ML17291A777

Safety Evaluation Supporting Amend 137 to License NPF-21

ML17291A777
Person / Time
Site:	Columbia
Issue date:	05/02/1995
From:	Office of Nuclear Reactor Regulation
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NUDOCS 9505100211
Download: ML17291A777 (76)

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DOCKET NO. 50-397 In its letter dated July 9, 1993 (Reference 1),

as supplemented by letters dated October 8, 1993, October 25, 1993, January 6,

1994 (Reference 2),

January 6,

1994 (Reference 3), February 2, 1994 (Reference 4),

Nay 3, 1994 (Reference 5),

Hay 13, 1994 (Reference 6), September 26,

1994, and October 12, 1994, the Washington Public Power Supply System (the Supply System, or the licensee) proposed that Facility Operating License No. NPF-21 and Appendix A (Technical Specifications

[TSs]) of Facility Operating License No. NPF-21 be amended.

The proposed changes would increase the licensed thermal power level of the reactor from the current limit of 3323 megawatts thermal (HWt) to a limit of 3486 NWt.

This request conforms to the generic power uprate program for boiling-water reactors (BWRs) established by the General Electric Company (GE) (Reference 7) and approved by the U.S. Nuclear Regulatory Commission (NRC) staff in a letter dated September 30, 1991 (Reference 8).

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~OKUAtall On December 28,

1990, GE submitted a proposal (Reference

9) to create a

generic program to increase the rated thermal power levels of the BWR/4, BWR/5, and BWR/6 product lines by approximately 5 percent.

The report contained a proposed outline for individual license amendment submittals and discussed the scope and depth of plant-specific reviews needed and the methodologies to be used in these reviews.

In Reference 8, the NRC staff approved the program proposed by GE, on the condition that individual power uprate amendment requests meet certain requirements in the GE document.

The generic BWR power uprate program gives each licensee a consistent means to recover additional generating capacity beyond its current licensed limit, up to the reactor power level used in the original design of the nuclear steam supply system (NSSS).

The original licensed power level for most licensees was based on the vendor-guaranteed power level for the reactor.

The difference between the guaranteed power level and the design power level is often referred to as stretch power.

The design power level is used in determining the specifications for all major NSSS equipment, including the emergency core cooling systems (ECCSs).

Therefore, increasing the rated 9505i002ii 950502 PDR ADOCK 05000397 P

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thermal power limits within this program does not violate the design parameters of the NSSS equipment and does not significantly affect the reliability of this equipment.

The Supply System's amendment request to increase the current licensed power level of 3323 HWt to a new limit of 3486 HMt represents an approximate 4.9-percent increase in thermal power with a corresponding 5-percent increase in rated steam flow.

The licensee will increase power to the higher level by (1) increasing the core thermal power to increase steam flow, (2) increasing feedwater system flow by a corresponding

amount, (3) increasing reactor, pressure to ensure adequate turbine control margin, (4) not increasing the current maximum core flow, and (5) operating the reactor along higher flow control lines.

This approach is consistent with the BWR generic power uprate guidelines presented in Reference 9.

The operating pressure will be increased approximately 15 psi to ensure satisfactory pressure control and pressure drop characteristics for the increased steam flow.

The increased core power will be achieved by utilizing a flatter radial power distribution while still maintaining limiting fuel bundles within their constraints.

3. 0

$~$~0N The NRC staff reviewed the request for the MNP-2 power uprate amendment using applicable rules, regulatory

guides, sections of the Standard Review Plan, and NRC staff positions.

The NRC staff also evaluated the Supply System's submittal for compliance with the generic BWR power uprate program as defined in Reference 9.

Detailed discussions of individual review topics follow.

3.1 Re cto Co e and F

o ce The NRC staff evaluated the power uprate for its effect on such areas related to reactor thermohydraulic and neutronic performance as power/flow operating

map, core stability, reactivity control, fuel design, control rod drives, and scram performance.

The NRC staff also considered the effect of power uprate on reactor transients, anticipated transients without scram (ATMS),

ECCS performance, and peak cladding temperature for design-basis accident break spectra.

3. 1. 1 Fuel Design and Operation The licensee stated in Reference 1 that no new fuel designs would be needed in order to increase power.

This is consistent with the information in Reference 7.

The plant will continue to comply with fuel operating limits, including the maximum average planar linear heat generation rate (MAPLHGR) and operating limit minimum critical power ratio (OLMCPR) for future reloads.

The OLHCPR is determined on a cycle-specific basis from the results of reload analysis, as described in Reference 7.

The MAPLHGR and LHGR limits will also be maintained as described in this Reference.

The plant-specific safety analysis for WNP-2 is in References 10, ll, and 12.

Cycle-specific thermal limits will be included in the core operating limits report (COLR).

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1 3.1.2 Power /Fl ow Operating Hap The uprated power/flow map includes the operating domain changes for uprated power.

The map considers the increased core flow (ICF) range and an uprated extended load line limit (ELLL) analysis.

The maximum thermal operating power and maximum core flow correspond to the uprated power and the maximum core flow for ICF.

Uprated power has been rescaled so that it is equal to 100-percent rated.

The changes to the power/flow operating map are consistent with the previousTy approved generic descriptions given in Reference 13 and are acceptable to the staff.

3.1.3 Stability The licensee evaluated the effect of power uprate on core stability issues according to the generic guidelines for power uprate (Reference 7).

To determine the effect on core stability, the licensee reviewed the recommendations from GE Service Information Letter (SIL)-380, Revision 1; NRC Bulletin 88-07, Supplement 1;

and current efforts of the BWR Owners Group (BWROG), including interim corrective actions (ICAs) recommended by GE and the BWROG.

In addition, the licensee revised the power/flow maps on the basis of an updated version of the STAIF code, as discussed in Reference 13.

Use of the code was approved by the NRC staff, with limitations, in Reference 14.

The power/flow maps proposed by the licensee were determined to preserve a

decay ratio of 0.9 or less, which is consistent with NRC Bulletin 88-07, Supplement 1 (Reference 15).

Operating within the proposed limits on the power/flow maps is intended to provide adequate margin to ensure thermal-hydraulic stability, and thus maintain the level of protection against power instability that is provided by the STAIF code to presently authorized power conditions.

Ongoing activities by the BWROG and the NRC are addressing ways to minimize the occurrence and potential effects of power oscillations that have been observed for certain BWR operating conditions (as required by General Design Criterion 12 of 10 CFR Part 50 Appendix A).

GE has documented information and cautions concerning this possibility in SIL-380 and related comaunications.

The NRC has documented its concerns in NRC Bulletin 88-07 and Supplement 1 to that bulletin.

While a more permanent resolution is being developed, the technical specifications and associated implementing procedures proposed by the licensee using the STAIF code will restrict plant operation in the high

power, low core flow region of the BWR power/flow operating map.

Specific operator actions have been established to provide clear instructions for the possibility that a reactor inadvertently (or under controlled conditions) enters any of the defined regions.

Specific restrictions discussed in Bulletin 88-07 and Supplement 1 apply to uprated operation.

Final resolution will continue to proceed as directed by the joint effort of the BWROG and the NRC.

This is acceptable to the NRC staff.

3. 1.4 Control Rod Drives (CROs) and CRD Hydraulic System The CRO system controls gross changes in core reactivity by positioning neutron-absorbing control rods within the reactor.

The CRD system is also required to scram the reactor by rapidly inserting withdrawn rods into the core.

The licensee evaluated the CRD system at uprated steam flow and dome pressure.

The increase in dome pressure at uprated power will increase the bottom head pressure a corresponding amount.

Although the increased pressure will slow rod insertion initially, the reactor pressure will eventually become the primary source of pressure to complete the scram.

Hence, the higher reactor pressure will improve scram performance after the initial degradation.

Increased reactor pressure has little effect on scram time, and CRD performance during power uprate will conform to current TS requirements, as will be verified through required periodic measurement of individual CRD and group scram times.

Power uprate conditions reduce the operating margin between available and required drive water differential pressure.

For CRD insertion and withdrawal, the required minimum differential pressure between the hydraulic control unit (HCU) and the vessel bottom head is 250 psi.

The licensee analyzed plant CRD pump and system data, and the CRD pumps were found to have sufficient capacity.

The flow required for CRD cooling and driving is assured by the automatic opening of the system flow control valve, thus compensating for the small increase in reactor pressure.

The licensee has determined through evaluation that the flow control valves and CRD pumps are capable of operating within their acceptable range with power uprate.

Based on its review of the information provided by the licensee, the NRC staff concludes that the CRD system will continue to perform all its safety-related functions at uprated power with ICF, will function adequately during insert and withdraw modes, and is, therefore, acceptable.

3.2 Re t Coo t

te a d Con e

s e

The NRC staff reviewed the mechanical portions of the Supply System's power uprate amendment request to determine the effects of power uprate on the structural and'ressure boundary integrity of the piping systems and components, their supports, and reactor vessel and internal components.

The NRC staff's review is discussed below.

3.2.1 Nuclear System Pressure Relief The plant safety/relief valves (SRVs) and reactor scram functions relieve pressure from the nuclear system to prevent.overpressurization of the nuclear system during abnormal operating transients.

The setpoints for the relief function of the SRVs are increased 15 psi to accommodate the increased dome pressure for power uprate.

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The maximum operating steam dome pressure was selected to enable the turbine control valves (TCVs) to operate effectively at the higher steam flow condition corresponding to uprated power.

An appropriate increase in the SRV safety setpoints ensures that adequate differences are maintained between operating pressure and safety setpoints (siaeer margin),

and that the increase in steam dome pressure does not increase the number of unnecessary SRV actuations.

The evaluation in Section 3.2.2 of this safety evaluation (SE) discusses the capability of the nuclear boiler pressure relief system to accommodate the power uprate.

3.2.2 Code Overpressure Protection The design pressure of the reactor pressure vessel (RPV) and reactor coolant pressure boundary (RCPB) remains at 1250 psig.

The American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code allows a peak pressure of 1375 psig (110 percent of the design value), which is the acceptance limit for pressurization events.

The limiting pressurization event is a main steam isolation valve (MSIV) closure with a failure of the valve position scram.

The licensee's uprate analysis assumes (1) core power is 3702 MWt (106.2 percent of the uprated power of 3486 MWT), (2) end-of-cycle nuclear parameters, (3) six SRVs out of service, (4) no credit for the relief mode of the

SRVs, (5)

TS scram speed, (6) 3-second MSIV closure time, and (7) initial reactor dome pressure of 1050 psia.

The SRV opening pressures were +3 percent above the nominal safety setpoint for the available valves.

The analysis resulted in a peak vessel bottom head pressure of 1335 psig.

This value remains below the ASME Code limit of 1375 psig.

The NRC staff reviewed the results of the licensee's evaluation and finds this acceptable.

3.2.3 Reactor Pressure Vessel and Internals The licensee evaluated the reactor vessel and internal components considering load combinations that contain reactor internal pressure difference (RIPD),

loss-of-coolant-accident (LOCA), safety relief valve (SRV), seismic, annulus pressurization (AP), and fuel lift loads.

The licensee evaluated such LOCA loads as pool swell, condensation oscillation (CO),

and chugging for the WNP-2 power uprate and found that the original LOCA analyses remain unchanged because the containment conditions with the power uprate are within the range of test conditions used to define the LOCA dynamic loads.

The licensee evaluated the SRV containment dynamic loads that affect the reactor vessel and piping systems.

