ML20132C369

Safety Evaluation Supporting Amend 239 to License DPR-59

ML20132C369
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Site:	FitzPatrick
Issue date:	12/06/1996
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO AMENDMENT >H). 239 TO FACILITY OPERATING LICENSE NO. DPR-59 POWER AW HORITY OF THE STATE OF NEW YORK JAMES A. FITZPATRICK NUCLEAR POWER PLANT DOCKET NO. 50-333

1.0 INTRODUCTION

By "JPN-92-028, Forwards Application for Amend to License DPR-59,increasing Authorized Max Power Level by Approx 4.1% to 2536 Mwt from Current Limit of 2436 Mwt.Proprietary TRs NEDC-32016P & [[Report" contains a listed "[" character as part of the property label and has therefore been classified as invalid.,encl.Without Proprietary TRs|letter dated June 12,1992]] (Reference 1), as supplemented September 17, 1992, March 17, 1993, August 17, 1993, August 18, 1993, December 29, 1993, June 23, 1995, August 15, 1996, October 3, 1996, October 23, 1996, November 14, 1996, November 20, 1996 (JPN-96-045), November 20, 1996 (JPN 046), and November 27, 1996, the New York Power Authority (ilYPA or the licensee) submitted an application for Operating License DPR-59 and a request for changes to the James A. FitzPatrick Nuclear Power Plant (JAFNPP) Technical Specifications (TSs). The requested changes would increase the licensed thermal power level of the JAFNPP from the current limit of 2436 MWt to 2536 MWt. The request would also approve changes to the TSs to implement uprated power operation. This request is in accordance with the generic boiling water reactor (BWR) power uprate program established by the General l

Electric Company (GE) and approved by the U.S. Nuclear Regulatory Commission (NRC) staff in a letter dated September 30, 1991 (Reference 2). The September 17, 1992, March 17, 1993, August 17, 1993, August 18, 1993, December 29, 1993, June 29, 1995, August 15, 1996, October 3, 1996, October 23, 1996, and November 26, 1996, letters provided clarifying l

information that did not change the initial submittal proposed no significant hazards consideration determination.

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2.0 BACKGROUND

l On December 28, 1990, GE submitted GE Licensing Topical Report (LTR)

NEDC-31897P-1, in which it proposed 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 (Reference 3). The report contained a proposed outline for individual license amendment submittals and discussed the scope and depth of reviews needed and the methodologies used in these reviews.

In the letter dated September 30, 1991 (Reference 2), the NRC approved the program proposed in the report, on the condition that the individual power uprate amendment request meets certain requirements in the document.

l 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

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was based on the vendor-guaranteed poWr 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

.l determining the specifications for all major NSSS equipment, including the emergency core cooling systems (ECCS). Therefore, increasing the rated l

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

The licensee's amendment request to increase the current licensed power level of 2436 MWt to a new limit of 2536 MWt represents an approximate 4.1 percent increase in thermal power with a 4.8 percent increase in rated steam flow (an increase in vessel steam flow from 10.47 to 10.976 M1b/hr). JAFNPP will increase power to the higher level by:

(1) an increase in the core thermal power with a more uniform (flattened) power distribution to create increased steam flow, (2) a corresponding increase in the feedwater system flow, (3) no increase in maximum core flow, and (4) reactor operation primarily along extensions of current control rod position to core flow control lines.

This approach is consistent with the NRC-approved BWR generic power uprate guidelines presented in NEDC-31897P-A (Reference 3). The increased core power will be achieved by utilizing slightly flatter radial power distribution while maintaining the most limiting fuel bundles within their operating constraints.

The operating pressure of the reactor will be increased approximately 25 psi i

to assure satisfactory turbine pressure control and pressure drop l

characteristics with the increased steam flow.

3.0 EVALUATION i

The NRC staff reviewed JAFNPP's request for a power uprate amendment using applicable rules, regulatory guides, sections of the Standard Review Plan (NUREG-0800), and NRC staff positions. The NRC staff also evaluated JAFNPP's submittal (Reference 1) for compliance with the generic BWR power uprate program as defined in Reference 3.

Detailed discussions of individual review 4

j topics follow.

3.1 Reactor Core and Fuel Performance 3.1.1 Fuel Design and Operation All fuel and core design limits will be met by control rod pattern adjustments. New fuel designs are not needed for power uprate.

Power uprate will increase the core power density, and will have some effects on operating flexibility, reactivity characteristics, and energy requirements.

These 4

issues are discussed in the following sections.

3.1.2 Thermal Limits Assessment The operating limit minimum critical power ratio (MCPR) is determined on a cycle-specific basis from the results of reload analysis, as described in General Electric Report NEDC-31984P, " Generic Evaluations of General Electric Boiling Water Reactors Power Uprate," July 1991; and Supplements 1 and 2 (Reference 4). The maximum average planar linear heat generation rate (MAPLHGR) and linear heat generation rate (LHGR) limits will also be maintained as described in this reference. The plant-specific safety evaluation for JAFNPP is contained in References 5 and 6.

3.1.3 Power / Flow Operating Map Power uprate is achieved by expanding the upper portion of the operating map (power / core flow) along the current rod / flow control lines which have not changed, but have been renamed to reflect the redefinition of rated thermal power.

Full power operation under the Extended Load Line Limit Analysis (ELLLA), which was previously achieved at a minimum value of approximately 87 percent of maximum core flow, will now be achieved at approximately 93 percent of maximum core flow along the same rod / flow control lines.

The absolute power MWt at that point on the operating map will be higher since the rated thermal power limit is redefined.

3.1.4 Stability The BWR Owners' Group (BWROG) and the NRC are currently addressing methods to minimize the occurrence and potential effects of power oscillations that have occasionally been observed for certain BWR operating conditions.

Until a more permanent solution is developed, NYPA has adopted the generic interim operating constraints proposed by GE.

Existing plant procedures have incorporated operating constraints, in accordance with NRC Bulletin 88-07 and-Supplement I to that bulletin, which restricts plant operation in the high power / low core flow region of the power / flow operating map. The restricted i

operating regions of the power / flow map for power uprate have not changed, and operator actions upon entry into these regions will likewise remain the same.

This is consistent with the information presented in the generic evaluations provided by GE in Reference 4.

3.1.5 Reactivity Control 3.1.5.1 Control Rod Drives (CRDs) and Scram Performance The CRD system performance was evaluated at the uprated steam flow and system pressure. This included evaluation of the control rod insertion and withdrawal functions, as well as CRD cooling.

Results of the evaluation indicate that the increase in reactor pressure (approximately 35 psi) has little effect on scram insertion speed, and that the system will continue to perform all of its functions at uprated conditions. The licensee will continue to monitor, through TS surveillance requirements, the scram time performance in order to ensure that the original licensing bases for the CRD system are maintained.

This approach is consistent with that proposed by GE in Reference 4.

_ The FitzPatrick power uprate conditions with the increase of reactor dome pressure, temperature and steam flow rates are within the range of values specified in GE generic guidelines for the BWR/4 power uprate (Reference 3).

The CR0 system was evaluated at the normal maximum reactor dome pressure of 1055 psig which is higher than the nominal power uprate operating pressure of 1040 psig for FitzPatrick. Based on the review of the FitzPatrick power uprate submittal and the GE generic gui s lines, the staff concludes that the i

CRD mechanisms will continue to meet its design basis and the CRD system will continue to perform its safety function at uprated power.

3.1.5.2 Standby Liquid Control System (SLCS)

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 enrichment. The licensee may wish 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 the reactivity requirements of the SLCS.

The SLCS requirements are evaluated at each reload. The SLCS is designed to inject at a maximum pressure equal to that of the lowest safety / relief valve (SRV) setpoint. The SLCS pumps are positive displacement pumps, where a small pressure increase in the SRV setpoint will not impair the performance of the pumps. Therefore, the staff concludes that the ability of the SLCS to inject to the reactor will not be impaired by the power uprate.

3.2 Reactor Coolant System and Connected Systems The staff's review of the mechanical engineering portions of the FitzPatrick power uprate amendment request centered on the effects of power uprate on the structural and pressure boundary integrity of the piping systems and components, their supports, and reactor vessel and internal components.

