01/18/05 OFFICE EMCB/SC EMCB/SC EMEB/SC SRXB/SC IROB/SC PDI-1/SC NAME TChan*

SCoffin*

KManoly*

JUhle*

DTrimble*

RLaufer DATE 01/03/05 01/11/05 02/14/05 01/11/05 01/31/05 3/11/05 OFFICIAL RECORD COPY

Enclosure REQUEST FOR ADDITIONAL INFORMATION RELATED TO FIRSTENERGY NUCLEAR OPERATING COMPANY (FENOC)

BEAVER VALLEY POWER STATION, UNIT NOS. 1 AND 2 (BVPS-1 AND 2)

EXTENDED POWER UPRATE (EPU)

DOCKET NOS. 50-334 AND 50-412 By letter dated October 4, 2004, Agencywide Documents Access and Management System (ADAMS) Accession No. ML042920300, First Energy Nuclear Operating Company (FENOC, the licensee) proposed changes to BVPS-1 and 2 operating licenses to increase the maximum authorized power level from 2689 to 2900 megawatts thermal (MWt) (approximately 8%). The proposed amendment would also change the BVPS-1 and 2 Technical Specifications (TSs) to authorize operation with replacement Model 54F steam generators (SGs) for BVPS-1 and authorize full implementation of an alternate source term (AST) for both BVPS-1 and 2, and deletion of the power range, neutron flux high negative rate trip for both BVPS-1 and 2. Various other miscellaneous and administrative TS changes were also proposed not related to the EPU.

The Nuclear Regulatory Commission (NRC) staff has determined that it will need the additional information identified below to complete its review.

Section 5.12, Fire Protection Safe Shutdown (Appendix R), and Section 9.24, Additional Systems Reviewed 1.

In Review Standard for Extended Power Uprates (RS-001), Rev. 0, Attachment 1 to Matrix 5, ?Supplemental Fire Protection Review Criteria, Plant Systems states that...

power uprates typically result in increases in decay heat generation following plant trips.

These increases in decay heat usually do not affect the elements of a fire protection program related to (1) administrative controls, (2) fire suppression and detection systems, (3) fire barriers, (4) fire protection responsibilities of plant personnel, and (5) procedures and resources necessary for the repair of systems required to achieve and maintain cold shutdown. In addition, an increase in decay heat will usually not result in an increase in the potential for a radiological release resulting from a fire. However, the licensees application should confirm that these elements are not impacted by the extended power uprate...

The NRC staff notes that Section 5.12 of Attachment 2 to the BVPS-1 and 2, EPU application does not address the items above. Please address each of these items.

2.

In RS-001, Attachment 1 to Matrix 5, ?Supplemental Fire Protection Review Criteria, Plant Systems states that... where licensees rely on less than full capability systems for fire events..., the licensee should provide specific analyses for fire events that demonstrate that (1) fuel integrity is maintained by demonstrating that the fuel design limits are not exceeded, and (2) there are no adverse consequences on the reactor pressure vessel integrity or the attached piping. Plants that rely on alternative/dedicated or backup shutdown capability for post-fire safe shutdown should analyze the impact of the power uprate on the alternative/dedicated or backup shutdown capability...The licensee should identify the impact of the extended power uprate on the plants post-fire safe-shutdown procedures.

The NRC staff notes that Section 5.12, Fire Protection Safe Shutdown (Appendix R),

of Attachment 2 to the BVPS-1 and 2, EPU application does not address either of the items above. Please address each of these items.

Steam Generator Tube Integrity

1. Laser-welded sleeves were discussed for BVPS-2 in Section 4.7.2.4.6 (p. 4-95) of the Beaver Valley Power Station Extended Power Uprate Licensing Report September 2004" (EPU Licensing Report). Please confirm that the laser welded sleeves will still meet all applicable regulatory criteria, consistent with the original design/licensing basis of the sleeves, under EPU conditions.

2. Section 4.7.2.6 (p. 4-103) of the EPU Licensing Report discusses the tube plugging or repair limit for BVPS-2. The licensee concludes that the analyses and evaluations for the tube plugging or repair limit for the higher power level (2900 MWt) bound and support operation at the current power level. Discuss whether the Regulatory Guide (RG) 1.121 analysis has shown that the 40% through-wall plugging limit is adequate for EPU conditions.

3. FENOC letter, L-04-115 (September 1, 2004, ADAMS Accession No. ML042520356), stated that two tubes in SG C, BVPS-2, were plugged and stabilized due to a loose part. The part had caused degradation to one of the tubes and could not be removed. Since loose parts may affect tube integrity, please discuss the results of your analysis to confirm that the parts in your SG will not compromise tube integrity for the period of time between inspections under the EPU conditions.

4. The SG U-bend fatigue analysis for BVPS-2 (p. 4-87) showed that up to six tubes could require removal from service by plugging if the normal operating steam pressure falls below a certain value. The report states that tubes will be removed from service using sentinel plugs only or have cable tube dampers installed with plugging. What are the criteria that determine which preventive action will be used? Once a need for preventive action is identified, when will that action be taken?

Alternative Source Term Two of the issues affecting the release of radioactive iodine are:

- ability to maintain sump pH greater than 7 for 30 days following a loss-of-coolant accident (LOCA), and

- backleakage to the refueling water storage tank Do the previous evaluations of these issues, which were reviewed by the staff for selective AST implementation (ADAMS Accession No. ML032530204, 9/10/03), apply to both the EPU and containment conversion conditions? If not, please provide evaluations that do apply.

Chemical and Volume Control System (Including Boron Recovery System)

The EPU Licensing Report (p. 9-14 and 9-15) states that letdown line pressure will not have a significant impact on letdown flowrates and N-16 delay time. The report also states that the N-16 dose rate will increase due to the EPU and be managed through plant access restrictions and exposure monitoring. If this approach to managing N-16 dose rate has been reviewed and accepted by the NRC staff, please provide a reference.

Steam Generator Blowdown System (SGBS)

1. The application states that blowdown required to control secondary chemistry and SG solids will not be impacted by the EPU (p. 3-8 and 3-10 of the EPU Licensing Report.) What are the blowdown flow (lbm/hr) and velocity at present operating conditions and at EPU conditions for each unit? Compare these to the design values to demonstrate that EPU conditions are within design values.

2. The application states that the operating position of the flow control valves will be impacted (p. 3-8 and 3-10 of the EPU Licensing Report). Please discuss the amount and significance of this change.