The licensee determined that the increase of SRV loads resulting from the change of SRV setpoints are within the range of the original WNP-2 SRV load definition on the suppression pool boundary.

Therefore, the original SRV loads remain bounding for the power uprate condition.

The licensee reviewed the original analyses for the AP and found that the mass and energy release rates used for calculation of the original analyzed loads bound the uprated power conditions.

On the basis of this review, the NRC staff concurs with the licensee's determination that the LOCA, SRV, and AP design-basis loads remain bounding for the WNP-2 power uprate.

There is no change in the maximum allowable core flow for WNP-2 power uprate.

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The licensee determined that the fuel lift loads for power uprate conditions are less than those specified for the original design-basis for WNP-2.

The licensee determined that this was due primarily to improvements in the analytical model.

The licensee evaluated the stresses and fatigue usage factor 'for reactor

. vessel components in accordance with the requirements of the 1971 Edition of

'he ASHE Boiler and Pressure Vessel Code Section III, Subsection NB with Sunmer 1971 Addenda (Reference

16) to ensure compliance with the WNP-2 original code of record.

The load combinations for normal, upset, and faulted conditions were considered in the evaluation.

The maximum stresses at the critical locations for the shroud and top guide were sumaarized in Table 1 of Reference 6.

The fatigue usage factors of limiting components calculated for the uprated power level were listed in Table 3-4 of Reference 10.

The maximum cumulative usage factor is 0.696 located at the feedwater nozzle based on 40 years of operation.

No new assumptions were used in the analysis for the power uprate condition.

The NRC staff reviewed the results of the licensee's evaluation and concludes that the maximum stresses and fatigue usage factor submitted by the licensee are within the code-allowable limits and are, therefore, acceptable.

3.2.4 Control Rod Drive System The licensee evaluated the adequacy of the WNP-2 control rod drive mechanism (CRDM) in accordance with the ASME Boiler and Pressure Vessel Code Section: III, 1968 Edition, up to Winter 1970 Addenda (Reference 17).

The licensee found that the limiting component of the CROM was the indicator tube.

The maximum calculated stress was based on a maximum CRD internal water pressure of 1750 psig.

This basis is not affected by the power uprate.

The licensee calculated a maximum fatigue usage factor of 0.15 for the CRD main flange assuming 40 years of plant operation.

The increase in the reactor dome pressure, operating temperature, and steam flow rate as a result of the power uprate are bounded by the conditions assumed in the GE generic guidelines for the power uprate (Reference 5).

The CRDH was originally evaluated for a normal maximum reactor dome pressure of 1060 psig, which is higher than the power uprate dome pressure of 1020 psig.

The licensee also stated that the CROM has been tested at simulated reactor pressure up to 1250 psig, which bounds the analytic limit for the high-pressure scram setpoint of 1086 psig for the power uprate.

Based on its review of the licensee's information, the NRC staff concludes that the CRDM will continue to meet its design-basis and performance requirements at uprated power conditions, and is, therefore, acceptable.

3.2.5 Reactor Recirculation System The licensee will increase power to the uprated level by operating along higher rod lines on the power/flow map with allowance for increased core flow

( ICF).

The cycle-specific core reload analyses will consider the full core

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1 flow range, up to 115 million ibm/hr.

The licensee's evaluation of the reactor recirculation system performance at uprated power with ICF determined that the core flow can be maintained.

The design pressures for the reactor recirculation control (RRC) system components include the suction, discharge, and flow control valves, as well as recirculation pumps and piping.

Raising the steam pressure by 15 psi as a result of power uprate will raise the pump suction pressure by 17 psi and the pump discharge pressure by 45 psi.

The licensee states that these increases in normal operating pressures are bounded by the system design pressure.

Operation at uprated conditions will increase the RRC pump suction temperature by approximately 1 'F which is bounded by the system design temperature.

The pump speed and flow control valve position runback functions affected by power upr ate and ELLL are changed by the amendment.

The new cavitation interlock setpoint is 11 F.

The new flow control valve runback setpoint for the uprated power condition corresponds to a core flow of 48 percent of rated flow.

The licensee concluded that the changes caused by power uprate and ELLL are small and are bounded by the RRC design basis.

Startup testing of the RRC system is discussed in Section 3.8.4 of this SE.

The NRC staff reviewed the results of the licensee's evaluation and concludes.

that the existing RRC system design has sufficient margin to accommodate operation at the uprated power condition, and is therefore acceptable.

3.2.6 Reactor Coolant Piping and Components The licensee evaluated the effects of the power uprate considering higher steam flow rate, temperature, and pressure for thermal expansion, dynamic

loads, and fluid transient loads on the reactor coolant pressure boundary (RCPB) piping systems (the main steam (MS), recirculation, feedwater (FW),

CRD, reactor water clean-up (RWCU), residual heat removal (RHR), high-and low-pressure

ECCS, and standby liquid control (SLC) and reactor core isolation cooling (RCIC)).

The licensee's evaluation was performed in accordance with requirements of the ASIDE Boiler and Pressure Vessel Code Section III Subsection NB-3600, 1971 Edition with Addenda through Winter 1971 (Reference 18).

The licensee stated that stresses and fatigue usage factors were calculated for the power uprate, based on equations 9 thr ough 14 of Reference 18, for the

design, normal, upset, emergency, and faulted conditions.

The revised stresses resulting from the power uprate were compared with the code-allowable stresses for acceptability.

The licensee concluded that the code requirements are satisfied for the piping systems evaluated and that the power uprate will not have an adverse effect on the Class 1 piping system design.

The licensee evaluated pipe supports, equipment nozzles,

valves, guides,-

penetrations, and piping suspension devices by comparing the increased piping interface loads on the system components with the margin in the original design-basis calculation.

The increased interface loads are due to thermal expansion of the piping and components from the power uprate.

The licensee found sufficient margin between the original design stresses and the code

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limits to accommodate the stress increase for all service levels at the uprated power.

The licensee also evaluated the effect of power uprate conditions on thermal and vibration displacement limits for piping, system

springs, snubbers, and rigid supports and found the limits acceptable.

The NRC staff reviewed the results of the licensee's evaluation and concludes that the design of piping, components, and their supports is adequate to maintain the structural and pressure boundary integrity of the reactor coolant piping systems in'power uprate conditions, and is, therefore, acceptable.

3.2.7 Hain Steam Isolation Valves -(HSIVs)

The licensee evaluated the HSIVs, and found them consistent with the bases and conclusions of the generic evaluation (Reference 7).

Increased core flow alone does not change the conditions within the main steam lines, and thus cannot affect the HSIVs.

HSIV performance is routinely monitored as required by TSs to ensure that the original licensing basis for HSIVs is preserved.

This is consistent with the generic evaluation in Reference 7,

and is acceptable to the staff.

3.2.8 Reactor Core Isolation Cooling System (RCIC)

The RCIC system provides core cooling when the RPV is isolated from the main

'condenser and the RPV pressure is greater than the maximum allowable for initiation of a low-pressure core cooling system.

The licensee evaluated the RCIC system, and found it consistent with the bases and conclusions of the generic evaluation (Reference 7).

The licensee committed to additional testing to address all aspects of GE SIL 377 (Reference 19).

These tests will be conducted during power ascension testing for power uprate.

This is acceptable to the staff.

The licensee evaluated the capability of the RCIC system to operate under power uprate conditions.

The licensee found that the system response under the new load demands falls within the existing RCIC turbine and pump design margins.

The NRC staff reviewed the results of the licensee's evaluation and concludes that the RCIC systea will continue to meet its design-basis and performance requirements at uprate4 power conditions.

r 3.2.9 Residual Heat Removal (RHR) System The RHR system is designed to restore and maintain the coolant inventory in the reactor vessel and to remove decay heat from the primary coolant system after a reactor shutdown for both normal and postaccident conditions.

The RHR system is designed to operate in the low-pressure coolant injection (LPCI)

mode, shutdown cooling mode, suppression pool cooling mode, and containment spray cooling mode.

The LPCI mode of RHR is discussed in Section 3.3.2.2 of this SE.

The effects of power uprate on the remaining operating modes are discussed in the following paragraphs.'

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3.2.9. I Shutdown Cooling Mode The licensee evaluated the shutdown cooling mode of the RHR system.

The operational objective of this mode when used for normal shutdown is to reduce the bulk reactor temperature to 125 'F in approximately 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br />, using two RHR loops.

At the uprated power level, the decay heat is increased proportionally, which slightly increases the time required to reach the shutdown temperature.

The licensee considers that this increased time has no effect on plant safety.

Regulatory Guide 1.139, "Guidance for Residual Heat Removal,"

(Reference 20) states that cold shutdown capability (200 'F reactor fluid temperature) should be reached within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

Final Safety Analysis Report (FSAR) Section 15.2.9 indicates that cold shutdown can be reached in a much shorter time, even considering the availability of only one RHR heat exchanger.

For power

uprate, the licensee performed an alternate shutdown cooling analysis based on the criteria of Regulatory Guide 1. 139.

The results of this analysis show that for the power uprate condition, the reactor can still be cooled to 200 'F in less than the 36-hour criterion.

The NRC staff reviewed the results of the licensee's evaluation and finds this acceptable.

3.2.9.2 Suppression Pool Cooling Mode (SPCM)

The functional design basis as stated in the FSAR for the SPCM is to ensure that the pool temperature does not exceed its maximum temperature limit after a blowdown.

This objective is met with power uprate, since the licensee's analysis confirms that the pool temperature will stay below its design limit at uprated conditions.

Suppression pool temperature response is discussed further in Section 3.3. 1. 1 of this SE.

3.2.9.3 Containment Spray Cooling Mode In the containment spray cooling mode, the RHR system provides water from the suppression pool to spray headers in the drywell and suppression chambers to reduce containment pressure and temperature during postaccident conditions.

Power uprate will increase the containment spray temperature by only a few degrees.

This increase will have a negligible effect on the calculated values of drywell pressure, drywell temperature, and suppression chamber pressure, since these reach peak values before containment spray actuates.

The licensee evaluated the temperature increase and determined it does not affect the function or operation of the containment spray cooling mode.

The NRC staff reviewed the results of the licensee's evaluation and concludes that the RHR system will continue to meet its design-basis and performance requirements at uprated power conditions, and is, therefore, acceptable.

3.2. 10 Reactor Water Cleanup (RWCU) System The operating pressure and temperature of the RWCU system will increase slightly as a result of power uprate.

The licensee evaluated the effect of these increases and has found that uprate will not adversely affect

RWCU, system integrity.

Although increased feedwater flow to the reactor may slightly diminish the cleanup effectiveness of the RWCU system, the power uprate will not require a change in current TS limits for reactor water chemistry.

The NRC staff reviewed the results of the licensee's evaluation and concludes that the RWCU system will continue to meet its design-basis and performance requirements at uprated power conditions, and is, therefore, acceptable.

3.3 ee ed S f e t re The NRC staff reviewed the effect of power upr ate on containment system performance, the standby gas treatment system (as affected by increased iodine loading),

post-LOCA combustible gas control, the control'oom atmosphere control system, and the emergency cooling water system.

The NRC staff performed this review to verify that the uprate would not impair the ability of these systems to perform their safety functions to respond to or mitigate the effects of design-basis accidents.

The NRC staff also considered the effects of the power uprate on high-energy line breaks, fire protection, and station blackout.

3.3. 1 Containment System Performance Section

5. 10.2 of Reference 9 requires the power uprate applicant to show acceptability of the uprated power level for: (1) containment pressures and temperatures, (2)

LOCA containment dynamic loads, and (3) safety-relief valve dynamic loads.