3.2.1 Nuclear System Pressure Relief The purpose of the nuclear pressure relief system is to prevent overpressurization of the NSSS during abnormal operational transients. The main steam line safety / relief valves (SRVs) provide this protection.

In Reference 3, GE evaluated the impact of uprated conditions; namely, increased temperatures, pressures, and flow rates on the SRVs. GE determined that the function and structural integrity of the SRVs would not be compromised by power uprate. The only change in the nuclear pressure relief system for power uprate is an increase in SRV setpoints to accommodate an approximate 35 psi increase in reactor dome pressure. These setpoints will be increased to maintain an adequate simer margin during reactor operation.

.- 3.2.2 Motor-0perated Valve (MOV) Capability at Increased Line and Valve i

Differential Pressures NYPA evaluated the effect of power uprate Line Pressures (LPs), Differential Pressures (DPs), and post-accident maximum ambient temperatures on Motor Operated Valves (MOVs) within the scope of GL 89-10, " Motor-Operated Valve Testing and Surveillance." As a result of these evaluations, no M0V modifications or field adjustments (i.e., torque or limit switch adjustments) will be required to support power uprate.

The increased thrust required to operate the MOVs due to expected increased line and valve differential pressures is within the capabilities of the existing valve actuators. The staff finds the results of this analysis acceptable.

3.2.3 Reactor Overpressure Protection The design pressure of the FitzPatrick reactor vessel and reactor pressure coolant boundary will remain at 1250 psig after power uprate.

The American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code) allowable pressure limit for pressurization events is 1375 psig. The licensee has analyzed the limiting pressurization event, which is a main steam isolation valve (MSIV) closure with failure of the reactor to automatically scram on MSIV position.

The results of the analysis are documented in GE Topical Report NEDC-32016P, " Power Uprate Safety Analysis for the James A.

FitzPatrick Nuclear Power Plant." Two SRVs were assumed to be out-of-service, and an initial operating pressure of 1055 psig was used in the analysis. The analysis also assumed 102 percent of 2536 MWt,100 percent core flow, and an automatic scram on high neutron flux during the event. At the uprated conditions, a peak pressure of 1326 psig resulted, but remained below the 1375 psig ASME Code allowable limit. The staff concludes that overpressurization protection will remain adequate after power uprate.

3.2.4 Reactor Recirculation System l

The reactor recirculation flow control system cavitation interlocks will l

remain the same.

These interlocks are based on subcooling in the external recirculation loop, and therefore, are a function of thermal power and l

feedwater flow. However, the reactor pressure vessel dome will be operated at an increased pressure which will result in below-rated power level than previously used in the design basis analysis. Slightly more subcooling will result at higher pressures for the same core power-flow point. Therefore, it would be possible to lower the cavitation interlocks but it was judged not to

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be necessary.

The licensee states that the required pump head and pump flow 1

will increase power demand to the pump motors by less than two percent, thus, they concluded that operation in the power uprate condition is within the capability of the recirculation system.

Because the maximum flow at the uprated condition is within the maximum design flow, the uprate analysis is i

acceptable.

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1,

However, the submittal does not specifically discuss the possibility of pump cavitation and pump vibrations due to the increased pump speed and increased

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pump flow which might overcompensate the increased vessel dome pressure and feedwater subcooling. Therefore, at the staff's request the licensee proposed a license condition to monitor pump motor vibration during initial power ascension to uprated power conditions (Letter from William J. Cahill, Jr. to the USNRC, " Response to Request for Additional Information Regarding Power Uprate" November 14, 1996).

The staff has reviewed the licensee's submittals and finds the licensee's approach acceptable.

3.2.5 Main Steamline Isolation Valves (MSIVs)

The performance of the MSIVs with regard to reactor coolant pressure boundary requirements such as closure time and leakage could be impacted by the increased operating pressure; however, the pressure increase is relatively small (less than 3 percent) and performance will be monitored by surveillance requirements in the plant TSs to ensure the original licensing basis for the MSIVs is preserved.

The testing will be performed in accordance with Technical Specification 4.7.D.I.c.(2) which requires, "with the reactor at a reduced power level, fast close each main steam isolation valve, one at a time, and verify closure time," test acceptance criteria require valve closure between the limits of 3 seconds and 5 seconds. This testing will be performed during power ascension at approximately 70 percent of uprate power. The staff finds this acceptable in that it will ensure acceptable performance of the MSIVs.

3.2.6 Reactor Core Isolation Cooling (RCIC)

The RCIC system provides core cooling when the reactor pressure vessel (RPV) is isolated from the main condenser, and RPV pressure is greater than the maximum all_owable for initiation of a low pressure cooling system. The licensee has assessed the RCIC system in a manner consistent with the bases and conclusions of Section 4.2 of Reference 1.

However, the FitzPatrick RCIC system has a smaller turbine than those used on some newer BWRs, and may not require the GE SIL No. 377 modifications, based on the RCIC startup transients

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observed at the plant.

A' modification of the Reactor Core Isolation Cooling system which replaces the steam admission valve addresses the concerns of General Electric SIL No. 377, RCIC Startup Transient Improvement with Steam Bypass.

Reactor Core Isolation Cooling system performance testing will be performed in accordance with Technical Specification 4.5.E.1, which requires that, "The RCIC pump shall deliver at least 400 gpm for a system head corresponding to a reactor pressure of 1195 psig to 150 psig." for the power uprate condition.

Performance and/or operability testing will be performed during the power ascension at reactor pressures of approximately 150 psig and 1000 psig and at approximately 100 percent of uprate power. The staff finds this acceptable in that it will ensure operability of the RCIC system at the uprated power.

t 3.2.7 Residual Heat Removal (RHR) System The RHR system is designed to restore and maintain the coolant inventory in the reactor vessel and to provide decay heat removal following reactor shutdown for both normal and post-accident conditions.

The RHR system is designed to operate in the low pressure coolant injection (LPCI) mode, shutdown cooling mode, suppression pool cooling mode, containment spray cooling mode, and fuel pool cooling assist mode. The LPCI mode is discussed in Section 4.2.2 of this report. The effects of power uprate on the shutdown cooling mode and suppression pool cooling modes are discussed in the following paragraphs.

3.2.7.1 Shutdown Cooling Mode The effect of power uprate on the shutdown cooling mode is that the time required to reach shutdown temperature is slightly increased due to the increase in decay heat.

Regulatory Guide 1.139, " Guidance for Residual Heat Removal," requires demonstration of cold shutdown (212 *F reactor temperature) capability within 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />. The licensee has estimated based on uprated conditions that the time to reach cold shutdown is 21 hours2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br />, even considering the availability of only one RHR loop.

Therefore, the 36-hour criterion can still be met with power uprate, 3.2.7.2 Suppression Pool Cooling and Containment Spray Mode The Suppression Pool Cooling (SPC) and Containment Spray (CC) modes are designed to provide sufficient cooling to maintain the containment and suppression pool temperatures and pressures within design basis loss-of-coolant accident (LOCA). This objective is met with power uprate, since the peak suppression pool temperature analysis by the licensee confirms that the pool temperature will stay below its design limit at uprated conditions.

Power uprate increases the containment spray temperature by a few degrees.

This has a negligible effect on the calculated values of drywell pressure, drywell temperature, and suppression pool chamber pressure, since these parameters reach peak values prior to actuation of the containment spray. The licensee stated that the capability of the CC mode is therefore acceptable for power uprate. The effect of higher suppression pool temperature on the net positive suction head (NPSH) of the RHR pumps during SPC and CC modes is also discussed in 3.9 Reference 5.

The results show that there is adequate NPSH margin for the RHR and Core Spray (CS) pumps. The effects of power uprate on the above-mentioned cooling modes of the RHR system are acceptable to the staff.

3.2.8 Reactor Water Cleanup System (RWCU)

The operating temperature and pressure of the RWCU system will increase slightly as a result of power uprate. The licensee evaluated the effect of these increases and concluded 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

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I uprate will not require a change in TS limits for reactor water chemistry.

Therefore, the power uprate will not significantly affect the operation or coolant boundary integrity of the RWCU system.

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3.3 Enaineered Safety Features (ESF) 3.3.1 Emergency Core Cooling Systems (ECCS) 1 The long-term bulk pool temperature response with power uprate was evaluated 1

for the DBA LOCA. An increase in the peak pool temperature resulted, but still remained below the design limit of 220 *F.