3. Does the licensees evaluation of the impact of EPU on the SGBS apply to the Unit 1 replacement SGs?

4. Has the blowdown system experienced degradation due to flow accelerated corrosion? Did the blowdown system evaluation consider a potential increase in flow accelerated corrosion due to higher secondary system flow rates (including particulates) under EPU conditions?

Protective Coating Systems Section 10.14 of the EPU Licensing Report states that the protective coatings inside the BVPS-1 and 2 containments comply with the design-basis accident (DBA) testing requirements of American National Standards Institute (ANSI) N101.2, and that these existing coatings are acceptable and compatible with the environments associated with EPU conditions. In order to perform the necessary evaluation of the coating systems in containment, the NRC staff requests that the licensee provide the following additional information:

1. Section 10.14 of the licensing report (p. 10-12) states that the higher radiation levels corresponding to EPU conditions will be bounded by the normal plus DBA tested values. It also states that the other DBA conditions are not expected to change significantly as a result of EPU. Will all qualification parameters (temperature, pressure, chemical environment, radiation level) for Service Level 1 coatings be bounded at EPU conditions by the normal plus DBA tested values?

2. Describe the actions that will be taken if the qualification of Service Level 1 coatings are not bounded by the EPU/DBA conditions, since coating failure could threaten performance of the emergency core cooling system (ECCS) sump after a LOCA.

3. The Updated Final Safety Analysis Reports (UFSARs) state that the DBA simulation tests were for the most part (BVPS-1, page 5.2-51) or in general (BVPS-2, page 6.1-5) consistent with ANSI N101.2. Please discuss any lack of consistency with N101.2 in the qualification of Service Level 1 coatings used in containment.

4. Has any major coatings work been performed inside containment since the original application? If so, was it done using a qualified coating bounded by EPU conditions?

Effect of EPU on Flow-Accelerated Corrosion (FAC)

In Section 10.4 (p. 10-4) of the EPU Licensing Report, the licensee discusses the FAC program. The licensee also states that the EPU increases the operating pressure, temperature, and velocity in several systems, and that these changes have been used to update the FAC program. In order to evaluate the licensees FAC program, the NRC staff requests that the licensee provide the following additional information:

1. Provide the name of the predictive code used to evaluate FAC at BVPS-1 and BVPS-2.

Briefly describe how the code is applied.

2. Describe the criteria used in the FAC program for (1) selecting piping segments for inspection, (2) selecting points at which to make thickness measurements, (3) determining the frequency of thickness measurements, (4) selecting thickness measurement methods, and (5) making replacement/repair decisions.

3. Please describe any significant changes in FAC rates anticipated as a result of EPU conditions.

4. For each of the five components considered most susceptible to FAC before the EPU and after the EPU, discuss the change to the velocity, temperature, pressure, and predicted corrosion rate resulting from the change to EPU conditions.

Electrical Systems

1. Section 9.17.1 of the EPU Licensing Report: The licensee stated that the Iso-phase Bus rating of 28,000 amperes (forced air-cooled) is adequate. However, the bus could see a current of 29,557 amperes under EPU conditions with 95% of rated voltage ((1070 x 1000)/(1.7321 x 22 x 0.95)). Please explain.

2. Section 9.17.3: The licensee stated that the load flow analysis shows that the BVPS-1 unit station service transformer (USST), 1C, is slightly overloaded on the Y winding.

A. Please provide the calculated load on the Y winding.

B. Please provide the impact of the overload on the Y winding on the following:

1) the load terminal voltages

2) the winding temperatures

3) the transformer life

4) the bus duct (Y winding to switchgear) rating

3. Section 9.17.4: The licensee stated that the load flow analysis shows that BVPS-1 system station service transformer (SSST) 1A is slightly overloaded on the Y winding.

A. Please provide the calculated load on the Y winding.

B. Please provide the impact of the overload on the Y winding on the following:

1) the load terminal voltages

2) the winding temperatures

3) the transformer life

4) the bus duct (Y winding to switchgear) rating

4. Section 9.18.2 A. The licensee stated that there is an increased load on the charging pump motors.

Provide the increased load and how it impacts the voltages and the operation (such as loading, voltage regulation, spare capacity) of the emergency diesel generators (EDGs).

B. Address the impact of increased motor loads (reactor coolant pump motor, condensate pump motor, charging pump motor) on the degraded voltage relay setpoints.

C. Evaluate the impact of these increased loads on the life of the motors mentioned above.

5. Section 9.18.3: Evaluate the impact of the increase in loads on the life of the cables.

6. Sections 9.18.4 and 9.18.5 A. The licensee stated that EDG loading is not affected by the load increase....

Provide a detailed explanation of why the load increases on the 4160V and 480V motors does not affect the EDG loading.

B. The BVPS-1, UFSAR, Section 8.5.2, states that the EDG loads do not exceed 2745 kW.

Please confirm that the EDG load does not exceed 2745 kW during EPU condition.

C. If the load has increased above 2745 kW, provide the effect of operation at EPU conditions on maintenance and on the EDG support systems (such as fuel storage and transfer systems).

7. Section 9.19.1 A. Provide the service factor of containment air recirculation (CAR) motors and the impact on the motor life if the service factor is 1.0 at EPU conditions.

B. Has operation of the CAR motors been reviewed and accepted by the motor manufacturer at EPU conditions?

C. Provide a detailed discussion of the changes in the ampacity of the feeder cables of the existing 480V motors (CAR fan motor and control rod drive mechanism shroud fan motor).

8. Section 9.19.3: Address the impact of the load changes on the 125 V dc system in this section.

9. Section 9.20 A. Identify the nature and quantity of megavolt-amperes reactive (MVAR) support necessary by each Beaver Valley unit to maintain post-trip loads and minimum voltage levels.

B. Identify what MVAR contributions each Beaver Valley unit is credited in its support of the offsite power system or grid.

C. Identify any changes in the MVAR quantities associated with items A. and B. post EPU.

D. Discuss any compensatory measures necessary to adjust for any shortfalls in item C.

above.

E. Evaluate the impact of any MVAR shortfall listed in item D. above on the ability of the offsite power system to maintain minimum post-trip voltage levels and to supply power to safety buses during peak electrical demand periods. The subject evaluation should document any information exchanges with the transmission system operator.

F. What is the demonstrated capability of the generator in MW output and MVARs?

G. Were any Category D events (North American Electric Reliability Council planning standard IA) experienced in the past? If yes, then describe the mitigation measures taken. Provide an evaluation of the Category D faults for EPU conditions.

H. Provide details about the model validation for the grid stability study and about on-line contingency capabilities available during EPU conditions.

I.

On page 9-45 of the submittal, the licensee stated that the instability does not adversely affect the ability to meet General Design Criterion (GDC) 17. Please provide a detailed explanation.