Appendix G of Reference 9 prescribes the approach to be used by power uprate applicants for performing required plant-specific analyses.

The licensee performed the necessary analyses and discussed its results in its application.

The analyses results are discussed below and summarized in Table 3.1 of this SE.

Appendix G of Reference 9 states that the licehsee needs to analyze short-term containment responses using the staff-approved H3CPT code.

M3CPT is used to analyze the period from break initiation through initiation of pool cooling.

M3CPT output is also used as input for dynamic loads analyses and equipment qualification analyses.

Appendix G of Reference 9 states that the licensee needs to analyze long-term containment heatup (suppression pool temperature) for the limiting safety analysis report events to show that pool temperatures will be within the required limits for containment design temperature, NUREG-0783 (Reference 22) local pool temperature, net positive suction

head, and other considerations (e.g.,

pump seals, piping design temperatures).

Licensees performing these analyses are to use the SHEX code and ANS 5; 1-1979 decay heat assumptions consistent with Reference 21.

SHEX, which is partially based on H3CPT, is a

long-term code used to analyze the'eriod from break initiation until after peak pool heatup.

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3.3. 1.1 Containment Pressure and Temperature

Response

The WNP-2 containment consists of a drywell/vent system and a wetwell.

The drywell/vent system is designed for a pressure of 45 psig and a temperature of 340 F.

The wetwell is designed for a pressure of 45 psig and a temperature of 275 'F.

(a)

- er S

ression Pool em er t e

es onse Pool Tem erat re (2)

(3)

The licensee evaluated the long-term response of the bulk suppression pool temperature for 102 percent of an analyzed power level of 3629 NWt.

The design-basis accident (DBA) LOCA peak temperature was calculated to be 204 'F which is bounded by the original design value of 212 F.

Current TS require that the ultimate heat sink be operable at a

temperature less than or equal to 77 'F.

The new analysis conservatively assumes that the service water temperature is 90 F (the previous analysis assumed a service water temperature of 95 F).

Thus, existing TSs ensure adequate ultimate heat capacity and temperature.

Section 6.4.1.1.2 of Reference 1 states that the containment cooling analysis does not assume that the post-LOCA RHR cooling capacity is increased.

The licensee determined that the highest pool temperature resulting from the most severe non-LOCA event, an alternate shutdown event, would be 210 'F.

This result is bounded by the design value.

Th'e NRC staff reviewed the results of the licensee's analysis and concludes that the bulk suppression pool temperature response is acceptable for power uprate conditions.

L cal Pool Tem e

ture With S Dischar e

The higher SRV setpoints accompanying power uprate result in an approximately 3.3-percent increase in quencher mass flux and an increase in the total heat load.

The licensee analyzed the local suppression pool temperature response for the limiting SRV discharge event.

The resultant temperature is bounded by the 200 F local temperature limit prescribed by NUREG-0783 (Reference 22).

The NRC staff reviewed the results of the licensee's.analysis and concludes that an uprated power level will not introduce the possibility of SRV discharge condensation instability.

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C b'

The licensee evaluated steam bypass of the suppression pool due to leakage between the drywell and wetwell airspace during a

LOCA to ensure that there is sufficient time for manual actuation of containment spray to prevent the containment pressure from exceeding the design limit.

The allowed time for operator action is 30 minutes.

The licensee found that power uprate has negligible impact on bypass effects, and that sufficient time exists for manual operator action.

- 12 (b)

The NRC staff reviewed the results of the licensee's evaluation and concludes that the steam bypass response will remain acceptable after power uprate.

t i ment Atmos her Tem er t re R s o se The licensee stated that the containment drywell design temperature of 340 'F was determined based on a bounding analysis of the superheated containment atmosphere which could result from blowdown of steam to.the drywell during a

LOCA.

Power uprate involves an increase in the nominal reactor vessel operating pressure from 1005 to 1020 psig.

The original design analysis assumed a pressure of 1040 psig, which bounds the new value.

The licensee determined that the original analysis bounds the power uprate condition, and other power uprate changes (such as higher decay heat) do not significantly affect the peak containment temperature

response, and thus found that the LOCA short-term peak drywell temperature response is unchanged.

The licensee's analysis indicated that in the event of a LOCA, the maximum bulk pool temperature can reach 204 F due to power uprate.

Mith saturation conditions, the licensee found that the maximum wetwell atmosphere temperature of 204 F will remain below the wetwell design temperature of 275 'F.

The NRC staff reviewed the licensee's evaluation and concludes that the containment atmosphere temperature response for uprated power operation is acceptable.

(c)

S ort-Term Containme t P essure Res onse The licensee analyzed the short-term containment response to a double-ended guillotine break of a recirculation suction line to demonstrate that the accident consequences of power uprate and ELLL operation would not result in exceeding the containment design limits.

The analysis encompassed the drywell pressure, wetwell pressure, and drywell-wetwell differential pressure for the period from break initiation thr ough blowdown for the 3702 HMt condition.

On the basis of these results (see Table 3.1), the staff concludes that the short-term containment pressure response is acceptable.

The licensee proposed to modify the TS definitions, limiting conditions for operation, surveillance requirements, and bases relating to the current calculated peak containment internal pressure (P,) of 34.7-psig to reflect a new value of P, 38 psig.

The licensee selected this value to conservatively encompass the proposed uprated power with ELLL peak pressure of 35.1 psig plus possible unforeseen future plant changes and additional uprate.

This ensures that containment leak rate testing is conducted at greater than or equal to peak calculated containment

pressure, to maintain the validity of the post-LOCA radiological analysis.

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l The NRC staff reviewed the results of the licensee's analysis and concludes that the containment pressure response following a postulated LOCA will remain acceptable after power uprate.

3.3. 1.2 Containment Dynamic Loads (a)

Conta ic o ds Reference 9 specifies that the licensee should determine if the containment

pressure, temperature, and vent flow conditions calculated with the M3CPT code for power uprate are bounded by the analytical or experimental conditions on which the previously analyzed LOCA dynamic loads were based.

If the new conditions are within the range of conditions used to define the loads, then LOCA dynamic loads are not affected by power uprate and thus do not require further analysis.

Reference I states that the licensee performed analyses that verified that containment

pressure, temperature, and vent flow conditions are bounded.

The licensee determined that the containment response is negligibly affected by power uprate, since the loads are bounded by the test conditions used to define the original loads.

The short-term analysis demonstrates that the uprate would not significantly affect parameters important for LOCA containment dynamic loads (e.g., drywell and wetwell pressure, vent flow rate, and suppression pool parameters).

The NRC staff reviewed the results of the licensee'.s analysis and concludes that LOCA containment dynamic loads will remain acceptable for power uprate.

(b)

SR Containment 0 namic pads The licensee stated that SRV containment dynamic loads include discharge line loads, pool boundary pressure

loads, and drag loads on submerged structures.

These loads are influenced by SRV opening setpoints, discharge line configuration, and suppression pool configuration.

The SRV setpoint is the only one of these affected by power uprate.

Reference 9 states that if the SRV setpoints are increased, the licensee needs to show that the SRV design loads have sufficient margin to accommodate the higher setpoints.

The licensee reanalyzed the containment dynamic loads to reflect increased SRV opening setpoints (15 psi increase in Group I) and increased SRV setpoint tolerance (from +IX/-3X to i 3X).

The licensee discusses the effects of power uprate on the pool boundary and submerged structure loads in Reference 12.

The analysis was performed using GE methods for plants having X-quenchers.

The results of the reanalysis indicate that the loads remain below their design values.

The NRC staff reviewed the results of the licensee's analysis and concludes that SRV containment dynamic loads are acceptable.

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S m art e t Press ri ation FSAR Section 6.2.1.2 discusses the analysis for the two primary containment subcompartments for effects of pressurization.

The subcompartments are (I) the annulus between the biological shield wall and reactor vessel and (2) the drywell head.

The licensee determined that these subcompartments do not need to be reanalyzed for power uprate.

The licensee's conclusion is based on the phenomena of the blowdown process.

Subcompartment pressurization is controlled by the initial break flow mass-energy release rates which are governed primarily by the initial conditions of the reactor coolant inventory.

The mass-energy release rates are not significantly affected by power uprate.

The NRC staff agrees with the licensee that power uprate will not affect subcompartment pressurization, and concludes that subcompartment pressurization effects will remain acceptable for uprated power.

3.3. 1.3 Containment Isolation The Reference 9 methodology does not address a need for reanalysis of the isolation system.

The isolation system is not adversely affected by power uprate.

HSIVs are designed and installed in such a manner that increased steam flow will not adversely affect their capability to close within specified time limits.

A generic analysis described in Reference 7, paragraph 4.7, concluded that existing HSIVs are acceptable for power uprate for all BWR 4/5/6 facilities subject to plant-specific confirmation that associated Class 1E components such as limit switches and solenoid valves are qualified for any increase in environmental stress resulting from power uprate.

As discussed in Section 3.8.2 of this SE, the licensee has an acceptable approach to assure qualification of safety-related electrical equipment resulting from any increase in environmental stress.

3.3. 1.4 Containment Combustible Gas Control Section 2.4.5 of Reference 23 states that plant-specific submittals need to confirm the capability of the combustible gas control system, and also need to address any procedural or equipment setpoint changes that may be required to ensure adequate containment atmosphere combustible gas control.

The WNP-2 combustible gas control systems include an inerted containment and redundant postaccident hydrogen recombiners.

The inerted containment serves to accomaodate the early hydrogen produced as a result of metal-water

reaction, whereas the recombiners serve to preclude the molecular oxygen and hydrogen buildup that could result later from radiolysis.

The amount of hydrogen that is assumed to result from metal-water reaction is prescribed by Regulatory Guide 1.7 as a function of core geometry and is not affected by power uprate.

The licensee confirmed that the postaccident hydrogen recombiners have sufficient capacity to accommodate the increased radiolytic oxygen production resulting from an increase in rated power level.

Accordingly, the NRC staff concludes that there are no combustible gas control concerns that would preclude the proposed increase in rated power.

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The effect of power uprate and the increase in RPV dome pressure on each ECCS is addressed below.

Also as discussed in the FSAR, compliance to the net positive suction head (NPSH) requirements of the ECCS pumps is conservatively based on a containment pressure of 0 psig and the maximum expected temperature of pumped fluids.

The pumps are assumed to be operating at the maximum runout flow with the suppression pool temperature at its NPSH limit (212 'F).

Assuming a

LOCA occurs during operation at the uprated power, the suppression pool temperature will remain below its NPSH limit.

Therefore, power uprate will not affect compliance to the ECCS pump NPSH requirements.

3.3.2.1 High-Pressure Core Spray (HPCS)

System The licensee evaluated the HPCS system and found that system design and operation is bounded by the generic evaluation for power uprate (Reference 7).

The staff concludes that it provides an acceptable basis for the design and operation of the HPCS system for power uprate, and is therefore acceptable.

3.3.2.2 Low-Pressure Core Injection (LPCI) System Mode of RHR The licensee evaluated the RHR system for LPCI operation, and found that the

hardware for the low-pressure portions of the RHR is not affected by power uprate.

The upper limit of the low-pressure ECCS injection setpoints will not be changed for power uprate; therefor e, the low-pressure portions of these systems will not experience any higher pressures.

The power uprate will not result in an increase in the licensing and design flow rates of the low pressure ECCS.

In addition, the uprated power will not require an increase in the RHR system shutdown cooling mode flow rates and operating pressures.