Calculations were performed i

to ensure available NPSH would not be significantly impacted.

Results show that the bulk pool temperature increases, which would lower the NPSH, but is offset by the increase in the torus wetwell airspace pressure which would increase the NPSH.

Based on the above, the staff finds that the power uprate will not affect UFSAR compliance with ECCS NPSH requirements.

l 3.3.1.1 High Pressure Coolant Injection (HPCI) s The HPCI system design basis is to provide reactor vessel inventory makeup during small and intermediate break loss-of-coolant accidents (LOCA) and j

reactor vessel isolation events. The HPCI system is designed to provide its rated flow over a reactor prest,ure range of 150 psig to a maximum pressure j

based on the lowest SRV safety setpoint.

The SRV opening actpoints will be increased for power uprated to maintain adequate simmer margin.

Increasing the SRV setpoint pressure has a potential impact on the maximum operating i

pressure for the HPCI system.

The effect of power uprate on HPCI system operability, including potential system modifications, was addressed by GE in Reference 4.

The required flow rate remains unchanged. However, the HPCI pump and turbine operational requirements at uprated conditions are increased.

The pump total dynamic head is increased by approximately 3 percent due to SRV setpoint increase. The speed and power requirements of the steam turbine are also increased. The licensee adopted the assessment of turbine overspeed as described in the generic topical report and has implemented GE SIL 480 for the HPCI system.

The licensee will conduct high Pressure Coolant injection system performance testing in accordance with Technical Specification 4.5.C.1, which requires that, "the HPCI pump shall deliver at least 4250 gpm against a system head corresponding to a reactor vessel pressure of 1195 psig to 150 psig" for the power uprate condition.

Performance and/or operability testing will be performed during the power ascension at reactor pressures of approximately 150 psig and 1000 psig and at approximately 100 percent of uprate power. The staff finds this acceptable.

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3.3.1.2 RHR System (Low Pressure Coolant Injection, LPCI)

The licensee has adopted the bounding generic evaluation provided in the GE l

topical report (Reference 4) for the LPCI mode of the RHR system.

This

-analysis is applicable to FitzPatrick. The licensing and design flow rates i

plus the system operating pressure will not be increased.

Therefore, the staff finds, there is no impact on the LPCI system due to power uprate.

3.3.1.3 Low Pressure Core Spray (LPCS) System 1

The licensee has adopted the bounding generic evaluation provided in the GE topical report (Reference 4) for the LPCS system. This analysis is applicable i

to FitzPatrick. The licensing and design flow rates plus the system operating pressure will not be changed. Therefore, there is no impact on the LPCS system due to power uprate.

3.3.1.4 Automatic Depressurization System (ADS)

The ADS uses 7 of 11 safety / relief valves (SRV) to reduce reactor pressure following a small break LOCA with high pressure ECCS failure.

This function 1

allows LPCI and low pressure CS to flow to the vessel. The pressure safety / relief valves open automatically upon coincident signals of reactor vessel low-low-low water level and discharge pressure indication of any low pressure cooling system (LPCI or LPCS), but only after a two minute delay.

Plant design requires a minimum flow capacity for the SRVs and that ADS l

initiate on low-low-low water level. The ADS initiating logic and ADS valve control are adequate for power uprate.

j 3.3.2 ECCS Performance Evaluation The ECCS performance under all LOCA conditions, and their analysis models, must satisfy the acceptance criteria and requirements of 10 CFR Section 50.46 and Appendix K to 10 CFR Part 50. The results of the ECCS/LOCA analysis using NRC-approved methods are presentei below.

A plant-specific analysis was performed for FitzPatrick using BP/P8x8R, GE8x8EB/NB-3, and Gell fuel types. The licensee used staff-approved SAFER /GESTR methodology to assess the ECCS capability for meeting the 10 CFR Part 50.46 criteria.

The results of the break spectra calculations show that the DBA recirculation line suction break with battery failure is the limiting case.

For the limiting fuel type (BP/P8x8R), the nominal peak cladding temperature (PCT) is calculated to be 1153 "F with corresponding Appendix K PCT of 1613 'F.

The licensing basis PCT is calculated to be 1620'F. The upper bound peak cladding temperature (UBPCT) is calculated to be less than or equal to 1510 'F.

The l

licensing basis PCT is less than 2200 'F and the UBPCT is less than the licensing basis PCT, therefore, the requirements of Appendix K are satisfied.

4

4 The licensee did not reevaluate ECCS performance for single loop operation (SLO) under power uprated conditions; however, the results of the 1980 FitzPatrick SLO analysis concluded that a MAPLHGR multiplier of 0.84 was appropriate for 8x8R and P8x8R fuel. This conclusion is applicable for the SAFER /GESTR analysis basis for all GE8x8 and Gell fuel designs because the SAFER /GESTR model results in more efficient heat removal during the boiling transition period that the previous evaluation model used to derive the multiplier for SLO.

The Extended Load Line Limit Analysis (ELLLA) provided an expanded operating rod line of the power / flow operating domain. The low flow effects on the ECCS analyses results were addressed generically in Reference 8 and received NRC approval (Reference 9).

The concern at low flow is the potential for hot fuel rod early dryout, which might result in PCT outside the licensing bases.

It was concluded that, for FitzPatrick, early dryout will not occur; this 4

conclusion is also valid for the SAFER /GESTR analyses. As an added conservatism, early dryout was assumed, yet the PCT remained within the Appendix K licensing basis.

Based on its review, the staff concludes that the ECCS performance satisfies the requirements of 10 CFR Part 50.46 and 10 CFR Part 50, Appendix K. The staff finds that the power uprate will not affect the ECCS performance.

4 3.4 Reactor Safety Performance Features 3.4.1 Reactor Transients Reload licensing analyses evaluate the limiting plant transients.

1 Disturbances of the plant caused by a malfunction, a single failure of equipment, or personnel error are investigated according to the type of initiating event. The licensee will use its NRC-approved licensing analysis methodology to calculate the effects of the limiting reactor transients as identified in the generic guidelines. The relatively small changes in rated power are not expected to affect the selection of limiting events. The events explicitly evaluated for the power uprate analysis are:

- Turbine Trip with Bypass Failure

- Generator Load Rejection with Bypass Failure

- Feedwater Controller Failure (Max. Demand)

- Inadvertent HPCI Activation

- Loss of Feedwater Heating

- Rod Withdrawal Error

- Slow Recirculation Increase

- Loss of Feedwater Flow (bounded by the generic evaluation in section 3.1 of Reference 4)

The limiting UFSAR transients were reevaluated by the licensee using the GEMINI transient analysis methods with uprated power input parameters.

The transients were analyzed at the uprated power and maximum allowed core flow point on the power / flow map for uprated operational conditions.

The current safety limit minimum critical power ratio (SLMCPR) was shown to be applicable 4

' for uprated conditions and then used to calculate the minimum critical power ratio (MCPR) operating limits. The limiting transient, Generator Load l

Rejection with Bypass Failure, yielded the greatest change in critical power ratio (CPR). This ACPR was added to the SLMCPR to determine a conservative operating limit minimum critical power ratio (OLMCPR). The NRC staff finds that use of NRC-approved methodology as described in Reference 5 will ensure l

that the effects of transients will be within applicable design and safety l

limits.

l 3.4.2 Special Events l

3.4.2.1 Anticipated Transients Without Scram (ATWS) l FitzPatrick. performed a plant-specific ATWS analysis for power uprate (Reference 13).

This study is based on the 1979 generic BWR/4 ATWS analysis performed by GE and documented in NEDE-24222, " Assessment of BWR Mitigation of I

ATWS, Volume II."

Specific modifications were made to the analyses to reflect FitzPatrick plant-specific features in addition to the uprate input parameters.

FitzPatrick Emergency Operating Procedures (EOPs) were used in l

assessing the time to initiate the SLC system and to trip FW pumps, to manually control RPV level, and to start the RHR system in suppression pool cooling mode.

Five ATWS events were analyzed: (1) MSIV closure with all SRV's in service, (2) MSIV closure with two SRVs out-of-service, (3) Turbine Trip, (4)

Inadvertent Opening of a Relief Valve (IORV), and (5) Loss of Feedwater.