J.

Did the Pennsylvania, New Jersey, and Maryland interconnection (PJM) review the grid stability studies performed by Power Systems Energy Consulting (PSEC) Group and American Transmission Systems Incorporated (ATSI)? If yes, then provide the results of the PJM review.

10. Section 10.7 A. Address the adequacy of the DC system for the first hour of the station blackout (SBO) event for EPU conditions.

B. EDG loading will increase due to EPU. The increase loading will reduce the spare capacity. BVPS-1 and 2 utilize emergency AC power from the other unit as an Alternate AC (AAC) power source. Please address the impact of reduced capacity available as the AAC power source.

11. Section 10.10.1 A. Provide details about the impact of containment conversion (CC), replacement steam generators (RSG) and EPU conditions on the environmental qualification (EQ) of the equipment.

B. Explain any EQ impact due to main steam line break (MSLB) in the main steam valve house.

C. Do any EQ components need replacement due to EPU?

D. The EPU submittal asks to increase the reactor power level to 2900 MWt. However, in this section the reactor power level is stated as 2910 MWt. Explain the difference.

12. Section 10.10.3 A. Provide a comparison of the revised inside containment LOCA/MSLB temperature and pressure profiles to the current profiles.

B. Provide a comparison of the revised temperature and pressure profiles inside the main steam valve houses and the service building to the current profiles.

C. Describe the scaling factors used to determine the gamma and beta radiation total integrated dose (TID) values for post-EPU conditions.

D. Describe the shielding/reduction methods used for beta radiation for post-EPU equipment qualification, including the junction boxes with weep holes or open sides.

E. This section indicates that the alternative test data was used to qualify equipment.

Provide details of the alternative test data.

Instrumentation and Controls

1. Plant Licensing Requirements Manual (LRM) 3.8 (Leading Edge Flow Meter (LEFM)), action 1b, requires the use of feedwater venturis for performing calorimetric heat balance calculations when the LEFM is inoperable and also requires thermal power to be reduced to 98.6% of rated thermal power (RTP) measured when the LEFM is not in service, until the LEFM is restored to operable status. This 98.6% thermal power is equivalent to the original 100% licensed thermal power, prior to the use of the LEFM with a 2% margin added, per Title 10 of the Code of Federal Regulations (10 CFR), Part 50, Appendix K, to assure that ECCS analyses remain valid. Venturi loops universally used for measuring feedwater flow are of known accuracy so that the 2% Appendix K margin is assured and generally increased when venturis are fouled during prolonged use. Venturi conditions at BVPS-1 and 2, however, raise concerns with regard to adequacy of ECCS analyses.

In a public meeting conducted by the NRC staff on September 17, 2004, regarding the use of ultrasonic flow meter (UFM) devices for feedwater flow measurement, the licensees staff stated that BVPS-1 and 2 are essentially identical, but for most of its life, BVPS-2 has produced about 1.5% more power than BVPS-1, hence, violating Appendix K requirements.

It was concluded therefore, that with both units using venturis for measuring feedwater flow, BVPS-2 was overpowering by about 1.5%, as claimed to have been confirmed by the newly installed LEFM. Using the LEFMs for feedwater flow measurement, power at BVPS-2 was reduced by 1.5% to match power with BVPS-1.

The licensees staff stated that the LEFM system installed at BVPS-1 provided readings that were essentially the same as the BVPS-1 feedwater venturi meters and, therefore, validating the accuracy of the LEFM with the venturi loop. However, a pre-installation review of various balance of plant indications at BVPS-2 (including generator electric output indication), not a vary precise measurement of reactor power, revealed an overpower of 1.5%. The LEFM installation at BVPS-2 identified that the venturi measurements of feedwater flow indicated 40 MWt power less than that indicated by the LEFM measurements, which is approximately 1.5% of BVPS-2 RTP. The venturis used for measuring feedwater flow at both units, prior to the use of LEFMs, were laboratory-calibrated.

NRC Staffs Concern As stated by the licensee, venturis for both units were laboratory-calibrated and their measurement uncertainty is expected to remain the same, per American Society of Mechanical Engineers (ASME) Standards PTC 6 and 19.1. The licensee, however, has not explained the cause of the difference in venturi measurements between BVPS-1 and 2 and has not identified corrective measures for assuring better accuracy of feedwater flow at BVPS-2, when the LEFM is not in service, so that the required Appendix K 2% margin is maintained above indicated power.

Furthermore, the licensee needs to reconcile views presented by Caldon to the Advisory Committee on Reactor Safeguards (ACRS) on July 8, 2004 (ADAMS Accession No. ML042080030), regarding use of venturis for confirming accuracy of their LEFM instruments when used in power uprates. Caldon has stated that venturis can not be used to confirm accuracy of their instrument because preponderance of data show that in general, nozzles/venturis can only be counted on to measure accuracy within an uncertainty of +/-

1.5%.

2. Section 5.10.3 in Enclosure 2 of the licensees application defines margin as the difference between the total allowance (TA) and the channel statistical allowance (CSA) and states that the acceptance criterion for the RTS/ESFAS [reactor trip system/engineered safety features actuation system] setpoints is that the margin is greater than or equal to zero.

Section 5.10.4 states that all of the RTS/ESFAS functions have acceptable margins, and, therefore, are acceptable for the nuclear steam supply system (NSSS) power of 2910 MWt.

Please explain what is included in TA and its basis. Also explain how CSA is calculated.

Also explain the relationship between the proposed EPU of 2900 MWt and the NSSS power of 2910 MWt.

3. Trip setpoints and allowable values (AVs) in Table 5.10-2 and those for overtemperature T and overpower T reactor trip were calculated using documents listed in Section 5.10.5 of. Provide these documents for the NRC staffs review. Provide references only if the listed topical reports (WCAPs) were previously reviewed and approved by the NRC staff.

4. Explain overtemperature T and overpower T formulae in Table 3.3-1 and function of lead/lag filters to accommodate the revised time constants as mentioned in Section 9.25.2 of Enclosure 2. Also clarify why these formulae have different lead/lag compensators and number of time constants than those used for calculating these two reactor trip functions in the Westinghouse Standard Technical Specifications.

5. The NRC staff has determined that setpoint AVs established by means of ISA 67.04, Part II, Method 3 (Method 3), do not provide adequate assurance that a plant will operate in accordance with the assumptions upon which the plant safety analyses have been based.

These concerns have been described in various public meetings. The presentation used in public meetings in June and July 2004 to describe the NRC staffs concerns is available on the public website under ADAMS Accession No. ML041810346.