The NRC staff reviewed the licensee's evaluation and concludes that the licensee's findings are consistent with the bases and conclusions of the generic power uprate evaluation (Reference 7),

and are therefore acceptable.

3.3.2.3 Low-Pressure Core Spray (LPCS) System The licensee evaluated the LPCS system, and found that the hardware for the LPCS is not affected by power uprate.

The power uprate will not affect the upper limit of the LPCS injection setpoints; therefore, the low-pressure portions of this system will not experience any higher pressures.

The power uprate will not increase the licensing and design flow rates of the low-pressure ECCS.

The licensee determined that the short-term LPCS system response to a LOCA will continue to bound the long-term system response.

The NRC staff reviewed the licensee's evaluation and concludes that the licensee's findings are consistent with the bases and conclusions of the generic power uprate evaluation (Reference 7) and are, therefore, acceptable.

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1 3.3.2.4 Automatic Depressurization Systems (ADSs)

The ADSs use SRVs to reduce reactor pressure following a small-break LOCA with HPCS failure.

This function allows low-pressure coolant injection (LPCI).and core spray (CS) to flow to the vessel.

The licensee evaluated the ADS initiation logic and ADS valve controls and found that they are adequate for

. power uprate.

The licensee determined that the ECCS design requires a minimum

'low capacity equi.valent to five SRVs/ADS valves.

SRV setpoint tolerance and out-of-service requirements are established to assure this requirement is met.

ADS initiates (after a time delay) on Low Mater Level 1 in conjunction with a signal that at least one LPCI or LPCS pump is running with permissive from Low Mater Level 3.

The licensee determined that power uprate does not affect the ability to perform these functions.

The NRC staff reviewed the licensee's evaluation and concludes that the licensee's findings are consistent with the bases and conclusions of the generic power uprate 'evaluation (Reference

7) and are, therefore, acceptable.

3.3.3 Emergency Core Cooling System (ECCS)

Performance Evaluation The ECCSs are designed to protect against hypothetical LOCAs caused by ruptures in primary system piping.

The ECCS performance under all LOCA conditions and their analysis models need to satisfy the requirements of 10 CFR 50.46 and 10 CFR Appendix K.

The licensee evaluated the Siemens Nuclear Power (SNP) 8x8 and 9x9-9x fuel, used in MNP-2, with NRC-approved methods.

The results of the ECCS-LOCA evaluation are discussed in the followi.ng paragraphs.

The licensee used the NRC-approved SAFER/GESTR (S/G) methodology to assess the ECCS capability for meeting the 10,CFR 50.46 criteria.

The licensee performed S/G-LOCA analysis for WNP-2 with SNP 8x8 and 9x9-9x fuel in accordance with NRC requirements.

The results (Reference ll) demonstrate conformance with the ECCS acceptance criteria of 10 CFR 50.46 and Appendix K.

A sufficient number of plant-specific break sizes were evaluated to establish the behavior of both the nominal and Appendix K peak cladding temperature (PCT) as a function of break 'size.

Different single failures were also investigated in order to clearly identify the worst cases.

The licensee performed the WNP-2 specific analysis using a conservatively high peak linear heat generation rate (PLHGR) and a conservatively low minimum critical power ratio (MCPR).

In addition, some of the ECCS parameters were conservatively established relative to ECCS design performance.

The nominal (expected)

PCT is 992 'F.

The statistical upper bound PCT is below 1440 F.

The licensing basis PCT for WNP-2 is 1440 'F which is well below the acceptance criterion of 10 CFR 50.46 PCT limit of 2200 'F.

The analysis also conforms to the other acceptance criteria of 10 CFR 50.46.

Compliance with each of the'elements of 10 CFR 50.46 is documented in Table 4-2 of Reference 11.

The NRC staff reviewed the results of the licensee's analysis and concludes that WNP-2 complies with the NRC S/G-LOCA licensing analysis requirements.

The licensee also evaluated the ECCS performance for single-loop operation (SLO) using the S/G-LOCA methodology.

The DBA size break is also limiting for SLO.

The licensee's evaluation, using the same assumptions in the S/G-LOCA calculation with no NAPLHGR reduction, yields a calculated nominal and Appendix K PCT of 1184 F and 1504 'F, respectively.

Since the PCT was below the 10 CFR 50.46 limit of 2200 'F, the licensee determined that no NPLHGR reduction is required for SLO.

In Reference 24, the NRC staff asked the licensee to recongile the fact that the S/G-LOCA analysis PCT results for SLO were higher than those presented for two-loop operation, and to address the concern that the licensee did not provide a statistical analysis of the upper

'ound PCT for this case.

The licensee responded in a letter dated January 6,

1994 (Reference 3), stating that the upper bound PCT for WNP-2 is 1450 F,

which is below the 1600 F limit.

The current MNP-2 TS applies a 0.01 adder to the safety limit minimum critical power ratio (SLHCPR) when in SLO due to increased uncertainties.

This is acceptable to the staff.

The licensee also evaluated the applicability of the S/G-LOCA methodology to MNP-2 which operates with Siemens Nuclear Power (SNP) 8x8 and 9x9-9x fuel.

The dimensions and characteristics of the SNP fuel are similar to GE fuels.

The HAPLHGR and PLHGR values used in the WNP-2 analysis are based on inputs for the SNP 8x8 and 9x9-9x fuel shown on Figures 5-2 through 5-5 in NEDC-32115P, respectively.

The S/G-LOCA analysis is valid for fuel designs with comparable geometry and for MAPLHGR and PLHGR values less than or equal to those shown in the previously mentioned figures in NEDC-32115P.

Since the geometry and characteristics of the SNP fuel used in WNP-2 are simi,lar to those of a typical GE BWR plant, the S/G-LOCA methodology is considered applicable to WNP-2 with SNP fuel.

The licensee evaluated the impact of ICF, up to 115 Hlb/hr, on LOCA results at the 3629 HMt power level (corresponding to a 110-percent increase in steam flow) using S/G-LOCA methodology for WNP-2.

The evaluation results for a DBA recirculation line break with the same single failure (HPCS diesel) and using the same Appendix K and nominal assumptions show a decrease in the nominal PCT when compared to the base case.

This decrease in PCT for the nominal ICF case is due to (1) the better heat transfer during flow coastdown from the higher initial flow and (2) less subcooling in the downcomer which results in reduced break flow and later core uncovery.

The NRC staff reviewed the results of the licensee's analysis and concludes that the licensee's evaluation demonstrates that the ECCS systems are acceptably designed for power uprate conditions.

3.3.4 Standby Gas Treatment System (SGTS)

The SGTS is designed to ensure controlled and filtered release of particulate and halogens from primary and secondary containments to the environment during abnormal and accident situations in order to maintain thyroid doses within the 10 CFR Part 100 guidelines.

The system consists of two 100-percent-capacity,

parallel, redundant flow trains.

Each flow train consists of a moisture separator, two electric heater

banks, a prefilter, a high-efficiency particulate air (HEPA) filter, an electric strip heater, a charcoal adsorber

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I bed, an additional electric strip heater and charcoal adsorber

bed, a

HEPA afterfilter, and two 100-percent-capacity exhaust fans.

The capacity of the SGTS was designed to provide one secondary containment volume change per day and thereby maintain the reactor building (RB) at a slight negative pressure of 0.25-inch water gauge with respect to the outside atmosphere.

The negative pressure prevents the unfiltered release of radioactive material from the RB to the environment.

The licensee stated that the proposed slight uprate in power (4.9 percent) will not change the ventilation design aspects of the SGTS.

The licensee concluded that the power uprate by itself will not have any adverse impact on the capability of the SGTS to meet this design objective.

The NRC staff reviewed the licensee's evaluation and concludes that the power uprate by itself will not have any adverse impact on the capability of the SGTS to meet the SGTS design objectives since it does not change the ventilation design aspects of the SGTS.

The NRC staff recognizes that the iodine loading will increase marginally because of the proposed power uprate.

The SGTS design uses filters that meet Regulatory Guide (RG) 1.52 criteria with respect to the, design, testing, and maintenance criteria of engineered safety feature (ESF) grade filters.

One of the criteria deals with the filter loading capability.

The licensee stated that although the iodine loading will increase slightly, it will remain well below the original design capability of the filters.

The NRC staff reviewed the licensee's evaluation and concludes that the SGTS will continue to meet the RG 1.52 criteria and is, therefore, acceptable for power uprate operation.

On these

bases, the NRC staff concludes that SGTS will continue to meet its design objectives for uprated power operation.

3.3.5 Other ESF Systems 3.3.5.1 MSIV Leakage Control System The licensee's containment analysis determined that the peak post-LOCA containment pressures will not incr ease beyond the original design basis due to uprated power operation.

The NRC staff concludes that operation at power uprate conditions will not affect operation of the MSIV leakage control system.

3.3.5.2 Main Control Room Atmosphere Control System (CRACS)

The CRACS, containing air handling and emergency filtration units, is designed to maintain the control room envelope at a slightly positive pressure (I/8" water gauge) relative to the outside atmosphere and thus minimize unfiltered inleakage of contaminated outside air into the control room following an accident.

The system accomplishes the design objective by bringing in controlled and filtered outside air to keep the control room operator doses within GDC 19 limits during an accident.

The NRC staff determined that since plant operation at the uprated power level does not change the design and operational aspects of the control room emergency filtration system, there will not be a significant increase in unfiltered inleakage of contaminated outside air into the control room following an accident.

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The staff recognizes that following a LOCA, iodine loading in the makeup air filters and recirculation air filters will increase marginally under upr ate conditions.

The licensee committed in the FSAR to design, test, and maintain the control room emergency filtration system filters in accordance with RG 1.52 guidelines.

Therefore, the NRC staff concludes that the filters will continue to be valid for the control room atmosphere control system at uprate operation.

On this basis, the staff concludes that the uprated power level will have little or no impact on the CRACS meeting its design objectives.

3.4 t

mentatio n

Contro beany of the TS changes proposed in the licensee's application (Reference I) involve changes to the reactor protection system (RPS) trip and interlock setpoints.

These changes are intended to maintain the same margin between the new operating conditions and the new setpoints that existed before the power uprate.

The conservative design calculations for the initial licensing of MNP-2 resulted in setpoints that provided excess reactor coolant flow capacity and corresponding margins in the power conversion system.

For MNP-2, these margins (e.g.,

5 percent rated steam flow) result in the capability to increase the core operating power level by approximately 4.9 percent.

This SE is limited to the setpoint changes for the identified instrumentation and is predicated on the assumption that the analytical limits used by the licensee are based on application of approved design codes.

A review of the licensee's submittal indicates that plant-specific calculations were performed using methods recommended by the Instrument Society of America as outlined in GE Topical Report NEDC-31336-P (Reference 25).

The licensee proposed the following setpoint changes:

(a)

PRM Flow B ase Sim ted Thermal Po e

(I)

Flow Biased Change trip from 0.66M + 51X to 0.58M + 59X.

Change Allowable Value from 0.66M + 54X to 0.58M + 62X.

(2)

Flow Clamped No change in trip setpoint.

Change Allowable Value from <115.5X to gll4.9X.

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Re ctor ess Steam Oome Pressure Hi h

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Change trip from 1037 psig to 1060 psig.

Change Allowable Value from 1057 psig to 1074 psig.

(c) i Steam Hi low Change trip from 105.5 psid to 115.6 psid.

Change Allowable Value from 108 psid to 124.6 psid.