For all the events analyzed, peak RPV pressures were below the service level C pressure of 1500 psig and suppression pool temperatures were below the NEDE-2422 criterion of 190 'F.

The limiting peak RPV pressure was 1495 psig for the case of MSIV closure with two SRVs out-of-service. The limiting peak suppression pool temperature was 173*F (assuming a service water temperature i

of 75 'F) for the case of an inadvertent opening of a relief valve. These results are based on a typical BWR single lope, Mark I containment plant (NUREG-0460) which are comparable to FitzPatrick calculated values, therefore they are acceptable to the staff.

3.4.2.2 Station Blackout The licensee indicated that the plant response and coping capabilities for a station blackout (SBO) event are Impacted slightly by operation at the uprated power level due to the increase in the operating temperature of the primary i

coolant system, decay heat, and main steam safety / relief valve setpoints. The licensee analyzed the impact of these increases on the condensate water requirement and the temperature heat-up in the areas which contain equipment necessary to mitigate the SB0 event and the IE batteries. The licensee concluded that no changes to the required coping period or to the systems and equipment used to respond to an SB0 event are required.

Evaluation of emergency diesel generator and Class IE battery capacities following loss of power was also determined by the licensee to be sufficient to maintain safe shutdown for uprated conditions.

• Based on its' review, the staff finds that the impact to an SB0 event due to the operation at uprated power will be insignificant and that no changes to the required coping time and to systems and equipment used to respond to an SB0 event are required.

3.4.2.3 High Energy Line Breaks The slight increase in the operating pressure and temperature caused by the power uprate results in a small increase in the mass and energy release rates following high-energy line breaks (HELB). This results in a small increase in the subcompartment pressure and temperature profiles and a negligible change in the humidity profile. The licensee reviewed the HELB for the subject piping systems (main steam, feedwater, high pressure coolant injection, reactor core isolation cooling, reactor water cleanup) and as a result concluded that the existing HELB temperature and pressure analyses envelope uprated conditions.

Evaluation of the subject piping systems also concluded that there is no change in postulated break locations due to uprated conditions.

Based on its review of the information provided, the staff concludes that the analyses for high-energy line breaks remain bounding and is acceptable for power uprate.

3.5 Containment System Performance The FitzPatrick containment responses to various postulated accidents that constitute the design basis for the containment are provided in the updated final safety analysis report (UFSAR). Operation with uprated power changes some of the conditions for the containment evaluations.

Primary containment temperature and pressure response following a postulated LOCA is important when determining the potential for offsite release of radioactive material, in determining ECCS pump net positive suction head (NPSH) requirements, and in determining environmental qualification requirements for safety-related equipment located inside the primary containment. The containment pressure and temperature response and its effects have been reanalyzed by the licensee to demonstrate the plant's capability to operate with uprated power in the following sections.

3.5.1 Containment Pressure and Temperature Response Short-term and long-term containment analyses of containment pressure and temperature response following a large break inside the drywell are reported in the UFSAR.

The short-term analysis is directed primarily at determining the peak drywell pressure response during the initial blowdown of the reactor vessel inventory to the containment following a DBA LOCA.

The long-term analysis is directed primarily at determining the peak pool temperature response. The licensee indicated that the analyses were performed at uprated power to yield the limiting containment pressure and temperature response in accordance with Regulatory Guide 1.49 and NEDC-31897, " Generic Guidelines for General Electric BWR Power Uprate." The staff finds these results acceptable.

- - - -- -. - -... - - -. - - - - -. _.- - - - - - -.I 3.5.1.1 Long-Term Suppression Pool Temperature Response (1) Bulk Pool Temperature The licensee indicated that the long-term bulk suppression pool temperature response was evaluated for the DBA LOCA at 102 percent of the uprated power by using the SHEX computer code. The SHEX code utilizes more refined models than used by the M3CPT/HXSIZ code in the original analysis to determine the suppression pool temperature.

SHEX code is capable of modeling containment response to more accident scenarios than the HXSIZ code. Many of the models used in SHEX code are the same as or very similar to those used in the M3CPT code used to calculate the short-term containment temperature and pressure i

response following a LOCA.

In a safety evaluation on the GE BWR power uprate generic analysis, dated July 11, 1992, the staff stated that, although the SHEX code is not yet formally approved on a generic basis, use of SHEX code in place of M3CPT/HXSIZ code is acceptable on a plant-specific basis, provided adequate information was provided to justify its use.

The licensee for FitzPatrick has referenced a comparison of the peak suppression pool temperatures obtained using the SHEX computer code and the M3CPT/HXSIZ computer code with a consistent set of inputs for a DBA-LOCA which the staff has reviewed for the Fermi-2 power uprate amendment. The peak suppression pool temperature ottained with SHEX and HXSIZ showed close agreement (<l'F).

GE indicated that since FitzPatrick and Fermi-2 are similar plants (BWR/4 with Mark I containment) and because the basis model equations for SHEX and HXSIZ are similar and verification of SHEX has shown that it performs accurate mass and energy balances on the suppression pool during a LOCA, it is expected that the results of SHEX and HXSIZ LOCA analysis for FitzPatrick would show similar agreement (approx. l'F difference) as shown for Fermi-2. Based on the above review, the staff finds the use of SHEX code acceptable for FitzPatrick.

The licensee has performed analyses at current rated and uprated power using the SHEX code and updated plant parameters associated with power uprate and consistent with the technical specifications. The licensee assumed a higher dome pressure, a higher initial suppression pool temperature and a higher RHR service water temperature, which are all conservative in nature. The licensee also indicated that the current analyses have been done using a more realistic decay heat model used for the UFSAR analysis (ANS/ ANSI 5.1 verses May-Witt).

The staff finds the use of the ANS 5.1 decay heat model acceptable.

The licensee indicated that using the above assumptions, SHEX predicted a peak suppression pool temperature of 208.7 'F for uprated power conditions and a peak pool temperature of 206.7 'F at the current rated power which is more conservative than the UFSAR value of 204 'F based on the current power level.

The uprated peak suppression pool temperature of 208.7 'F remains below the pool design temperature of 220 *F.

l The licensee stated that GE has also performed another analysis at uprated power with inputs which maximize the peak suppression pool temperature and, for that peak temperature, minimize the suppression chamber pressure. This analysis was performed to determine suppression pool temperature and suppression chamber pressure for use in evaluating the ECCS and RHR pump NPSH adequacy. The licensee indicated that the inputs included initial conditions that minimize the amount of non-condensible gas in the drywell and suppression chamber airspace, and modeling of continuous containment spray operation, which reduces the suppression chamber pressure during the event.

For this case, the available ECCS pumps NPSH was determined to be approximately 12 psig at the peak suppression pool temperature. The licensee indicated that the required NPSH for RHR is 2.55 psig at the peak suppression pool temperature of 209 'F.

Thus, adequate margin is available to prevent cavitation. The other l

ECCS pumps do not require credit for the increase in containment pressure due to the LOCA.

Based on its review as discussed above, the staff concludes that the containment long-term peak bulk suppression pool temperature response will remain acceptable after power uprate.

(2)

Local Pool Temperature With SRV Discharge The licensee indicated that since the FitzPatrick plant has quenchers, no evaluation of this limit is considered necessary.

Elimination of this limit l

for plants with quenchers on the SRV discharge lines is justified in GE report NED0-30832, " Elimination of Limits on Local Suppression Pool Temperature for SRV Discharge with quenchers." However, the local suppression pool temperature has been evaluated at uprated power, and was found to be l

acceptable.

l l

Based on its review, the staff concludes that power uprate will not affect the local pool temperature limit.

l 3.5.1.2 Containment Gas Temperature Response The licensee indicated that the containment drywell design temperature of 309 *F was determined based on a bounding analysis of the blowdown of steam to l

the drywell during a LOCA. The maximum value calculated for power uprate conditions for the same event is 288 'F.

This value for power uprate l

conditions was determined using Mark I Long Term Program (LTP) methods with break flow rates and enthalpies calculated using the GE LOCA Analytical Model for peak drywell pressure. Therefore, the drywell gas temperature response with power uprate will not exceed the design value of 309 'F.

The licensee also indicated that the wetwell gas space peak temperature response is calculated assuming thermal equilibrium between the pool and l

wetwell gas space.