The NRC staff is currently formulating generic action on this subject. It is presently clear, however, that the NRC staff will not be able to accept any requested Technical Specification (TS) changes that are based upon the use of Method 3, unless the method is modified to alleviate the NRC staffs concerns. In particular, each setpoint limit in the TSs must ensure at least 95% probability with at least 95% confidence that the associated action will be initiated with the process variable no less conservative than the initiation value assumed in the plant safety analyses. In addition, the operability of each instrument channel addressed in the setpoint-related TSs must be ensured by the TSs. That is, conformance to the TSs must provide adequate assurance that the plant will operate in accordance with the safety analyses. Reliance on settings or practices outside the TSs and not mandated by them is not adequate.

The NRC staff has determined that AVs computed in accordance with ISA Method 1 or 2 do provide adequate assurance that the safety analysis limits will not be exceeded. The NRC staff has also determined that an entirely different approach, based upon the performance of an instrument channel rather than directly upon the measured trip setting, can also provide the required assurance. This alternative approach, designated performance-based TSs (PBTSs), sets limits on acceptable nominal setpoints and upon the observed deviation in the measured setpoint from the end of one test to the beginning of the next. This approach has been accepted for use at Ginna, and is discussed in a safety evaluation (SE) available via ADAMS at Accession No. ML041180293. The referenced SE is specific to Ginna, and is cited here only as a general example for other plants. It is up to the licensee to modify the approach as necessary to meet the indicated objectives for the particular plant(s) in question. In addition, licensees are welcome to propose alternative approaches that provide the indicated confidence, but such alternative approaches must be presented in detail and must be shown explicitly to provide adequate assurance that the safety analysis assumptions will not be violated.

The Nuclear Energy Institute (NEI) has indicated an intent to submit a white paper concerning this matter for NRC consideration. Receipt of that white paper is anticipated in late October or early November 2004. Licensees may choose to endorse whatever approach and justification is described in that white paper, or to act independently of the NEI. If the NEI approach is found to be acceptable to the NRC staff, it will be necessary for each licensee who chooses to use it to affirm that the salient conditions, practices, etc.

described in it are applicable to their individual situations.

Please indicate how you wish to proceed in regard to the setpoint-related TS changes addressed in your request. The following are examples of acceptable actions:

A. Demonstrate that the approach that you have used to develop the proposed limits provides adequate assurance that the plant will operate in accordance with the safety analyses. Show that operability is ensured in the TSs.

B. Suspend consideration of setpoint-related aspects of your request pending generic resolution of the NRC staffs concern.

C. Revise your request to incorporate Method 1, Method 2, or PBTSs.

D. Revise your request to incorporate some other approach that you demonstrate to provide adequate confidence that the plant will operate in accordance with the safety analyses and show that operability is ensured in the TSs.

Pressurized Thermal Shock (PTS)

1. FENOCs [FirstEnergy Nuclear Operating Companys] EPU submittal states in Section 4.1.2.5, that...neutron fluence projections have increased versus those calculated fluence projections reported in WCAP-15571." In addition, it is stated that the fluence used in WCAP-15570, the applicable report for the BVPS-1 pressure temperature (PT) limit curves, bound the first portion of the uprating. However, Section 4 in WCAP-15570 states that the fluence values are from Section 6 of WCAP-15571. Comparison of the values verified that this is the case. Please clarify how the EPU uprated neutron fluence values for BVPS-1 were derived/calculated, where the uprated neutron fluences are reported, and how the changes made to the BVPS-1 1/4T and 3/4T neutron fluences (to account for the uprated power conditions) will impact the NRC staffs requests made in RAI 4.1.2-3 (on PT limit changes) and in RAI 4.1.2-7 (on uprated upper shelf energy (USE) assessments).

2. Regarding the BVPS-1 PTS assessment reported in WCAP-15569, it is stated in Section 1, that the neutron fluences used for the RTPTS value calculations were derived from the WCAP-15571 (BVPS-1 Capsule Y Report, November 2000). Tables 6 and 7 in WCAP-15569 confirms this to be the case. The NRC staff has determined that WCAP-15571 does not clearly indicate the reported neutron fluence values, as derived from the Capsule Y dosimetry measurements, and calculations were appropriately increased to account for the 8% uprated power conditions. Please clarify how the neutron fluence values for the clad/base-metal interface were calculated to account for the 8% power conditions, where the uprated neutron fluences are reported, and how the changes made to the BVPS-1 neutron fluences for the clad/base-metal interface (to account for the uprated power conditions) will impact the calculated, uprated RTPTS values for the BVPS-1 reactor vessel (RV) beltine base-metal and weld materials.

3. Table 4.1.2-3A indicates reduced applicability of the BVPS-1 PT limit curves from 22, 28 and 45 effect full power years (EFPYs) of operation (as previously referenced in WCAP-15570) to 21.78, 27.58, and 44.0 EFPYs. However, the NRC staffs review of WCAP-15570 indicates that the WCAP did not include any technical bases or calculations for reducing the applicability of the curves to 21.78, 27.58, and 44.0 EFPYs, as based on the neutron fluence values associated with the 8% uprated power conditions. Please clarify how the reductions in the applicability of the PT curves to 21.78, 27.58, and 44.0 EFPYs were calculated (as based on 8% uprated power conditions) and where the bases for these reductions were reported.

4. The staff has determined that FENOC has submitted the uprated PT limit curves to the NRC (for information only) as part of the BVPS-1and 2 8% EPU. The NRC staff has also determined that license Amendments Nos. 256 to the BVPS-1 Technical Specifications (TS) and 138 to the BVPS-2 TS granted FENOC the right to relocate the BVPS-1and 2 PT limit curves into a Pressure Temperature Limits Report (PTLR). TS 6.9.6, Pressure and Temperature Limits Report (PTLR) provides the administrative TS requirements for making changes to the BVPS-1 and 2 PT limit curves and for implementing the BVPS-1 and 2 PTLR. Paragraph c. of TS 6.9.6 requires the following:

The PTLR shall be provided to the NRC upon issuance for each reactor fluence period and for any revision or supplement thereto.

Clarify how submission of only the uprated, revised PT limit curves for BVPS-1 and 2 satisfies the intent of TS 6.9.6.c, as opposed to submitting the entire PTLR. Otherwise, submit the uprated, PTLR for BVPS-1 and 2 as part of the 8% EPU request.