(d) od lock o

tor - Fl w-Biased-U sc 1

Change trip from 0.66W + 40X to 0.58W + 48X.

Change Allowable Value from 0.66W + 43X to 0.58W + 51X.

(e)

APRH Rod Bloc low-Biased Neutron Fl x U scale Change trip from 0.66W + 42X to 0.58W + 50.0X.

Change Allowable Value from 0.66W + 45X to 0.58W + 53.0X.

(f)

T rbine Sto Valve and Turbine Co trol Valve Fast Closure Sc am ass The turbine first-stage pressure setpoint was changed to reflect the expected pressure at the new 30-percent power point.

1 The licensee's submittal dated July. 9, 1993 (Reference

1) did not describe the methodology used for instrument setpoint calculations.

Therefore, in a letter dated November 9, 1993 (Reference 30), the NRC staff requested additional information regarding the instrument setpoint methodology.

The licensee, in a letter dated February 2, 1994 (Reference 4), responded to the staff's request and confirmed that it used GE licensing Topical Report NEDC-31336-P (Reference

25) for calculating instrument setpoints.

The NRC staff previously reviewed this topical report and accepted it with some minor exceptions.

The NRC staff is reviewing these exceptions and will resolve them generically.

They do not affect the staff's evaluation of the proposed WNP-2 change..

The proposed setpoint changes are intended to maintain the existing margins between operating conditions and reactor trip setpoints.

Thus, margins to the new safety limits will remain the same as the current margins.

These new setpoints also do not significantly increase the likelihood of a false trip nor failure to trip upon demand.

The NRC staff concludes that the licensee's instrument setpoint methodology and the resulting setpoint changes incorporated in the TSs for power uprate are consistent with the WNP-2 licensing basis and are, therefore, acceptable.

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ar S stems 3.5.1 Spent Fuel Pool Cooling and Cleanup System (FPCCS)

The FPCCS is designed to remove the decay heat generated by the stored spent fuel assemblies.

,The system consists of fuel pool cooling pumps, heat

.'xchangers, skimmer surge tanks, filter demineralizers, associated

piping, valves and instrumentation.

The system is designed to transfer the decay heat generated by the spent fuel to the reactor building closed cooling water

system, and can be cooled by the service water system during abnormal operation.

Spent fuel discharged to the pool following operation at the upr ated power level will increase the heat load on the FPCCS.

The licensee evaluated spent fuel pool heat loads and radiological consequences for plant operations at the uprated power level.

The results of the evaluation indicate that the fuel pool temperature can be maintained below the 150 F

FSAR limit with two fuel pool cooling heat exchangers in service under the maximum expected normal heat load using uprated power.

The licensee analyzed the failure of a single cooling train under the same heat loading conditions and determined that the, 155 'F FSAR limit for a single failure would not be exceeded.

The FPCCS is capable of maintaining pool temperatures less than 145 'F for the emergency full core offload with the fuel pool at maximum capacity.

Supplemental RHR cooling is required immediately after the full core offload.

The FSAR limit for this scenario is 175 'F.

The licensee verified that the spent fuel racks are designed to withstand the higher fuel temperatures expected from the power uprate.

The licensee also determined that the capacity of the standby service water system is adequate to cool the spent fuel pool with the anticipated increase in heat. load.

The licensee conducted its evaluation by reviewing the original design requirements and bases for the

FPCCS, the current plant operating conditions, and the assumptions made for the proposed power level increase.

Findings indicate that the proposed power upr ate will hav'e a minimal effect on the FPCCS.

Although normal radiation levels around the pool are expected to

increase, pr'imarily during fuel handling operations, the licensee considers this increase acceptable and does not anticipate a significant increase in operational doses.to personnel or equipment.

The NRC staff reviewed the results of the licensee's evaluation and concludes that operation at uprated power will not prevent the FPCCS from performing its design function.

In addition to the licensee's evaluations, the NRC staff identified an issue associated with the adequacy.of spent fuel pool cooling in NRC Information Notice (IN) 93-83, "Potential Loss of Spent Fuel Pool Cooling Following a Loss of Coolant Accident (LOCA)," October 7,

1993, and in a.10 CFR Part 21 notification dated November 27, 1992.

The staff is evaluating this

issue, as well as broader issues associated with spent fuel storage safety, as part of the NRC generic issue evaluation process.

If the generic review concludes that additional requirements in the area of spent fuel pool safety are warranted, the staff will address those requirements to the licensee under separate cover.

Based on its review of the licensee's evaluation and consideration of potential generic issues identified in NRC IN 93-83, the NRC staff concludes that operation at uprated power will have little or no impact on the spent fuel pool cooling system operation at WNP-2.

3.5.2 Cooling Mater Systems The licensee evaluated the effect of power uprate on the various plant water

systems, including safety-related and non-safety-related service water

systems, the closed-loop cooling water system, and the plant ultimate heat sink.

The licensee's evaluation took into account the increased heat loads, temperature, pressures, and flow rates.

The following sections discuss the specific evaluations performed.

3.5.2. I Safety-Related Loads The safety-related heat loads are rejected to one of two safety-related service water systems.

These systems are the emergency equipment service water system and the residual heat removal service water system.

All heat removed from these systems is then rejected to the ultimate heat sink (UHS).

The NRC staff's evaluation of the effects of uprated power level operation on each of these subsystems appears below.

(a) er enc ui ment Service Mate S stem EESWS The licensee evaluated the EESMS for its ability to provide cooling water to ECCS and ESF components, heating, 'ventilation, and air conditioning (HVAC) systems, and diesel generator coolers.

The system provides water from the UHS spray pond through the various equipment coolers and returns it to the pond through a spray network.

The licensee found that the loads on the EESMS were not power dependent and thus remain unchanged for LOCA conditions following uprated operation.

The NRC staff reviewed the licensee's evaluation and concludes that the proposed power uprate will have little or no effect on the EESMS meeting its design objectives.

(b) v Se v ce Water S st The RHR service water system provides safety-related cooling water to the RHR system under normal and postaccident conditions.

The system provides water from the UHS (spray pond) through the RHR heat exchangers and returns it to the pond via a spray network.

The licensee evaluated the effects of power uprate on the RHR service water system, and found that the system's post-LOCA RHR cooling capacity remains unchanged.

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The NRC staff reviewed the results of the licensee's evaluation and concludes that power uprate will not have a significant effect on the cooling requirements for the RHR service water system.

3.5.2.2 Non-Safety-Related Loads The licensee stated that the increase in plant service water system (PSMS) heat loads should he proportional to the uprated power level operation.

The licensee evaluated the PSMS to ensure that it is capable of supplying sufficient cooling water to remove the additional heat load from the proposed power uprate without making modifications to the existing system.

Since the PSMS does not perform any safety function, the NRC staff did not review the effect of the uprated power level operation on the PSMS design and performance.

3.5.2.3 Hain Condenser/Circulating Mater System/Normal Heat Sink The main condenser, circulating water system, and normal heat sink condense steam in the condenser and reject heat to the circulating water system.

This maintains a vacuum in the condenser to provide for efficient turbine performance by maintaining condenser backpressure.

The licensee evaluated the performance of the main condenser and found that the condenser, circulating water system, and normal heat sink were adequate for uprated power operations.

Since the main condenser, circulating water system, and normal heat sink do not perform any safety function, the staff did not review the effect of the uprated power level operation on the design and performance of these systems.

3.5.2.4 Reactor Building Closed Cooling Water (RBCCM) System The RBCCM system is designed to remove heat from the non-safety-related equipment located in the reactor building during normal plant operations and provide a barrier between systems carrying radioactive fluids and the non-radioactive service water system.

The licensee evaluated the RBCCM system and found that the increase in heat load due to the power uprate will not have a significant impact on the capability of the RBCCM system to perform its design function.

The licensee determined that the maximum heat load to the RBCCM system including the additional heat load resulting from the power uprate will be 46.5 HBtu/hr, which is below the 50.0 HBtu/hr capacity of the RBCCM system.

The licensee stated that it will modify flow to equipment affected by the power uprate and cooled by the RBCCM system as required to support the power uprate.

The NRC staff reviewed the results of the licensee's analysis and concludes that the effect of uprated power operations on the RBCCW system is negligible and the system has sufficient heat removal capability to accommodate operations at the increased power level.

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I 3.5.2.5 Ultimate Heat Sink (UHS) System The UHS is a source of safety-related cooling water for the standby service water system during normal shutdown and accident conditions.

The UHS consists of a spray pond system capable of performing its safety function without offsite power available.

The UHS provides a cooling capability for a period of 30 days without any outside makeup water.

I The licensee evaluated the UHS at upr ated power conditions and found that the post-LOCA UHS water temperature will decrease from the temperature predicted in the licensee's current FSAR.

The NRC staff attributed this decrease to the licensee's use of a more refined decay heat model (ANSI/ANS 5.1-1979) that assumes a lower integrated heat addition.

This model was not available when the WNP-2 FSAR was developed, and the NRC staff has not accepted the model for use at WNP-2.

The NRC staff requested (Reference

26) that the licensee address this issue.

The licensee responded to the request for additional information (RAI) (Reference 5) and discussed the results of an analysis using the decay heat model used in the FSAR.

The licensee concluded that even though the proposed power uprate will increase the evaporation rate of the spray pond system from I percent to 4 percent, the minimum required inventory of the UHS system will not be affected.

In conducting both analyses, the licensee conservatively assumed a 5-percent increase to all major components which comprise the total heat load to the UHS.

The licensee found that the UHS system will continue to provide a sufficient quantity of water at a temperature less than 88.6 'F (FSAR peak spray pond design temperature) following a DBA-LOCA, and no changes to the TS for the UHS are required.

The NRC staff reviewed the results of the licensee's evaluation and concludes that the UHS system design is adequate for the uprated power operation.

3.5.3 Standby Liquid Control System (SLCS)

The licensee evaluated the SLCS and found that the ability of the SLCS to achieve and maintain safe shutdown is not directly affected by core thermal power; rather, it is a function of the amount of excess reactivity present in the core; and as such, is dependent upon fuel-loading techniques and uranium enrichment.

The SLCS is designed to inject at a maximum pressure equal to that of the lowest safety/relief valve setpoint.

The SLCS pumps are positive displacement

pumps, and the small (15 psig) increase in the lowest safety/relief valve setting as a result of power uprate will not impair the performance of the pumps.

The NRC staff reviewed the licensee's analysis and concludes that the ability of the SLCS to inject water to the reactor will not be impaired by power

'prate.

In the future,

however, the licensee may want to increase fuel enrichments in order to meet fuel energy requirements for longer fuel cycles.

The increased excess reactivity associated with this increase in fuel enrichment will affect

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th'e reactivity requirements of the SLCS.

The licensee will evaluate the SLCS requirements for future operating cycles on a cycle-specific basis.

3.5.4 Power-Oependent Heating, Ventilation, and Air Conditioning (HVAC)

Systems The licensee evaluated the impact of higher process fluid temperatures in piping for all HVAC systems, including units in the reactor building, turbine building, drywell; steam tunnel, radwaste building, and control room area.

The licensee stated that the uprated heat loads in the radwaste building, control room area, drywell, steam tunnel, and reactor building have minimal impact on maintaining the design environmental temperature parameters since the uprated parameters are within the scope of the original SE performed for WNP-2.

The licensee stated that operations at the uprated power level would cause some areas in the turbine building to exceed their design temperatures.

The licensee evaluated the effect of the elevated temperatures and found that there is no safety equipment in these areas for which the equipment design capability is exceeded.