Since the power uprate analysis has increased the maximum pool temperature to 209 'F, the wetwell gas space temperature will also increase to 209 'F.

This is still lower than the wetwell space design temperature of 220 'F.

l

..,4 e

3 l i

Based on the above review, the staff concludes that the containment gas temperature response will remain acceptable after power uprate.

3.5.1.3 Short-Term Containment Pressure Response The licensee indicated that the short-term containment response analyses were performed for the limiting DBA LOCA, which assumes a double ended guillotine i

break of a recirculation suction line to demonstrate that operation with power uprate will not result in exceedirg the containment design pressure limits.

l The short-term analysis covers the blowdown period during which the maximum i

drywell pressure and differential pressure between the drywell and wetwell i

occur. These analyses were performed at 102 percent of the uprated power level using the GE M3CPT computer code which was reviewed and accepted by the l

NRC during the Mark I LTP for application to Mark I plants including l

FitzPatrick. The break flow rates and enthalpies for the uprated analyses l

were determined from the GE LOCA Analytical Model consistent with the Generic i

Power Uprate Guidelines. At uprated power level, these methods predicted a maximum containment pressure of 41.2 psig for the limiting DBA LOCA against j

the 38.3 psig determined during Mark I long-term program at the current power level. The containment is designed for a pressure of 56 psig.

Therefore, the maximum pressure of 41.2 psig at uprated power remains below the containment design pressure.

Based on its review, the staff concludes that the FitzPatrick containment pressure response following a postulated LOCA will remain acceptable after the power uprate.

3.5.2 Containment Dynamic Loads 3.5.2.1 LOCA Containment Dynamic Loads NEDC-31897 requires that the power uprate applicant 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 4

loads, then LOCA dynamic loads are not affected by power uprate and thus do not require further analysis.

The licensee indicated that the results of the short-term LOCA containment response analyses were primarily used to evaluate the LOCA dynamic loads such as pool swell, condensation oscillation, chugging and vent thrust loads. This evaluation showed that the containment response conditions with power uprate are within the range of test conditions used to define the pool swell, condensation oscillation and chugging design loads. Also, the vent thrust loads with power uprate are calculated to be less than the plant-specific values calculated during the Mark I Containment LIP. The LOCA dynamic design loads at power uprate conditions kre therefore acceptable.

- -=.

3.5.2.2 Safety Relief Valve (SRV) Containment Dynamic Loads The SRV containment dynamic loads include discharge line loads, pool boundary l

loads and drag loads on the submerged structures. These loads are influenced by the SRV opening setpoint pressure, discharge line configuration and suppression pool configuration. Of these parameters, only the SRV setpoint pressure is affected by power uprate. NEDC-31897 (Reference 3) states that if i

l the SRV setpoints are increased, the power uprate applicant will show that the SRV design loads have sufficient margin to accommodate the higher setpoints.

The licensee indicated that the FitzPatrick SRV opening pressure could be increased to 1195 psig without exceeding the allowable stresses in the affected components. With a 3 percent tolerance in the nominal opening pressure setpoint, the nominal opening setpoint corresponding to an opening pressure of 1195 psig would be 1160 psig. The highest SRV nominal opening pressure for FitzPatrick with power uprate will be 1145 psig and thus, will l

remain bounded by the 1160 psig allowable value.

Based on the above, the staff finds that the SRV containment dynamic loads for FitzPatrick with power uprate will not exceed the existing allowable stresses and therefore, will be l

acceptable at power uprate conditions.

3.5.2.3 Subcompartment Pressurization The licensee stated that the annulus or subcompartment pressurization is not part of the original licensing basis for FitzPatrick and therefore this analysis was not performed for power uprate. The licensee also stated that a review of other plants indicates that the effects of a 4 percent to 5 percent thermal power uprate on annulus pressurization loads are not significant.

Also NEDC-31897 does not require subcompartment reanalysis.

The staff finds that the licensee's response that the effect of proposed power uprate on annulus loads will not be significant is acceptable.

3.5.3 Containment Isolation The licensee indicated that containment isolation is not affected by power uprate. The uprated peak drywell pressure remains bounded by the original Appendix J test pressure. Additional assurance is provided by the FitzPatrick program to comply with Generic Letter 89-10 for motor-operated valves (MOV) which assures that safety-related MOVs impacted by power uprate are evaluated at uprated conditions and any resulting necessary modifications are implemented prior to operation at uprated conditions.

Based on its review, the staff finds that the operation of the plant at uprated power level will not impact the containment isolation system.

3.5.4 Standby Gas Treatment System (SGTS)

The SGTS is designed to ensure controlled and filtered release of particulates j

and halogens from primary and secondary containment to the environment during l

abnormal and accident situations in order to maintain offsite thyroid doses

_ _ _ _ _. - - _ _. _ within the 10 CFR Part 100 limits. The system consists of two identical, physically and electrically separated,100 percent capacity air filter trains.

Each train is sized to change one secondary containment (SC) air volume per day while maintaining the SC at a slight negative pressure of 1/4 inch water gauge with respect to the outside atmosphere. Maintaining this negative pressure serves to prevent unfiltered release of radioactive material from the SC to the environment.

In its submittal, the licensee noted that the proposed uprate in power (4.1 percent) by itself will not have any adverse impact on the capability of the SGTS to meet the above design objective since it does not change the ventilation design aspect of the SGTS. After further review, the staff finds the SGTS acceptable for power uprate.

The staff recognizes that iodine loading in the filters will increase marginally (4.1 percent) due to the proposed pown uprate.

The SGTS design utilizes filters that meet the intent of Regulatory Guide (RG) 1.52 guidelines with respect to the design, testing, and maintenance criteria of ESG grade filters. One of the criteria deals with the filter loading capability. The licensee has determined that although the iodine loading will increase slightly, it will remain well below the original design capacity of the filters. Based on the above, the staff concludes that the SGTS will continue to meet the guidelines of RG 1.52 and remain acceptable for the proposed uprated power operation.

Based on the above findings, the staff concludes that the uprated power level operation will have an insignificant impact on ti.e capability of the SGTS to meet its design objectives.

3.5.5 Primary Containment Atmosphere Control and Dilution System The licensee indicated that the primary containment atmosphere control and dilution system provides nitrogen gas as required to the containment for the containment inerting and de-inerting, makeup, dilution control and supply to containment instrumentation. The containment is normally inerted, and nitrogen is added as necessary to maintain oxygen concentrations below 4 percent. The containment atmosphere dilution (CAD) system can supply nitrogen at a variable pressure sufficient to inject nitrogen up to the containment design pressure to maintain the post-LOCA oxygen concentration below the allowable limit of 5 percent.

The licensee indicated that the CAD system nitrogen addition rate was evaluated for power uprate and it was determined that the original analysis envelops uprated power conditions and that it can provide adequate flow and pressurization of the containment after a LOCA. The hydrogen generated calculations provided by the licensee indicate that less hydrogen will be liberated due to core-wide metal-water reactions than previously predicted.

This slight decrease is primarily due to significantly reduced predicted fuel cladding temperatures during a postulated LOCA.

Based on the above review, the staff concludes that the operation of the plant

! at uprated power level will not impact the primary containment atmosphere control and dilution system as the original analysis envelops the uprated power conditions.

3.6 Other Enaineered Safety Features (ESF) and Power Distribution Systems 3.6.1 Main Control Room Atmosphere Control System l

The main control room atmospheric control system function is provided by the control room emergency ventilation system (CREVS). The CREVS provides for control room habitability during emergency conditions.

The system provides safety related cooling, radiation protection and toxic protection. The CREVS design meets the NUREG-0800, Standard Review Plan, Section 6.4, guidance. The zone isolation is provided with incoming air filtered and a positive pressure maintained by the system fans.

The control room emergency zone (CREZ) volume is 148,300 cubic feet and it includes (a) the main control room includin shift supervisor office, hall areas, kitchen, and toilet / washroom areas,gand (b) operation department's office and HVAC equipment room.

l The staff review is based on the licensee's revised submittal for Response to the NUREG-0737, Item III.D.3.4, Control Room Habitability, dated March 2,1995, and was supplemented by several telephone conversations for clarifications during the weeks of November 25, and December 2, 1996. The

. control room is served by two safety-related, seismic Category I, full capacity redundant trains of the CREVS.