5. In Section 4.1.2 of Enclosure 2 of the EPU Licensing Report for BVPS-1 and 2, FENOC has referenced the following surveillance capsule reports that are applicable to the BVPS-1 and 2 RV assessments (i.e., PTS, USE, and PT limits): (1) WCAP-15771, Revision 0, Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program (November 2000), and (2) WCAP 15675, Revision 0, Analysis of Capsule W from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program (August 2001). The NRC staff has calculated the following average Copper (Cu) and Nickel (Ni) Weight-percent (Wt.-%) values for the BVSP-1 and 2 surveillance plate and weld materials, as based on taking an average of the Wt.-% Cu and Ni values reported in Topical Report Nos. WCAP-15571 and WCAP-15675:

BVPS-1 (WCAP-15771, Revision 0)

Surveillance Plate Surveillance Weld Heat No. C6317-1 Heat No. 305424 Wt.-% Cu:

0.205 (average) 0.234 (average)

Wt.-% Ni:

0.534 (average) 0.618 (average)

BVPS-2 (WCAP-15675, Revision 0)

Surveillance Plate Surveillance Weld Heat No. C0544-2 Heat No. 83642 Wt.-% Cu:

0.050 (average) 0.080 (average)

Wt.-% Ni:

0.560 (average) 0.070 (average)

The NRC staff requests confirmation whether or not these average Wt.-% Cu and Ni values are valid. State what the Wt.-% Cu and Ni values are for the surveillance plate and weld materials that are within the scope of the BVSP-1 and 2 Reactor Vessel Radiation Surveillance Programs (include applicable surveillance material heat numbers and associated RV plate/weld component ID numbers). Provide a reference to the pedigree of chemistry assays/tests used in the Cu and Ni Wt.% value determinations for these surveillance capsule materials.

6. In Section 4.1.2 of Enclosure 2 of the EPU Licensing Report for BVPS-1 and 2, FENOC states that the following topical reports contain the latest uprated PTS assessments for BVPS-1 and 2: (1) WCAP-15569, Revision 0, Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 1 (November 2000), and (2) WCAP-15676, Evaluation of Pressurized Thermal Shock for Beaver Valley Unit 2 (August 2001). The NRC staff requests that FENOC provide its technical bases for changing the previously reported Wt.-% Cu and Ni values for the following BVSP-1 and 2 RV beltline plate and weld materials:

BVSP-1:

Lower Shell Plate B6903-1 (Heat No. C6317-1): State the basis for increasing the Wt.-% Cu value from 0.200 to 0.210. NOTE: This is the limiting material in the BVSP-1 RV.

BVSP-2:

Intermediate Shell Plate B9004-1 (Heat No. C0544-1): State the basis for decreasing the Wt.-% Cu value from 0.070 to 0.065 and increasing the Wt.-% Ni value from 0.530 to 0.550.

Intermediate Shell Plate B9004-2 (Heat No. C0544-2): State the basis for decreasing the Wt.-% Cu value from 0.160 to 0.060. NOTE: This material in represented in the BVPS-2 Reactor Vessel Radiation Surveillance Program.

7. In Section 4.1.2 of Enclosure 2 of the BVPS-1and 2 EPU Licensing Report, FENOC states that the USE assessments for the RV beltline plate and weld components were reassessed using the uprated EPU neutron fluences and that the uprated USE values and assessments for the BVPS-1 materials were included in WCAP-15571, Revision 0, Analysis of Capsule Y from Beaver Valley Unit 1 Reactor Vessel Radiation Surveillance Program (November 2000) and for BVPS-2 in WCAP 15675, Revision 0, Analysis of Capsule W from FirstEnergy Nuclear Operating Company Beaver Valley Unit 2 Reactor Vessel Radiation Surveillance Program (August 2001). However, upon review of these WCAPs, the NRC staff has determined that the WCAPs provided updated USE assessments only for those RV beltline plate and weld materials that are represented in the BVSP-1 and 2 Reactor Vessel Radiation Surveillance Programs and did not provide a complete set of USE assessments 1

The NRC staff needs to emphasize that Section III.B.3 of 10 CFR Part 50, Appendix H, Reactor Vessel Material Surveillance Program Requirements, requires that proposed RV surveillance capsule withdrawal schedules be submitted to the NRC with a technical justification and be approved prior to implementation.

The NRC staff also needs to emphasize that, in Commission Memorandum and Order, CLI-96-013, the Commission made a decision that changes to RV materials surveillance program withdrawal schedules for the RV beltline plate and weld components, as evaluated for uprated neutron fluences for the components at the 1/4T locations of the RVs.

(A)

If WCAP-15571, Revision 0, and WCAP-15675, Revision 0, have been referenced in error and the uprated USE assessments are provided in other WCAPs or topical reports, reference which WCAPs/topical reports contain the complete sets of uprated EPU USE assessments for the BVPS-1 and 2 RV beltline plate and weld components (as evaluated in accordance with the uprated neutron fluences for the components at the 1/4T location of the RVs), and indicate whether the reports have been submitted to either the NRCs Document Control Desk or into ADAMS.

(B)

Otherwise, provide the uprated EPU USE assessment calculations and results for all BVPS-1 and 2 RV plate, forging, and weld components that will accumulate an updated, end-of-current operating license (EOL) neutron fluence in excess 1X1017 n/cm2 (i.e., neutron fluence values based on neutron energies E $ 1.0 MeV, as assessed for the 1/4T location of the RVs).

Account for the impacts of FENOCs response to RAI 4.1.2-1 on the response to this RAI, as determined to be relevant to the uprated USE values for the BVPS-1 and 2 RV beltine base-metal and weld materials.

8. In Section 4.1.2 of the BVPS-1 and 2 EPU Licensing Report, FENOC states that the most recent RV surveillance capsule withdrawal schedule for BVPS-1 is given in Table 7-1 of WCAP-15571, Revision 0 (Capsule Y report, November 2000) and that the most recent RV surveillance capsule withdrawal schedule for BVPS-2 is given in Table 7-1 of WCAP-15675, Revision 0 (Capsule W Report, August 2001). Part 50, Appendix H,Section III.B.3 of 10 CFR, requires that proposed changes to RV surveillance capsule withdrawal schedules be submitted for NRC review and approval. The NRC staff makes the following requests:

(1)