The NRC staff reviewed the licensee's evaluation and concludes that operations at the uprated power level will have minimal impact on the plant HVAC systems and is, therefore, acceptable.

3.5.5 Fire Protection Systems The licensee stated that operation of the plant at the uprated power level will not affect the fire suppression or fire detection systems.

The licensee stated that there are no physical plant configuration or combustible load changes resulting from the uprate,.and that the safe shutdown systems and equipment used to achieve and maintain cold shutdown conditions do not change.

The licensee's plan for addressing unresolved issues associated with certain post-fire safe-shutdown operator actions are documented in the licensee's letter of July 15, 1994.

The licensee's proposed final resolution of these issues is documented in the licensee's letter dated Janu'ary 25, 1995.

The NRC staff reviewed the issues and found that they do not adversely impact the proposed power uprate.

The NRC staff will complete its review of the post-fire safe-shutdown operator actions as a separate licensing issue.

The NRC staff agrees that power uprate will not affect fire detection and suppression systems and, therefore, concludes that the fire protection program is not affected by power uprate.

3.5.6 Power Conversion Systems The steam and power conversion systems and associated components (e.g.,

the turbine/generator, condenser and steam jet air ejectors, turbine steam

bypass, feedwater and condensate

systems, etc.)

were originally designed to use 105 percent of the rated power available from the NSSS.

Since the requested uprated values are less than or equal to the values used in existing analyses,

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the NRC staff concludes that the operation at uprated power should not have a

significant impact on the power conversion systems.

3.6 aste S ste s

a d R diatio So rces 3.6. 1 Liquid Waste Management System

'he liquid waste processing system collects, monitors, processes,

stores, and returns processed radioactive waste to the plant for reuse or for discharge.

The single largest source of liquid waste is from the backwash of the condensate filter demineralizers.

With the power uprate, the average time between backwash/precoat will be reduced slightly.

Also, the activated corrosion products in liquid wastes are expected to increase proportionally to the powe} uprate.

However, the total volume of processed waste is not expected to increase appreciably since the only significant increase in processed waste comes from the more frequent backwashes of the condensate filter demineralizers.

The licensee evaluated plant operating effluent reports and the slight increase expected from the power uprate, and found that the requirements of 10 CFR Part 20 and Appendix I to 10 CFR Part 50 will continue to be met.

Based on its review of available plant data and experience with other power

uprates, the NRC staff concludes that there will not be a significant adverse effect on liquid effluents from the proposed power uprate.

3.6.2 Gaseous Waste management Systems The gaseous waste management systems collect, control, process,

store, and dispose of gaseous radioactive waste generated during normal operation and abnormal operational occurrences.

The gaseous waste management systems include the offgas system, standby gas treatment system (SGTS),

and various building ventilation systems.

The licensee states that the systems are designed to conform to the requirements of 10 CFR Part 20 and Appendix I to 10 CFR Part 50.

Building ventilation systems control airborne radioactive gases by using combinations of such devices as HEPA and charcoal filters, and radiation monitors that signal automatic isolation dampers or trip supply and/or exhaust fans, or by maintaining negative air pressure, where required, to limit the migration of gases.

The licensee states that activity of airborne effluents released through building vents is not expected to increase significantly with power uprate because the amount of fission products released into the coolant depends on the number and nature of the fuel rod defects, and is approximately linear with respect to core thermal power.

The release of gaseous effluents is an administratively controlled variable, and is not a function of core power.

Based on its review of available plant data and experience with other power

uprates, the NRC staff concludes that the proposed power uprate will not have a significant adverse effect on airborne effluents.

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I 3.6.3 Radiation Sources in the Core and Coolant During reactor operation, the coolant passing through the core region becomes radioactive as a result of nuclear reactions.

Radioactive materials in the reactor core are produced in direct proportion to the fission rate.

Thus, the levels of radioactive materials (for both fission products and activation products) produced are expected to increase by a maximum of 4.9 percent.

The licensee evaluated the effect of the power uprate on coolant activation

products, activatdd corrosion products, and fission products and found that they are expected to be approximately equal to current measured data which is within the design basis of the plant.

Based on its review of available-plant data and experience with other power

uprates, the NRC staff concludes that the power uprate will not have a

significant adverse effect on radiation sources in either the core or reactor coolant.

3.6.4 Radiation Levels The licensee evaluated the effect of power uprate on radiation levels in the WNP-2 facility during normal and abnormal operation as well as from postulated accident conditions.

The licensee found that radiation levels from both normal and accident conditions may increase slightly.

However, any such increases would be slight and would be bounded by conservatism in the original plant design and analysis.

The licensee also found that individual worker exposures will be maintained within acceptable limits by the site "as low as reasonably achievable (ALARA)" program, which controls access to radiation areas.

The licensee determined that operation at the uprated power level will not significantly affect offsite doses associated with normal operation since these releases are administratively controlled and are expected to remain below the levels in Appendix I to 10 CFR Part 50.

Based on its review of available plant data and experience with other power

'prates, the NRC staff concludes that no significant adverse effect on radiation levels will result (either onsite or offsite) from the planned power uprate conditions; 3.7 e

orm ce valua i 3.7.1 Reactor Transients The licensee evaluated the limiting plant transients.

The licensee evaluated disturbances of the plant caused by a malfunction, a single failure of equipment, or personnel error according to the type of initiating event.

The licensee will use its NRC-approved licensing analysis methodology to evaluate the effect of the limiting reactor transients.

The licensee identified the limiting events for WNP-2 in Reference 1, which are the same as those in the generic report on power uprate.

The generic guidelines also identified the analytical

methods, the operating conditions that are to be assumed, and the criteria that are to be applied.

Representative changes in core critical power ratios (CPRs) fo} analyzed transients were given; however, specific core operating limits will be supplied for each specific fuel cycle in the core operating limits report (COLR).

The licensee discussed the power uprate with ELLL operation for a representative core using the GEMINI transient analysis methods listed in the generic report.

The licensee will.confirm the acceptability of the safety limit minimum critical power ratio (SLHCPR) for each operating fuel cycle at the time of the reload analysis using NRC-approved methodology.

The licensee evaluated limiting transients for each category to determine their sensitivity to core flow, feedwater temperature, and cycle exposure.

The licensee used the results from these analyses to develop the licensing basis for transient analyses at uprated power with ELLL operation.

The licensee discussed the limiting transient results in Table 9-2 of Reference l.

The licensee will perform cycle-specific analyses at each reload and will provide the results to the NRC in the COLR.

This approach is acceptable to the staff.

3.7.2 Design-Basis Accidents The licensee evaluated plant-specific radiological consequences at uprated

'conditions for loss-of-coolant-accident (LOCA), the fuel handling accident (FHA), and the control rod drop accident (CRDA).

The licensee calculated whole-body and thyroid doses at the exclusion area boundary (EAB) and low population zone (LPZ) for all accidents, and in the main control room for the LOCA.

The licensee found that the plant-specific results for power uprate will remain well below the reference values of 10 CFR Part 100.

The NRC staff compares doses resulting from the accidents analyzed to the applicable guideline in Table 3.2 of this SE.

On this basis, the NRC staff concludes that the analy'zed consequences of postulated accidents remain within NRC staff acceptance criteria and are, therefore, acceptable.

3.7.3 Anticipated Transients Without Scram (ATWS)

GE performed a generic evaluation for the ATWS events.

The evaluation is discussed in Section 3.7 of Supplement 2 of Reference 7.

This evaluation concludes that the results of an ATWS event are acceptable for the fuel and the reactor pressure vessel (RPV),

and that the containment response is also acceptable for a power uprate of 4.3 percent.

The WNP-2 power increase is 4.9 percent, which is 0.6 percent above the generic evaluation.

The licensee performed a WNP-2-specific ATWS analysis for a 10-percent power uprate for the limiting transients to provide assurance that the generic results will be met for WNP-2.

The licensee found the results of this analysis for the ATWS event acceptable for the fuel and

RPV, and the containment response is also acceptable for a 4.9-percent power uprate.

GE performed a cycle-specific analysis with the major parameters and characteristics of the SNP fuel; the

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results were bounded by the generic ATWS analysis.

The licensee found that the plant's response to an ATWS event is acceptable.

The NRC staff determined that the licensee's results were bounded by the generic analysis and therefore concludes that the plant design for response to an ATMS is acceptable.

3.7.4 Station Blackout (SBO)

The licensee evaluated the WNP-2 plant responses to a postulated SBO at a

steam flow increase of 110 percent for power uprate.

This corresponds to an increase of reactor thermal power to 3629 HMt from 3323 NMt.

The MNP-2 response to a postulated SBO uses the RCIC and HPCS for core cooling.

The licensee performed a coping evaluation to demonstrate plant response for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> using HPCS with backup provided by the RCIC system.

The RCIC system is the preferred source for initial operation.

The licensee found that no changes to the systems or equipment used to respond to a SBO are necessary due to power uprate.

The suppression pool temperature remained within design conditions.

The licensee found that all equipment that takes suction from the suppression pool will operate acceptably when power is restored.

The individual considerations evaluated for power uprate included the following:

the regulatory basis; the event scenario; condensate inventory and reactor coolant inventory; station battery load; compressed air supply; and loss of ventilation to the control room, reactor protection system rooms and switchgear

rooms, HPCS pump and auxiliary rooms, RCIC room, containment, suppression

pool, and spent fuel pool.

The NRC staff reviewed the results of the licensee's evaluation and concludes that the power uprate will have little or no effect on the plant's coping capabilities for an SBO event, and that no changes are needed to the required coping time or,to systems and equipment used to respond to an SBO event.

Therefore, the NRC staff concludes that the plant design for response to an SBO is acceptable.

3.8 dd t on 1

As owe U rat 3.8.1 High-Energy Line Breaks (HELBs)

The licensee will need to slightly increase the RPV dome operating pressure to supply more steam to the main turbine.

This slight increase in the operating pressure and temperature resulting from plant operations at the power uprate will cause a small increase in the mass and energy release rates following an HELB outside the primary containment.

This results in a small increase in the subcompartment pressure and temperature profiles and a negligible change in the humidity profile.

The licensee evaluated the HELB for the subject piping systems and found that there is no change in the postulated break locations from the uprated conditions.

The licensee evaluated HELBs for the systems with the most limiting environment qualification profiles (temperature,

pressure, and humidity) and found that for the RCIC line break, the original analysis envelopes the uprated conditions, and for the RWCU line break, the

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duration, and have minimal impact on equipment qualification.

The licensee evaluated the effects of power uprate on plant systems and components due to moderate-energy line breaks (HELBs).

The licensee determined that the original HELB analysis bounds the conditions resulting from the proposed power uprate.

The licensee evaluated the calculations supporting the disposition of potential targets of pipe whip and steam jet impingement from the postulated HELBs and determined that they are adequate for the safe-shutdown effects in the uprated power condition.

The licensee determined that existing pipe whip restraints and jet impingement shields and their supporting structures are'dequate for the power uprate.

The NRC staff reviewed the results of the licensee's evaluation and concludes that the analysis for HELBs is acceptable for the proposed power uprate.

3.8.2, Equipment gualification (Eg) 3.8.2.1 gualification of Electrical Equipment The licensee will evaluate safety-related electrical equipment inside and outside containment to ensure qualification for the normal and accident conditions expected in the areas in which the equipment is located.

For equipment located inside the containment, the licensee determined that normal and accident design conditions for temperature,

pressure, and humidity in the affected areas are unchanged for power uprate.