Each train consists of an air l

handling-cooling unit, recirculation / exhaust fan, emergency makeup filtration l

unit and emergency control room supply fan. 'There is no automatic isolation capability for the CREVS during accident conditions.

The CREVS is designed for fail-safe operation in the event of instrument or equipment failure.

If the CREZ exhaust temperature exceeds 98* F, both air handling units and chillers will start automatically to provide maximum cooling.

l Each seismic Category I emergency makeup filtration unit consists of a 1

prefilter, high efficient particular air (HEPA) filter, two 2-inch charcoal adsorbers in series, post-HEPA filter, 100 percent capacity booster fan with a l

capability of providing 10001 10 percent cubic feet per minute (cfm) and j

associated ductwork and instrumentation and controls.

The emergency makeup is used during radiological conditions to pressurize the control room zone to a positive pressure of 1/8-inch W.G. relative to potentially contaminated surrounding areas to minimize unfiltered inleakage of contaminated outside air into the control room emergency zone. Seismically qualified primary and secondary emergency air intakes are provided for emergency makeup air; however, only the primary intake is tornado missile protected. Therefore, the i

licensee's control room dose evaluation does not take credit for dual intake j

in accordance with the guidance of NUREG-0800, Standard Review Plan, Section 6.4.

The primary and secondary air intake valves are manually operated from i

the control room to choose the most suitable air intake during the accident conditions.

i

1 i

, I l

The HEPA filter efficiency is greater than 99.9 percent when handling air from 98-100 percent relative humidity (RH) based on the DOP test method. HEPA filter and charcoal adsorber surveillance tests for CREVS and SGTS are conducted in accordance with Technical Specifications (TS) 4.11.A and 4.7.B.1.c, respectively. The CREVS components are connected to safety-related power except for the kitchen and toilet exhaust fan and motor-operated modulating damper, and CREVS motor-operated modulating intake and exhaust dampers. These excepted components are seismically designed and fail in their safe position upon loss of power.

The new charcoal for the CREVS and standby gas system (SGTS) is purchased in accordance with the guidance of Regulatory Guide 1.52, Revision 0 (1973),

Table 1, Item 5.b with the test conditions of 30' C, 101 kPa and 95 percent RH.

Even though each CREVS emergency makeup filtration unit has two, 2-inch beds in series, only one, 2-inch bed is credited in the licensee's DBA dose analyses. A representative charcoal sample from one of the 2-inch beds is laboratory tested annually under the licensee's administrative procedure. The SGTS and CREVS utilize the identical test conditions which call for the laboratory testing of the used charcoal at 25' C,.101 kPa and 70 percent RH.

However, only the SGTS has electric heaters to reduce the RH to 70 percent RH.

The acceptance criteria for the laboratory testing of the used charcoal for the CREVS and SGTS is a methyl iodine penetration of $0.5 percent and $1.0 percent, respectively, when tested in accordance with ASTM D3803, 1979.

While testing the CREVS used charcoal at 70 percent RH is not as conservative as testing at 95 percent RH, the licensee conservatively takes credit for only one of the two, 2-inch charcoal beds and assumes only a 90 percent efficiency for the filtration unit.

The licensee replaced the used charcoal with the new activated charcoal a year ago and it is expected that the charcoal would meet the 99.5 percent methyl iodine removal acceptance criteria using the more conservative test protocol of 25* C,101 kPa and 95 percent RH. Therefore tne staff finds the assumption of 90 percent efficiency for the emergency makeup filtration unit to be acceptable. Additionally, the licensee is considering revising their administrative control procedure to comply with the latest codes and standards for the charcoal testing.

These conditions may be reflected in future adoption of the new improved standard technical specifications.

Upon receipt of a high radiation alarm from the radiation monitor the CREVS is manually placed in the isolation mode. The radiation monitor is installed in the intake duct of the CREVS. The control room intake radiation monitor is not safety-related since the control room is not automatically isolated and other non-safety related indicators, such as area radiation monitors, are available to alert the control room operators of radioactive releases.

Following the DB LOCA, the licensee assumed that the rate of unfiltered inleakages to be 15,0000 cfm until the CREVS is placed in the isolation mode, and then 2,100 cfm thereafter, based on a single failure of motor-operated intake valve 70M0V-108 to close and a failure to manually close bypass intake damper 70DNPR-105.

For the DB main steam line break (MSLB), the licensee used an unfiltered inleakage of 100 cfm for the entire accident, because the CREVS

_ _ _ _ is placed in the isolation mode if the reactor coolant activity levels reach 0.01 pc/gm.

Additionally, five air cylinders (330 cubic feet per cylinder) and five masks with air lines are located in the control room which conform to Regulatory l

Guide 1.3 and NUREG-0800, Standard Review Plan, Section 6.4, guidance. The Emergency Plan Procedures require periodic checks on the bottled air supply.

The control room espacity is adequate to maintain a staff of six persons for i

five days during the emergency conditions.

The staff has re-reviewed Section 4.5.4, Main Control Room Atmosphere Control l

System, based on the licensee's revised response to NUREG-0737, Item III.D.3.4, Control Room Habitability, dated March 2, 1995.

Based on our review, we find the revised submittal acceptable.

The design Basis Accident t

(DA) dose analysis and toxic releases evaluations will be provided in other sections.

Based on the above findings, the staff concludes that the uprated power level by itself will have little or no impact on CREVS meeting its design conditions.

l 3.6.2 Offsite and Onsite Power Distribution Systems / Grid Stability l

l The licensee has described their evaluation for the plant uprate regarding i

continued conformance to GDC 17. GDC 17 addresses onsite and offsite electrical supply systems for the electrical equipment important to safety.

Their evaluation included consideration of both the onsite and offsite power j

systems.

The staff has concluded that the licensee has evaluated the uprate impact on the necessary electrical systems a.,d components and the results of those i

evaluations provided the licensee with assurance that the safety functions of the electrical power system can be maintained. Therefore, the staff finds the uprate to be acceptable because conformance to GDC 17 of Appendix A to 10 CFR 50 is maintained.

3.7 Fuel Pool Cooling l

Three cooling systems are provided at Fitzpatrick to remove decay heat from stored spent fuel assemblies: the Spent Fuel Pool Cooling and Cleanup System l

(FPCCS); the Residual Heat Removal System in the Fuel Pool Cooling Assist Mode l

(RHR-FPCA); and the (Alternate) Decay Heat Removal System.

In the past, FPCCS and RHR-FPCA have been available in combination to maintain SFP temperatures below design limits under a variety of decay heat loads. The licensee recently installed the DHR system to provide added flexibility for SFP cooling during outages and because it was desirable for managing shutdown risk. The systems are designed to maintain the SFP below its design temperature of 150 'F under a variety of heat loads and offload scenarios. A description of j

each system's design information is provided below.

I l

l

! The FPCCS is designed to remove decay heat generated by the stored spent fuel 1

assemblies and maintain a pool outlet water temperature below 135 'F for a i

peak annual refueling heat load (partial core) of 10 M8tu/hr.

The system consists of two 100 percent pumps, two 50 percent capacity heat exchangers, and one 100 percent capacity filter. The two heat exchangers are sized for a combined heat load of 6.3 M8tu/hr with one pump operating, maintaining a pool temperature of l

125 'F.

With both pumps operating, FPCCS can remove 10 M8tu/hr while maintaining the fuel pool outlet temperature below 135 'F.

RHR-FPCA can be used during full core offloads to supplement FPCCS. Any one of four RHR pumps can be aligned to one of the two RHR heat exchangers in the fuel pool cooling assist mode.

RHR-FPCA and FPCCS with one pump and two heat exchangers combined have sufficient capacity to maintain SFP outlet water temperature less than 135 *F for a decay heat load of 24 M8tu/hr. The second FPCCS pump can also be operated to provide additional cooling.

The DHR system consists of two 100 percent capacity pumps, two 100 percent capacity heat exchangers, and a strainer, and is used independent from FPCCS and RHR-FPCA. The DHR system is provided with its own independent secondary system consisting of two pumps and two sets of cooling towers.

In the

" nominal" cooling configuration, the DHR system is operated with one primary pump and heat exchanger, and one secondary pump and set of cooling towers.