For those capsules that remain in the BVPS-1 and 2 reactors and are proposed for capsule removal during either the current operating term or anticipated extended period of operation, clarify whether the projected neutron fluence values associated with the anticipated times of withdrawal account for the impacts of the uprated 8%

power uprate on projected neutron fluence values, and if so, how. If the projected fluence values in the tables do account for the impact of the 8% power uprate, clarify whether the proposed withdrawal schedules are intended to be used by FENOC as the basis for the future surveillance capsule withdrawals from the BVPS-1 and 2 RVs, as performed in accordance with the BVPS-1 and 2 Reactor Vessel Radiation Surveillance Programs, and whether these proposed withdrawal schedules have been reviewed and approved by the NRC staff, as required pursuant to Section III.B.3 of 10 CFR Part 50, Appendix H. Otherwise, if unapproved, clarify whether FENOC is requesting the NRC staffs approval of the withdrawal schedules, pursuant to the review and approval requirements stated in the previous sentence.1 would not have be processed through the 10 CFR 50.90 license amendment process if the changes were verified as being in conformance with the withdrawal schedule provisions of American Society for Testing &

Materials (ASTM) Standard Practice E185. Thus, if FENOC is requesting the NRC staffs approval of the withdrawal schedules in Tables 7-1 of Topical Reports WCAP-15571, Revision 0, and WCAP-15675, Revision 0, and is requesting review and approval through the 10 CFR Part 50, Appendix H process, the NRC staff will place its review outside the scope of 10 CFR 50.90, 50.91, and 50.92 licensing action provisions if the NRC staff determines that the proposed withdrawal schedule changes are in conformance with the surveillance capsule withdrawal schedule provisions of ASTM Standard Practice E185-82. Under these circumstances the changes would not be subject to licensing action Hearing provisions, as would be consistent with the Commissions decision in CLI-96-013.

(2)

If the projected fluence values in the Table 7-1 of WCAP-15571 and Table 7-1 of WCAP-15675 do not account for the impact of the uprate 8% power uprate, clarify how the uprated 8% EPU will impact proposed withdrawal schedules for the surveillance capsules, including projected times of withdrawal (in EFPY) and associated projected neutron fluence values for the withdrawals. If FENOC determines that further changes to the surveillance capsule withdrawal schedules described in Table 7-1 of WCAP-15571 and Table 7-1 of WCAP-15675 are necessary as a result of this RAI, clarify whether and when FENOC intends to request the NRC staffs review and approval of the proposed changes, as required under 10 CFR Part 50, Appendix H, Section III.B.3.

9. NRC Review Standard NRR-RS-001, Revision 0, Review Standard for Extended Power Uprates (December 2002), provides the staffs standard review plan for evaluation license amendment requests for increasing power above 5-percent of the current core thermal power rated for a U.S. light-water reactor. Note 1 in Matrix 1 of Section 2.1 of NRR-RS-001, Revision 0, provides the following guidance for managing age-related degradation mechanisms in pressurized-water reactor (PWR) RV internals components:

In addition to the SRP, guidance on the neutron irradiation-related threshold for inspection for irradiation-assisted stress-corrosion cracking for BWRs is in BWRVIP-26 and for PWRs in BAW-2248 for E>1 MeV and in WCAP-14577 for E>0.1 MeV. For intergranular stress-corrosion cracking and stress-corrosion cracking in BWRs, review criteria and review guidance is contained in BWRVIP reports and associated staff safety evaluations. For thermal and neutron embrittlement of cast austenitic stainless steel, stress-corrosion cracking, and void swelling, licensees will need to provide plant-specific degradation management programs or participate in industry programs to investigate degradation effects and determine appropriate management programs.

WCAP14577, Rev. 1-A, established a threshold neutron fluence level of 1x1021 n/cm2 (E > 0.1 MeV) for the initiation of irradiation assisted stress corrosion cracking (IASCC) in Westinghouse-designed RV internals. The NRC staff has used 5x1020 n/cm2 (E > 1.0 MeV) as its threshold for the initiation of IASCC, loss of fracture toughness induced by neutron irradiation embrittlement, and void swelling in stainless steel or nickel-based alloy RV internal components. Recent operating experience with cracking of steam dryers at the Quad Cities Nuclear Plant has demonstrated that cracking of boiling-water reactor (BWR)

RV internal components may occur under BWR constant pressure power uprated conditions even after a short amount of time has elapsed at the uprated power conditions. The NRC staff is concerned that this may also become an issue for the extended power uprates that are proposed for PWR-designed facilities.

FENOCs EPU license amendment application did not indicate what the neutron fluence values (E > 0.1 MeV) will be for each RV internal components under the uprated 8-percent power conditions or what the limiting uprated neutron fluence value (E > 0.1 MeV) will be for the BVPS-1 and 2 RV internal components if the components are grouped collectively as a commodity group. FENOCs EPU license amendment application also did not provide any indication as to what measures FENOC would implement to conform to or meet the issues raised in Footnote 1 of Matrix 1 to Review Standard NRR-RS-001. With respect to evaluating the impact of the proposed EPU on the RV internal components at BVPS-1 and 2:

(1)

State what the specific uprated neutron fluence values (E > 0.1 MeV) will be for each RV internal component at BVPS-1 and 2, as impacted by the uprated 8% power conditions, or what the limiting uprated neutron fluence will be for each units RV internal components if the RV internal components are grouped collectively as a commodity group.

(2)

If the projected neutron fluence values for the RV internals are projected to exceed the threshold established by Westinghouse Electric in WCAP-14577, the NRC staff requests that FENOC either propose and identify an inspection program that will be utilized to manage the aging effects discussed in the paragraph above, or else provide a commitment to participate in and implement the EPRI MRPs research initiatives on age-related degradation of RV internal components and to submit the inspection plan for the BVPS-1 and 2 RV internals for NRC staff review and approval. If FENOC determines that aging management of the RV internals is necessary for the uprated power conditions and an inspection program is proposed as the basis for aging management, discuss the scope of the program and include specific details on which RV internal components and sample size of components will be inspected. Also discuss which aging effects will be monitored, which inspection methods and inspection qualifications will be used for the examinations, the frequency of examinations used for the inspections, what acceptance criteria will be used to evaluate any recordable and relevant flaw indications or evidence of distortion if void swelling is identified as an aging effect of concern, and the schedule for program implementation.

Material Properties of Components

1. Please provide a brief outline of your Alloy 600 management program intended to manage and identify mitigative actions to address primary-water stress corrosion cracking (PWSCC) in Alloy 82/182 weld locations in the RCS. Has any mitigative action taken place or is planned to be taken to manage PWSCC in the susceptible material?

2. Please summarize the results of volumetric examinations performed during the past inservice inspection of all Alloy 82/182 welds in the RCS. The ASME Code,Section XI, allows flaws to be left in service after a proper evaluation of the flaws is performed in accordance with Subsection IWB-3600. Indicate whether such flaws exist in any of the welds analyzed for leak-before-break (LBB) approval.

3. Discuss the material properties of the RCS at EPU conditions and how these new properties have been applied to, and affect, your LBB evaluations.