However, normal and accident radiation levels will increase proportionally to the uprated power.

For equipment located outside the containment, qualification is based on accident temperature,

pressure, and humidity resulting from an MSLB, or other HELB.

The licensee determined that normal and accident temperature,

pressure, and humidity conditions in the affected areas do not change with the uprated power level.

However, the maximum accident radiation levels used for qualification of equipment outside containment is based on a DBA/LOCA and will increase proportionally to the uprated power.

If the reevaluation of Eg inside and outside containment identifies equipment that has the potential to be affected by the power uprate, the licensee will resolve the qualification of the equipment by performing refined radiation calculations (location specific) or by reducing the qualified life of the equipment.

The NRC staff reviewed the licensee's evaluation and commitment to resolve qualification of safety-related electrical equipment, and concludes that the licensee's approach to the qualification of safety-related electrical equipment for power uprate is acceptable.

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k' 3.8.2.2 Eg of Mechanical Equipment With Non-Hetallic Components The licensee will reevaluate safety-related mechanical equipment with non-metallic components to identify equipment affected by the uprated radiation conditions.

If equipment is found to be affected by uprated radiation conditions, the qualification of the equipment will be resolved by

~ performing location-specific radiation calculations or by reducing the

'ualified life of.the equipment.

As stated in Section 10.2.l of Reference I, the normal and accident temperature,

pressure, and humidity inside and outside containment are unchanged by the power uprate, and thus do not thermally affect the non-metallic components of qualified mechanical equipment.

The licensee stated in a phone conversation on July 27, 1994, that these plant conditions also apply to Section 10.2.2 of Reference l.

The NRC staff reviewed the licensee's evaluation and commitment to resolve qualification of safety-related mechanical equipment, and concludes that the licensee's approach to the qualification of safety-related mechanical equipment with non-metallic components for power uprate is acceptable.

3.8.3 Balance-of-Plant Piping Systems The licensee evaluated the balance-of-plant (BOP) piping systems by comparing the original design-basis conditions with those for the proposed uprated conditions.

The licensee also performed stress analyses in accordance with require'ments of the code and the code addenda of record under the power uprate conditions.

For the limiting BOP piping systems, ratios of the maximum calculated stresses at the uprated power conditions to the corresponding code-allowable stresses are summarized in Tables 2 and 3 of Reference 6.

The results show that the calculated piping stresses are within the code-allowable limits.

The licensee did not identify any new postulated pipe break locations in the systems evaluated.

The licensee evaluated the supports of the BOP piping systems by reviewing the increase in the pipe support loadings and stresses due to increase in the

pressure, temperature, and flow rate in the affected piping systems.

The licensee determined that there is sufficient margin between the original design stresses and the code limits of pipe supports to accomnodate the stress increase due to the power upr ate.

The NRC staff reviewed the results of the licensee's evaluation and concludes that the BOP systems will operate at the proposed power uprate conditions without adverse effects on the piping system and pipe supports.

3.8.4 Startup Testing Program The licensee committed to a startup test program as described in Reference 9.

The startup test program includes system testing of such process control systems as the feedwater flow control system.

The licensee will collect steady-state operational data from 90-percent power up to the previously rated

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thermal power so that predicted equipment performance characteristics can be verified.

The licensee will conduct the startup testing program in accordance with its procedures.

The licensee also comaitted to include acceptance testing of the RCIC system in the startup test program.

The staff finds that the licensee's approach conforms to the test guidelines in Reference 9,

and is acceptable.

The NRC staff considers a testing program on the reactor recirculation (RRC) system necessary during startup subsequent to power uprate to demonstrate flow control over the entire flow range to enable a complete calibration of the flow control instrumentation, including signals to the process computer.

As stated in Reference 9, these tests should also ensure that no undue vibration occurs at uprate or ELLL conditions.

In response to a request for additional information dated October 26, 1993 (Reference 24), the licensee committed, in a letter dated January 6,

1994 (Reference 3), to do startup testing afte}

making the power uprate modifications, and prior to operating in the ELLL region.

The NRC staff finds this commitment conforms to the test guidelines in Reference 9,

and is acceptable.

3.8.5 Equipment Seismic and Dynamic gualification The licensee evaluated plant equipment for seismic and dynamic qualification.

The licensee found that (a)

Seismic loads are unchanged by power uprate.

(b)

The original LOCA dynamic loads (pool swell, condensation oscillation (CO) and chugging, annulus pressurization, and jet impingement) are not significantly affected by the power uprate conditions as discussed in Section 4.1.2 of Reference 10.

(c)

No new pipe break locations resulted from the uprated conditions.

(d)

The increased SRV loads due to the increase of SRV opening setpoint pressure and the setpoint tolerance are within the margins in the plant load definition for these loads.

(e)

The increased temperature,

pressure, and flow conditions for power uprate on various safety-related mechanical component internals do not exceed the equipment design criteria.

The NRC staff concludes that the above information confirms that the original seismic and dynamic qualification of the safety-related mechanical and electrical equipment is not affected by the power uprate conditions and is, therefore, acceptable.

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3.9 u t on o

ffect of Power U rate on Res onses to Generic o

ication In Reference 7,

GE submitted an assessment of the effect of power uprate on licensee responses to generic NRC and industry communications.

GE reviewed both NRC and industry communications to determine if parameter changes associated with power uprate could affect previously made licensee commitments or earlier responses.

A large number of documents were reviewed (more than 3000 items);

GE ndted that only a small number of these would be affected by power uprate.

The list of affected topics was then divided into those that could be bounded generically by GE, and those that would require plant-specific reevaluation.

The NRC staff audited the GE assessment in December 1991 and approved the assessment in Reference 23.

In addition to assessing those items requiring a plant-specific reevaluation, the licensee is also reviewing the potential effects of power uprate on internal commitments and procedures.

The licensee determined that the plant-specific issues were either acceptable for power upr ate, or have been revised to reflect the uprated conditions.

The licensee committed to resolve any changes to commitments to include the uprated conditions.

The NRC staff may audit these activities after plant startup following implementation of power uprate modifications.

The NRC staff finds this approach acceptable.

3.10 on a

es Re uest d b t S

1 S ste 3.10. 1 Safety/Relief Valve (SRV) Setpoint Tolerance and Out-of-Service Analysis The SRVs provide three main protection functions: (I) overpressure relief operation (power relief mode), in which the valves open automatically to limit a reactor vessel pressure rise; (2), overpressure safety operation (spring safety mode) to prevent reactor vessel overpressurization; and (3) depressurization operation; the automatic depressurization system (ADS) valves open automatically as part of the ECCS, for events involving small breaks in, the reactor pressure boundary.

GE submitted a generic topical report (Reference

28) for NRC review that supports changes to the current requirements for SRV setpoint tolerance changes.

The staff prepared a safety evaluation report (SER) that approved relaxing the setpoint tolerance limit from klX to i3X, when plant-specific analyses are submitted to support these changes (Reference 27).

The NRC SER specified that each licensee implementing these changes must prepare certain plant-specific analyses to include the following:

(a)

Transient analysis of all abnormal operating occurrences (AOOs) listed in Reference 28 using the i3-percent setpoint tolerance for the SRVs and using staff-approved methodology.

(b)

Analysis of the design overpressure event using the k3-percent tolerance limit to confirm that the vessel peak pressure does not exceed the ASHE Boiler and Pressure Vessel Code upset limit.

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(c)

The plant-specific analyses in Items (a) and (b) above should ensure that the number of SRVs analyzed corresponds to the number required to be operable in the TSs.

(d)

Reevaluation of the performance of high-pressure systems (pump capacity, discharge

pressure, etc.) motor-operated

valves, and vessel instrumentation and associated piping must be completed, considering the i3-percent limit.

(e)

Evaluation of the f3-percent tolerance on any plant operating mode such as increased core flow, extended operating domain (ELLL), and power uprate must be completed.

(f)

Evaluation of the effect of the %3X-percent tolerance limit on the containment response during LOCAs and the hydrodynamic loads on the SRV discharge lines must be completed.

The licensee submitted an evaluation (Reference

12) to support an increase in SRV setpoint tolerance and number of SRVs allowed out of service (OOS) for WNP-2.

The report was in support of modifying the current in-service opening pressure setpoint tolerance from +1X/-3X to %3X, and allowance for up to two ADS valves 00S.

The licensee performed its analysis assuming a thermal power level of 3629 HWt, corresponding to 104.1 percent of the uprated power level of 3486 HWt.

The analysis addresses a core flow operating range from 108.5 Hlbm/hr to 115 Hlbm/hr at the thermal power of 3486 HWt and an operating pressure of 1035 psia which corresponds to the WNP-2 power uprate with ELLL operation.

A setpoint tolerance of +3 percent above the nominal SRV setting was assumed in the analysis.

The SRV setpoint tolerance and OOS analysis integrates. the analysis supporting overpressure protection, containment

response, SRV load definition, emergency core cooling system (ECCS) loss-of-coolant accident (LOCA) analysis

events, abnormal operating occurrences (AOOs), anticipated transients without scram (ATWS), high-pressure core spray/reactor core isolation cooling (HPCS/RCIC) performance, standby liquid control system (SLCS) performance, common mode failure concerns, and Appendix R events.

The licensee's

analysis, assuming HSIV closure with indirect flux scram, with six SRVs

OOS, and an initial operating pressure of 1050 psia, resulted in a peak reactor vessel bottom head pressure of 1335 psig.

This complies with the ASHE Code allowable value of 1375 psig, and is acceptable.

The most limiting thermal transient is the load rejection with bypass failure event coincident with end-of-cycle (EOC) recirculation pump trip (RPT)

OOS.

The licensee analyzed this event to determine the operating limit HCPR to ensure safe plant operation.

The licensee.found that the peak HCPR occurs before the SRVs open at the setpoint tolerance, and the change, therefore, has no impact on thermal limits.

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h 41 The ECCS-LOCA analysis assumed that the SRVs would open at the +3 percent above the nominal setpoints with four SRVs OOS.

The break spectrum results show that, for the large breaks, the reactor vessel depressurizes through the break, while smaller breaks require the ADS for depressurization.

The small-break analysis assumed two ADS valves OOS.

The results shown in Table 4-2 of Reference 7 show that the peak clad temperature (PCT) for small breaks conforms to the 10 CFR 50.46 limit with various single failures assumed.

The HPCS and RCIC'performance were evaluated for loss-of-feedwater (LOFW) events.

The HPCS was evaluated assuming (I) no RCIC is available, (2) low-pressure ECCS pumps are available, (3) six SRVs OOS with relief mode 30 psi above the new nominal setpoints, (4)

SRVs close at 50 psi below the opening

pressure, and (5) the reactor water level initially at Level 3.

The results show that the water level remains above the active fuel, which is acceptable.

The RCIC was evaluated with the same assumptions as the HPCS analysis, except the HPCS and low-pressure ECCS pumps were assumed OOS.

The licensee's results show no core uncovery, which is acceptable.

Even though six SRVs were assumed OOS, there are other limiting concerns that will allow fewer OOS SRVs.

These limiting concerns will govern the number of SRVs that will be allowed OOS.

The licensee evaluated the SLCS performance.

The operating pressure range for SLCS is based on the lowest setpoint SRV available in the relief mode.

The ability of the SLCS pumps to inject their design flow at higher pressures is not affected because these pumps are positive displacement-type pumps and are designed to provide constant flow regardless of system pressure.

The licensee stated that the electric motors to drive these pumps have sufficient horsepower margin to meet the pump power requirements.