In this configuration, the system is designed to remove 30 MBtu/hr of decay heat, equivalent to the combined reactor core and SFP decay heat loads after approximately 4.5 days of decay.

In the " maximum" cooling mode, which consists of one primary pump, two heat exchangers, two secondary pumps, and both sets of cooling towers, the DHR system is capable of removing 45 MBtu/hr.

This heat load is equivalent to the combined reactor core and SFP decay heat loads after 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br />.

In either configuration, the DHR system is able to maintain SFP outlet temperature below 125 'F.

The licensee described the recently installed DHR system in a letter to the staff dated July 26, 1996, and provided additional information about their plans to manage decay heat in the spent fuel pool in a followup letter dated October 2, 1996.

For refueling outages where the DHR system is not available, the operating limitations on fuel movement will continue to remain in effect: 96 hour0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br /> must i

elapse after reactor shutdown before fuel movement to the SFP can begin, fuel movement must not exceed four fuel assemblies per hour, and the temperature of the SFP will be maintained between 68 'F and 125 'F for partial core offloads using the FPCCS, and 68 'F and 135 'F for full core offload using the FPCCS, l

with RHR in the fuel pool cooling assist mode. The licensee calculated that I

under design. conditions, the transitional SFP bulk temperature could reach a maximum temperature of 148 'F shortly after a full core offload. This is within the SFP structural design temperature of 150 'F.

With the DHR system available for the refueling outage, the licensee provided analysis that confirms the system can maintain SFP bulk temperature below

i 1 !

125 *F under all offload conditions.

For the bounding analysis, the licensee assumed a 36 hour4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> hold time after reactor shutdown before moving fuel resulting in a maximum heat load of 45 MBtu/hr.

No other limitations were included in the licensee's analysis (e.g., fuel assembly offload rate).

Compared to normal industry practice, the staff considers the licensee's assuuption of beginning fuel transfer 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> after shutdown to be conservative. As described in their October 2, 1996 letter, the licensee will maintain the RHR system in the FPCA or Shutdown Cooling mode as a backup to 1

DHR if refueling begins during the period from 36 hours4.166667e-4 days <br />0.01 hours <br />5.952381e-5 weeks <br />1.3698e-5 months <br /> to 4.5 days after i

shutdown.

With RHR available as a backup to DHR during this time, the plant is maintained in a configuration that will tolerate the failure of an active component in the DHR system with no degradation of decay heat removal function.

4 The licensee calculated that the capacity of the DHR system in the " nominal" mode will exceed the decay heat load ir. the SFP after the reactor has been shut down 4.5 days. The licensee stated, at this point, they will be capable of providing redundant cooling capability by maintaining the DHR system available to operate in the " maximum" cooling mode, or having RHR available in i

the Shutdown Cooling or RHR-FPCA mode. With the DHR system or RHR system maintained in this configuration, the operating system can tolerate the failure of an active component with no degradation of decay heat removal 4

function. While this is not required in the plant's current licensing basis, d

j the licensee finds this plant configuration conservative for managing shutdown risk.

As a result of reactor operations at uprated power levels, the licensee determined that the decay heat for any specific fuel discharge scenario will increase slightly, but will remain within the design limits of the affected systems. The results of the licensee's analyses indicate that sufficient i

capacity exists in the FPCCS and the RHR system to accommodate the increase in decay heat loads due to the proposed power uprate without increasing the maximum SFP temperature above the limits already accepted by the staff in Amendment 175 to license DPR-59. The licensee's analyses also indicate that when DHR is used during refuelings at uprated power, the maximum SFP temperature will not exceed the temperature limits already accepted by the staff in Amendment 175 for refueling operations using the FPCCS and RHR-FPCA.

For all refueling operations, analyses indicate that SFP temperature will be maintained below the design temperature of 150 *F.

On that basis, the staff finds that plant operations at the proposed uprated power levels are acceptable with respect to decay heat removal from the spent fuel pool.

Separately, an issue associated with spent fuel pool cooling adequacy was identified in NRC Information Notice 93-83 and its Supplement 1, " Potential Loss of Spent Fuel Pool Cooling Following a Loss of Coolant Accident (LOCA),"

i dated October 7, 1993 and August 24, 1995, respectively, 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 a

are warranted, the staff will address those requirements to the licensee under separate cover.

l 1 3.8 Water Systems l

l The licensee evaluated the impact of power uprate on the various plant water systems including the safety-related and non-safety-related service water systems, closed loop cooling water system, circulating water system, and the plant ultimate heat sink.

The licensee's evaluation considered incrcased heat 1

loads, temperatures, pressures, and flow rates.

3.8.1 Service Water Systems 1

l 3.8.1.1 Safety-Related Loads i

l l

These systems include the emergency service water (ESW) system and the residual heat removal service water (RHRSW) system. All heat removed by these systems is rejected to the atmosphere via the ultimate heat sink (VHS). The staff's evaluation of the effects of uprated power level operation on each of these systems is provided below.

3.8.1.1.1 Emergency Service Water System l

The ESW system was evaluated for its ability to provide cooling to ECCS components and other equipment essential to safe reactor shutdown during a design basis LOCA.

Based on its review, the staff finds that the original design loads for this system were based on equipment loads which are considerably greater than the anticipated equipment loads resulting from l

uprated power operation. Design flow rates will be maintained and remain adequate for uprate. Consequently, the staff concludes that the uprated power l

level operation has minimal impact on the ESW system operation.

3.8.1.1.2 Residual Heat Removal Service Water System l

The RHRSW system is designed to assure adequate cooling to Class I RHR heat l

exchangers during normal reactor shutdown cooling. As a result of the uprated i

power level operation, the following functions of the RHRSW system will be affected to a minor degree when operating in:

a.

the reactor shutdown cooling mode, b.

the spent fuel pool cooling (backup system) mode, and c.

the suppression pool cooling mode following a LOCA.

However, the shutdown cooling mode is the controlling mode for residual heat removal heat exchanger operation.

In its submittal, the licensee stated that the operational objective for normal shutdown is to reduce the bulk reactor temperature to 125 *F in 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, decay heat will increase proportional to uprate and will generate a i

small increase in the time required to reach reactor shutdown temperature.

In the analysis, which used an 82 *F delivery temperature compared to an original design temperature of 65 *F, the licensee determined the cooldown time to be

l l

l t 25 hours2.893519e-4 days <br />0.00694 hours <br />4.133598e-5 weeks <br />9.5125e-6 months <br />. The staff has reviewed the information provided and finds the licensee's evaluation acceptable and acknowledges that power uprate may i

l slightly increase cooldown time.

Power uprate will not prevent the RHRSW system from performing its design functions.

3.8.1.2 Non-Safety-Related Loads The normal service water system is designed to meet the following criteria:

- Provide a heat sink for the Turbine Building Closed Loop Cooling l

(TBCLC) and Reactor Building Closed Loop Cooling (RBCLC) systems during planned operations.

- Provide sufficient water for traveling screen wash nozzles, plant air conditioning and cooling units, mechanical vacuum pump, vacuum priming pump seal water coolers, steam jet air ejector precoolers, and other l

miscellaneous services.

The licensee indicated that the increase in normal service water (SW) system l

heat loads is projected to be approximately proportional to the uprate in power and that this increase is insignificant to the design of the system.

The SW system is capable of supplying sufficient water to remove the additional heat loads due to uprated power conditions.

In its submittal, the licensee indicated that although the SW system return water temperature may l

increase, it will not exceed current Technical Specification (TS) limits and will have no effect on the capability and design of the system.

Based upon (1) NRC staff experience with previous uprate applications for j

similar BWR plants, (2) the licensee's conclusion that the performance of the SW system will be adequate, and (3) since the SW system does not perform any safety function, the staff has not reviewed the impact of the uprated power level operation on the design and performance of this system.

3.8.2 Main Condenser and Circulating Water Systems The main condenser and circulating water systems are designed to provide the main condenser with a continuous supply of cooling water for removing heat rejected to the condenser by turbine exhaust, turbine bypass steam, and other i

l exhausts over the full range of operating loads thereby maintaining adequately low condenser pressure. The licensee indicated that the performance of the main condenser and circulating water system was evaluated and is adequate for uprated power level operation.