Operating Procedures

1. In Sections 10.15 and 10.9 of the EPU Licensing Report, it is stated that the EPU will require revision of operating procedures, training materials and the plant simulator.

Describe the changes that will be made to the operating procedures and training materials.

2. Describe the changes which will be made to the plant simulator. When will the changes to the plant simulator be implemented? To which ANSI standard will simulator fidelity and performance be measured?

3. Describe any new operator actions needed as a result of the proposed EPU. Describe changes to any current operator actions related to emergency or abnormal operating procedures that will occur as a result of the proposed EPU

4. Describe any changes the proposed EPU will have on the operator interfaces for control room controls, displays, and alarms. For example, what zone markings (e.g normal, marginal and out-of-tolerance ranges) on meters will change? What setpoints will change?

How will the operators be informed of the changes? Describe any controls, displays, alarms that will be upgraded from analog to digital as a result of the proposed EPU and how will operators be tested to determine proficiency?

5. Describe any changes to the safety parameter display system resulting from the proposed EPU. How will the operators know of the changes?

Plant Systems

1. Please address the impacts of the proposed power uprate and the change from a sub-atmospheric to an atmospheric containment on the licensee's response to GL 96-06.

2. Are the revised TS Bases for TS 3.7.1.2, "Auxiliary Feedwater System," and TS 3.7.7, "Control Room Habitability System," changes to the plant licensing basis? Why does EPU necessitate a change in the TS Bases for TS 3/4.7.1.2 (small-break LOCA (SBLOCA) is no longer the worst case event, and the loss of an EDG is no longer the worst-case single failure)? Is the 15-minute operator action time that is assumed to isolate the fault a change from previous assumptions?

3. Where does the existing licensing basis for TS 3/4.7.1.3 concerning the primary plant demineralized water (PPDW) storage tank specifically state that the 9-hour inventory is based on the RCPs being secured? Is this a change to the plant licensing basis? How much additional inventory is required for cooldown to RHR entry conditions (TMI Action Plan criteria)?

4. The Bases for TS 3/4.7.1.3 indicates that measurement uncertainty has not been included in the TS minimum volume requirement for PPDW inventory. A complete accounting of the minimum required TS PPDW volume that takes into consideration all postulated losses, line breaks, recirculation flow, and measurement uncertainty is needed. Identify and explain all instances where measurement uncertainty is not accounted for when confirming that indicated values are in accordance with the plant licensing basis.

Structural Loading Considerations

1. In Section 4.2.3.5 of the EPU Licensing Report, you stated that since application of LBB methodology has been licensed for the main coolant loop, consideration of breaks in the main coolant loop are not required for structural evaluations. The next limiting breaks to be considered are the branch line breaks. The hydraulic LOCA forces that are used in the reactor vessel LOCA analysis are for breaks in the 12" accumulator line (cold leg) and the 14" residual heat removal line (hot leg) for BVPS-1 and for breaks in the 4" line (cold leg) and the 3" line (hot leg) for BVPS-2. Confirm whether the current licensing basis is based on the application of LBB technology. Identify branch line breaks that are used for the RCS LOCA analysis for BVPS-1 and BVPS-2 at the EPU conditions. Confirm whether the pressurizer surge line break, the main steam line break and feewdwater line break are considered in the analyses for the EPU conditions. If not, provide technical justification for not including these pipe breaks.

2. In Section 4.2.4, provide a summary evaluation of flow-induced vibration (FIV) including identification of internal components that were reviewed and evaluated for FIV at EPU conditions. Also, provide the predicted maximum response due to FIV for EPU conditions.

3. On page 4-50, Section 4.4.5, you stated that the BVPS-1 and BVPS-2 Control Rod Drive Mechanisms (CRDMs) and Capped Latch Housings (CLHs) were evaluated for the EPU design parameters and the associated NSSS design transients. In most cases, the existing analyses and evaluations remained applicable and bounding. Where this was not the case, new calculations were performed for the limiting components and the results were evaluated to establish the structural acceptability of the CRDM and CLH pressure boundary components in accordance with the ASME Code. Specify limiting components where the re-analyses were required for operating at the EPU conditions. Also, provide calculated stresses and cumulative usage factors (CUFs) for these limiting components at the EPU conditions.

4. In Section 4.5, you indicated that the analyses and evaluations for pressurizer surge line stratification and the application of LBB methodology are addressed in this section. The analyses and evaluations for the reactor coolant loop piping and supports are addressed in Section 8.3. However, Section 8.3 does not address how the analysis was performed for the primary reactor cooling loop piping, components and supports and postulated pipe breaks considered for EPU analyses. Provide a summary description of analyses performed for both BVPS-1 and 2, including modeling, assumptions, forcing functions, loads and load combinations used in the analyses for the EPU, especially accounting for the effects of the replacement SGs on the dynamic response to the LOCA and seismic events for BVPS-1. Confirm whether the current EPU evaluations are in accordance with the licensing basis analysis methodology, loads and load combinations, and the Code of record.

If not, provide justification for the deviations. Confirm whether the stresses and CUFs provided for BVPS-1 piping, components and supports including the RV and internals at the EPU conditions include the dynamic effect of the replacement SGs. If not, provide technical justification. Also, provide maximum stresses and CUFs at the critical locations for the RV and RCS components (reactor coolant pump, replacement SG/original SG, and pressurizer) supports as a result of the RCS dynamic analyses for the EPU for both BVPS-1 and 2.

5. The licensees evaluation provided in Section 4.7.1 for the replacement SGs is rather qualitative. No quantitative results were given for the EPU conditions where the replacement SGs will be implemented. Please provide an evaluation that includes the calculated stresses and CUFs for the critical replacement SG shell, nozzles and internals (baffle, feedwater sparger, steam dryer, flow reflector, tubes) and U-bend tubes. Also provide an evaluation of FIV including fluid-elastic stability and turbulent and vorticity effects on tubes.

6. CUFs for both BVPS-1 and 2 in the main closure studs based on the design-basis calculation would exceed the Code allowable limit of unity for operation at the EPU conditions. Tables 4.1.1-1A and 4.1.1-1B indicate that the tabulated fatigue usage factors are nearly equal to 1.0. The CUFs for the main closure studs are calculated based on 10,400 and 14,000 cycles for BVPS-1 and BVPS-2, respectively, and pertain to plant loading and unloading, as opposed to 18,300 occurrences in the design-basis calculation.

Confirm whether and how you monitor the CUFs in the main closure studs such that the studs will be replaced when the loading/unloading occurrences reach 10,400 and 14,000 cycles for BVPS-1 and BVPS-2, respectively.