The effect of SRVs OOS was also evaluated for the limiting ATWS event.

The reactor is eventually shut down by. the SLCS during an ATWS.

Initially the power is reduced by the recirculation pump trip (RPT) signal.

The vessel experiences maximum pressure during the initial portion of the event.

After the RPT and subsequent actuation of the SRVs, the increasing pressure transient is terminated.

With four SRVs OOS, the peak pressure was calculated to be 1467 psig, which is below the ASIDE Code Service Level C value of 1500 psig.

The licensee found that four SRVs 00S does not violate the overpressurization criterion for an ATWS event.

The licensee evaluated the common mode failure aspects of the number of SRVs allowed to be OOS.

The condition with a large number of SRVs inoperable may be indicative of a common mode failure mechanism.

This is the limiting concern that determines how many SRVs will be allowed to be OOS.

The licensee also evaluated the increased SRV setpoint tolerance of k3 percent.

In order to ensure that the SRV setpoints do not drift beyond the allowable i3-percent tolerance

range, the staff 'requested in Reference 29 that the licensee place a restriction in the proposed technical specification (TS) for the SRVs that the valves be reset to %1-percent before returning them to service.

Without this restriction, the NRC staff expected that any valve found to have significantly drifted, yet not outside the k3-percent tolerance limit, would continue to drift, possibly outside the X3-percent range.

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Resetting the valves would provide assurance that the SRVs remain operable to accomplish their analyzed safety functions.

However, the licensee stated in Reference 6 that the imposition of this requirement -would result -in a significant hardship.

Specifically, the licensee stated that this requirement would cause multiple startups and shutdowns after refueling outages since the SRVs are setpoint tested and postmaintenance tested on-line with normal system

. pressure at low reactor power or decay heat.

The plant has Crosby spring-type SRVs which the licensee stated have not experienced significant setpoint drift.

The staff'generally agrees that the setpoint drift for the plant-specific model SRVs has not been as severe as for some other model SRVs used in the industry.

The licensee further stated that the plant-specific analyses for the power uprate condition conservatively assume that only 12 of the 18 plant SRVs are operable for performing the SRV safety function.

This additional margin provided in the plant-specific analyses gives reasonable assurance that the SRV setpoint drift would not result in the maximum allowable system pressure being exceeded.

Therefore, the staff agrees that the licensee's proposal to not incorporate the resetting of the SRVs to fl percent in the plant TS is acceptable.

The NRC staff reviewed the results of the licensee's

analyses, and concludes, that a setpoint tolerance increase from +1X/-3X to %3X, and allowing up to tvo SRVs to be 00S was properly analyzed by the licensee and is acceptable.

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'Jl TABLE 3.1 Containment Performance Results Current FghR Updated Methods QS~LCaso Updated Methods at Uprated Final it Uprated Power With

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Peak Drywall Pressure (psig) 34.7 348 34.8 3S.1 4S Peak Wctwcll Pressure gsig) 273 30.2 30.2 (4) 4S Peak Dryw~Wctwcll Pressure Differenc(paid)

IP.4 21.7 213 (4)

(4) 212 21S 210 212 (I) hnalysis performed at 3462 MWt (104.2% oforiginal rated power), domo pressure 1040 psig.

(2) hnalysis porformcd at 3702 MWt (102 x 110% oforiginal rated stcam Qow), dome pressure 1040 psig.

(3)

Update (curtent) methods.

(4)

Bamdcd by power uprato casa..

Note:

Tho value of P, to bo used for 10 CFR Part SO hppcadix J testing must bo more thaa or equal to tho peak calculated coatainment pressure resulting &om any dcsiga4asis accidctx. To bouad a potential futuro power upratc (110% oforiginal rated stcam Qow) and possiblo unforeseen lbture plant changes, tho valuo ofP, was conscrvativcly chosen by tho liccnsce to bo 38 psig.

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Table 3.2 LOCA Radiological Consequences DOSE (rem)

GUIDELINES (rem)

Wholo Body Doso '.7

'Ihyroid Dose 86.3 Low Population Zone:

Wholo Body Dose Thyroid Doso Control Room:

Whole Body Dose Thyroid Dose Beta Dose 4.1 95.8 0.4 15.4 3.5 5

30 30 Based on 1.02 X uprated power This valuo is a limitderived from General Design Criterion 19.

FHA Radiological Consequences LOCATION Exclusion Area:

DOSE (rem)

GUIDELINES (rem)

Whole Body Dose Thyroid Dose Low Population Zone:

Whole Body Doso Thyroid Doso 1.1 1.5 0.4 0.6 6

75 6

75 LOCATION Exclusion Area:

CRDA Radiological Consequences DOSE (rem)

GUIDELINES (rem)

Whole Body Dose Thyroid Dose Low Population Zone:

Whole Body Dose Thyroid Dose 0.03 0.3 0.02 0.7 6

75 6

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I.C C~CCNTC T

In accordance with the Commission's regulations, the Washington State official was notified of the proposed issuance of the amendment.

The State official had no coments.

5.0 NMENT L CO S

T 0 Pursuant to 10 CFR 51.21, 51.32, and 51.35, an environmental assessment and finding of no significant impact was published in the @de~a f)ece3s~t on Hay 2, 1995 (60 FR 21554).

In this finding, the Commission determined that issuance of this amendment would not have a significant effect on the quality of the human environment.

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The Comaission has concluded, based on the considerations discussed

above, that (1) there is reasonable assurance that the health and safety of the public will not be endangered by operation in the proposed

manner, (2) such activities will be conducted in compliance with the Commission's regulations, and (3) the issuance of the amendment will not be inimical to the common defense and security or to the health and safety of the public.

7. 0 ~RACES (1)

Letter to the NRC from J.

V. Parrish,

MPPSS, G02-93-180, requesting an amendment to the WNP-2 operating license and TS to increase licensed power level with extended load line limit, and a change in safety relief valve setpoint tolerance, dated July 9, 1993.

(2)

Letter to the NRC from J.

V. Parrish,

MPPSS, G02-94-004, requesting an administrative change to the WNP-2 TS to support the power uprate, January 6,

1994.

(3)

Letter to the NRC from Request for Additional 1994.

(4)

Letter to the NRC from RequesC'or Additional 1994.

J.

V. Parrish,

WPPSS, G02-94-006, "Response to Information, Power Uprate Review," January 6,

J.

V. Parrish,

WPPSS, G02-94-029, "Response to Information, Power Uprate Review," February 2, (5)

Letter to the NRC from Request for Additional (6)

Letter to the NRC from Request for Additional J.

V. Parrish,

WPPSS, G02-94-103, "Response to Information, Power Uprate Review," May 3, 1994.

J.

V. Parrish,

MPPSS, G02-94-115, "Response to Information, Power Uprate Review,"

May 13, 1994.

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(7)

(8)

(9)

GE Nuclear Energy, "Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NEDC-31984P, Class III (Proprietary), July 1991; NED0-31984, Class I (Non-Proprietary),

Harch 1992; and Supplements 1 and 2.

Letter from M. T. Russell, NRC, to P.

W. Marriott, General Electric Company, "Staff Position Concerning General Electric Boiling Water Reactor Power Uprate Program,"

September 30, 1991.

GE Nuclear Energy, "Generic Guidelines for General Electric Boiling Mater Reactor Power Uprate," Licensing Topical Report NED0-31897, Class I (Non-Proprietary),

February 1992; and NEDC 31897P-A, Class III (Proprietary)

June 1991.

(10)

GE Nuclear Energy, "Power Uprate Mith Extended Load'Line Limit Safety Analysis for MNP-2," Licensing Topical Report NEDC-32141P, Class III'*

(Proprietary),

June 1993.

(11)

GE Nuclear Energy, "Mashington Public Power Supply System Nuclear Project 2 SAFER/GESTR-LOCA Loss-of-Coolant Analysis," Licensing Topical Report '

NEDC-32115P, Class III (Proprietary),

Revision 2, July 1993.

(12)

GE Nuclear Energy, "Mashington Public Power Supply System Nuclear Project 2 SRV Setpoint Tolerance and Out-of-Service Analysis," Licensing Topical Report GE-NE-187-24-0992, Revision 2, Class III (Proprietary), July 1993.

(13) Letter to the NRC from J.

V. Parrish,

WPPSS, G02-94-221, requesting a

modification to the powe} /flow maps to reflect use of the Siemens Power Corporation STAIF code, September 26, 1994.-

4 (14) Letter from H. J. Virgilio, NRC, to R. A. Copeland, Siemens Power Corporation, "Acceptance for Referencing of Siemens Power Corporation Topical Report EHF-CC-074(P):

Volume 1,

'STAIF: A Computer Programs for BWR Stability Analysis in the Frequency Domain,'nd Volume 2,

'STAIF: A Computer Program for Stability in the Frequency Domain - Code gualification Report.'"

(15)

NRC Bulletin 88-07, Supplement 1,

"Power Oscillations in Boiling Mater Reactors,"

December 30, 1988.

(16)

ASHE Boiler and Pressure Vessel

Code,Section III, Division 1, 1971 Edition with Summer 1971 Addenda.

(17)

ASIDE Boiler and Pressure Vessel

Code,Section III, Division 1, 1968 Edition with Addenda through Winter 1970.

(18)

ASIDE Boiler and Pressure Vessel Code,Section III Subsection NB-3600, 1971 Edition with Addenda through Minter 1971.

(19)

GE Nuclear Energy Service Information Letter, SIL-377.

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4 H

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C[

(20) U.S. Nuclear Regulatory Commission, Regulatory Guide 1. 139, "Guidance for Residual Heat Removal,"

May 1978.

(21) Letter from A. Thadani, NRC, to G. L. Sozzi, General Electric Corporation, "Use of SHEX Computer Program and ANSI/ANS 5.1-1979 Decay Heat Source Term for Containment Long-Term Pressure and Temperature Analysis," July 13, 1993.

(22) U.S. Nuclear Regulatory Comnission, "Suppression Pool Temperature Limits for BWR Containments,"

NURG-0783, September 1991.

(23) Letter from W. T. Russell, NRC, to P.

W. Marriott, General Electric Company, "Staff Safety Evaluation of General Electric Boiling Water Reactor Power Uprate Generic Analyses," July 31, 1992.

(24) Letter from J.

W. Clifford, NRC, to J.

V. Parrish,

WPPSS, "Request for Additional Information, Power Uprate Review," October 26, 1993.

(25)

GE Nuclear Energy, "General Electric Instrument Setpoint Methodology,"

NEDC-31336P (Proprietary Report),

October 1986.

(26) Letter from J.

W. Clifford, NRC, to J.

V. Parrish,

WPPSS, "Request for Additional Information, Power Uprate Review," January 26, 1994.

(27)

NRC Safety Evaluation Report of Licensing Topical Report NEDC-31753P, March 8, 1993.

(28)

GE Nuclear Energy, "BWROG In-Service Pressure Relief TS Revision Licensing Topical Report,"

NEDC-31753P (Proprietary Report),

February 1990.

(29) Letter from J.

W. Clifford, NRC, to J.

V. Parrish,

WPPSS, "Request for Additional Information, Power Uprate Review," December 29, 1993.

(30) Letter from J.

W. Clifford, NRC, to J.

V. Parrish,

WPPSS, "Request for Additional Information, Power Uprate Review," November 9, 1993.

Principal Contributors:

R. Frahm, Sr.

C.

Wu G.

Hammer S. Klementowicz W. Long C. Gratton H. Garg J. Clifford Date:

Nay 2, 1995

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