7. In reference to Section 8.3, discuss the effect of the increased steam flow resulting from EPU that could cause excessive vibration in the main steam and feedwater lines. NB3622.3 requires that piping be designed so that vibration will be minimized. Provide a summary of the evaluation for flow effects on the main steam line vibration, which could be increased for the EPU condition. Confirm whether and how you plan to perform a vibration monitoring program at the startup for the EPU implementation.

8.

In reference to Section 8.3.3, provide the technical basis for not evaluating the piping and support systems where the increase in temperature, pressure and flow rate are less than 5% of the current rated design-basis condition. Your justification provided on page 8-94 is qualitative and nonspecific. For instance, you stated that conservatism may include the enveloping of multiple thermal operating conditions as well as not considering pipe support gaps in a thermal analysis. We can not draw a conclusion from these undefined qualitative statements. The technical justifications should be based on specific quantitative assessment or intuitively conservative deduction in order for the NRC staff to accept your conclusions.

9.

Section 10.1, Motor Operated Valves (MOVs), in the EPU Licensing Report for BVPS-1 and 2 states that the proposed 8% EPU will not change any program controls or existing licensing commitments for MOVs at BVPS-1 and 2. Section 10.1 states that the differential pressure applied in the MOV calculations is conservative, but, that plant modifications might affect some individual valve differential pressures. Section 10.1 also states that any required revised MOV settings and testing will be performed in accordance with the established MOV Program at BVPS-1 and 2. The licensee is requested to discuss with examples its evaluation of the impact of the EPU conditions on safety-related MOVs, including any changes in individual valve differential pressures. The discussion should address any changes in ambient temperature or operating voltage that might impact motor output of MOVs addressed in response to GL 89-10, Safety-Related Motor-Operated Valve Testing and Surveillance, and GL 96-05, Periodic Verification of Design-Basis Capability of Safety-Related Motor-Operated Valves.

10. Section 10.1 of the EPU Licensing Report states that evaluations to address GL 95-07, Pressure Locking and Thermal Binding of Safety-Related Power-Operated Gate Valves, have been reviewed to determine if there would be any adverse effects due to EPU operation at BVPS-1 and 2. The licensee is requested to discuss with examples its evaluation of safety-related power-operated gate valves in response to GL 95-07, including consideration of any changes in ambient temperature on the potential for pressure locking or thermal binding.

11. Section 10.2, the Air Operated Valve (AOV) Program of the EPU Licensing Report states that the program at BVPS-1 and 2 for testing, inspection, and maintenance of AOVs was reviewed for impact from EPU operation, and that these reviews did not identify any safety-related AOVs that would be adversely affected by EPU operation (other than the feedwater control valves that will be modified to increase flow capacity). The licensee is requested to discuss with examples its evaluation of safety-related AOVs for potential impact from EPU operation.

12. Section 10.5, the Inservice Testing (IST) Program of the EPU Licensing Report states that EPU analyses and plant modifications might change specific acceptance criteria for the IST Program at BVPS-1 and 2. The licensee is requested to discuss with examples its evaluation of the impact of EPU conditions on the performance of safety-related pumps, power-operated valves, check valves, and safety or relief valves, including consideration of changes in ambient conditions and power supplies (as applicable), and to indicate any resulting adjustments to the IST Program resulting from that evaluation.

13. Section 13, Testing of the EPU Licensing Report states that steady state data will be taken at 95% and 100% of the current RTP level at BVPS-1 and 2 in order to project operating performance parameters for the EPU power level before the current RTP level is exceeded. Section 13 also states that additional steady state operating data will be obtained and evaluated at approximately 2.5% increments between the current licensed RTP and the EPU power level with any data discrepancies resolved prior to proceeding with power ascension. Table 13-10 of the EPU Licensing Report provides a brief abstract of system vibration testing for EPU conditions at BVPS-1 and 2, but does not specify hold points during power ascension, systems to be monitored, data to be collected, or methods of data collection. The licensee is requested to discuss in more detail its procedures for avoiding adverse flow effects during power escalation and after achieving EPU conditions, including specific hold points and duration, inspections, plant walkdowns, vibration data collection methods and locations, and planned data evaluation.

14. The licensee is requested to discuss its evaluation of potential FIV effects due to the increase in steam flow resulting from EPU conditions at BVPS-1 and 2. The evaluation should include the RV internals, and steam and feedwater systems and their associated components, including impact on structural capability and performance during normal operations, anticipated transients (initiation and response), and design-basis conditions; and preparation for responding to the potential occurrence of loose parts as a result of the EPU. The evaluation should also include calculations, when applicable, of the fluid-elastic stability ratio, and stresses due to turbulent and vortex shedding.

Beaver Valley Power Station, Unit Nos. 1 and 2 cc:

Mary OReilly, Attorney FirstEnergy Nuclear Operating Company FirstEnergy Corporation 76 South Main Street Akron, OH 44308 FirstEnergy Nuclear Operating Company Regulatory Affairs/Performance Improvement Larry R. Freeland, Manager Beaver Valley Power Station Post Office Box 4, BV-A Shippingport, PA 15077 Commissioner James R. Lewis West Virginia Division of Labor 749-B, Building No. 6 Capitol Complex Charleston, WV 25305 Director, Utilities Department Public Utilities Commission 180 East Broad Street Columbus, OH 43266-0573 Director, Pennsylvania Emergency Management Agency 2605 Interstate Dr.

Harrisburg, PA 17110-9364 Ohio EPA-DERR ATTN: Zack A. Clayton Post Office Box 1049 Columbus, OH 43266-0149 Dr. Judith Johnsrud National Energy Committee Sierra Club 433 Orlando Avenue State College, PA 16803 J. H. Lash, Plant Manager (BV-IPAB)

FirstEnergy Nuclear Operating Company Beaver Valley Power Station Post Office Box 4 Shippingport, PA 15077 Rich Janati, Chief Division of Nuclear Safety Bureau of Radiation Protection Department of Environmental Protection Rachel Carson State Office Building P.O. Box 8469 Harrisburg, PA 17105-8469 Mayor of the Borough of Shippingport P O Box 3 Shippingport, PA 15077 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Resident Inspector U.S. Nuclear Regulatory Commission Post Office Box 298 Shippingport, PA 15077 FirstEnergy Nuclear Operating Company Beaver Valley Power Station ATTN: R. G. Mende, Director Work Management (BV-IPAB)

Post Office Box 4 Shippingport, PA 15077 FirstEnergy Nuclear Operating Company Beaver Valley Power Station Mr. B. F. Sepelak Post Office Box 4, BV-A Shippingport, PA 15077