Based upon (1) NRC staff experience with previous uprate applications for similar BWR plants, (2) the licensee's conclusion that the performance of the main condenser and circulating water systems will be adequate, and (3) since the main condenser and circulating water systems do not perform any safety function, the staff has not reviewed the impact of the uprated power level operation on the designs and performances of these systems.

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_ _ _ _ _. _ _ __ _ ~ - _ _ _,_

_ - 3.8.2.1 Discharge Limits The licensee compared the state discharge limits to current discharges and bounding analysis discharges for power uprate. The comparison demonstrates that the plant will remain within the state discharge limits during operations at uprated power level.

r Based on this comparison and NRC staff experience gained from review of power uprate applications for similar BWR plants, the NRC staff finds that plant operations at the proposed uprated power level will have an insignificant impact on the main condenser / circulating water discharge limits.

3.8.3 Reactor Building Closed Loop Cooling Water System The reactor building closed loop cooling water (RBCLCW) system is designed to provide required cooling to the 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 believes the increase in heat load due to power uprate to have an insignificant effect on RBCLCW system design.

Based on its review, the staff has concluded that the effect of uprated power operation on the RBCLCW system is negligible and that there is sufficient diversity and operating margin for this system to perform adequately at uprated conditions.

3.8.4 Turbine Building Closed Loop Cooling Water System The turbine building closed loop cooling water (TBCLCW) system is designed to remove heat from both generator-related and non-generator-related equipment in the turbine building and radioactive waste building.

The system also provides makeup seal water to mechanical vacuum pumps and condensate vacuum priming pumps. The licensee determined that the increase in heat loads from this equipment due to the uprated power level operation is expected to be negligible and that the TBCLCW system design cooling capacity will not be exceeded.

Based upon NRC staff experience with previous reviews of power uprate applications for similar BWR plants, the licensee's conclusion that the increase in heat loads to this system due to uprated power operations is insignificant and within design heat loads, and the fact that the TBCLCW system does not perform any safety-related function, the NRC staff has not reviewed the impact of the proposed uprated power operations on the design and performance of this system.

3.8.5 Ultimate Heat Sink i

Since the ultimate heat sink (UHS) is Lake Ontario, the licensee indicated j

that the environmental effects of uprate will be controlled at or near the same levels as for the original analyses. The licensee determined that the

.~-

! existing UHS system will continue to provide a sufficient quantity of water following a postulated LOCA for decay heat removal and that the TS for RHR reservoir level is adequate due to conservatism in the original water requirement calculations. The staff finds that the licensee's conclusion, that the UHS design is adequate for the uprated power level operation, is acceptable.

Based on its review, the staff concludes that uprated power operation will have little or no effect on the existing UHS to control the plant ambient temperatures, humidity, and flow of potential radioactive contaminants.

3.9 Power Dependent Heatina. Ventilation and Air-Conditionina (HVAC)

The licensee indicated that the uprated heat loads would have minimal or no impact on maintaining the design environmental temperature parameters since the uprated parameters are within the envelope of the original safety l

evaluation performed for the FitzPatrick station. Therefore, the design basis for the existing HVAC equipment and systems, the normal environmental temperatures used to establish qualified equipment life, and the area design temperature for all plant operating modes envelops the anticipated increase in heat load and is not impacted by power uprate.

Based on experience with previous power uprates and staff review, the staff concludes operation of the plant at uprated power will have minimal or no impact on the plant HVAC systems.

3.10 Fire Protection Fire suppression or detection is not expected to be impacted due to plant operations at proposed uprated power level since there are no physical plant configuration or combustible load changes resulting from the uprated power operations. The safe shutdown systems and equipment used to achieve and maintain cold shutdown conditions do not change and are acceptable for the uprated conditions. Also, the operator actions required to mitigate the consequences of fire are not affected.

Based on NRC staff review of the licensee's rationale and the experience gained from the review of power uprate applications for similar BWR plants, the NRC staff finds that the fire suppression and detection systems are not power dependent and will not be affected by plant operations at the proposed uprated power level.

_ _ _ _ _ 3.11 Other Systems Reviewed for Imoact by Power Vorate The licensee identified other systems which are not affected by plant operations at the proposed uprated power level. The NRC staff reviewed the following systems for which it has a review responsibility:

- Compressed Air

- Contaminated Equipment Vents

- Diesel Generator Room Ventilation

- Emergency Diesel Generators

- Hydrogen Storage and CO Purge Supply 2

- Lube Oil

- Demineralized Water

- Sample System

- Screenwell and Pumphouse Ventilation

- Sewage Treatment

- Tools and Servicing Equipment

- Radwaste Building Ventilation

- Stack and Stack Equipment Based on NRC staff review and experience of previous power uprate analysis, the NRC staff finds that plant operations at the proposed uprated power level has no impact on these systems.

3.12 Systems with Minimal Imoact l

The licensee identified and evaluated the systems which are affected in a very minor way by operation of the plant at the uprated power level. These include the following systems:

- Hydrogen Water Chemistry

- Local Panels and Racks

- Process Computer

- Extraction Steam

- Stator Cooling Water Based upon NRC staff experience with previous power uprate applications and the fact that these systems do not perform any safety-related function, the NRC staff has not reviewed the impact of the proposed uprated power operations on the designs and performance of these systems.

3.13 Turbine-Generator l

Evaluations for turbine operations with respect to design acceptance criteria to verify the mechanical integrity under the conditions imposed by the power uprate were performed.

Results of evaluations showed that there would be no increase in the probability of turbine overspeed nor associated turbine missile production due to plant operations at the proposed uprated power level. Therefore, the licensee concluded that the turbine could continue to be operated safely at the proposed uprated power levels.

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j Based on NRC staff review, the NRC staff finds that operation of the turbine j

at the proposed uprated power level is acceptable.

j 3.14 Power Conversion Systems

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The steam and power conversion systems and associated components (e.g. the i

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

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feedwater and condensate systems, etc.) were originally designed to utilize i

the energy available from the nuclear steam supply system and to accept the system and equipment flows resulting from continuous operation at 105 percent of the currently licensed rated power. Since the requested uprate is less than values used in the earlier analyses and NRC staff has experience with previous power uprate reviews for similar BWR plants, the staff finds that j

operation at uprated power should not have a significant impact on the power conversion systems.

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3.15 Radwaste Systems and Radiation Sources The licensee evaluated the impact of the proposed amendment to show that the applicable regulatory acceptance criteria continue to be satisfied for the uprated power conditions.

In conducting this evaluation, the licensee 3

considered the effect of the higher power level on source terms, on-site and i

4 j

off-site doses and control room habitability during both normal operation and i

accident conditions.

In reviewing this amendment request, the staff I

considered only the effects of this 2 percent uncertainty factor since the evaluations of radiological aspects of operations at a power level of 2550 had already been considered in the staff review of the original license application.

?

3.15.1 Liquid Waste Management j

The liquid radwaste system is designed to process the majority of the liquid wastes within the plant so that liquids discharged from the plant satisfy the 3

10 CFR Part 20 and 10 CFR Part 50 Appendix I requirements.

The licensee has noted that the volume of liquid waste is not expected to increase due to the proposed power uprate.

Reactor coolant cleanup flows, leaks, laboratory drains, dry solid waste, and spent resin quantities will remain essentially the same after uprate.

There will be an increase in activated corrosion a

products proportional to the proposed power uprate. Source term values are expected to increase for some radionuclides and decrease for others. The net effect is that the radwaste system will basically remain the same.

Based on a review of plant operation, effluent reports, and a consideration of the expected slight increase in effluents as a result of power uprate, the licensee determined that the requirements of 10 CFR Part 20 and 10 CFR Part 50 Appendix I will continue to be satisfied.

The licensee also evaluated the condensate demineralizer capability. The l

licensee concluded that the condensate deminaralizers had sufficient capacity to accommodate the proposed power uprate.

In addition, the licensee noted that the expected time between ultrasonic cleaning or regeneration is not expected to be significantly reduced at the uprated power level.

. - - -. =. Based on its review of available plant data and experience with previous power uprates, the staff agrees with the licensee's conclusion that the operation at uprated power levels will have no significant adverse effect on liquid 4

effluent and is, therefore, acceptable.

3.15.2 Gaseous Waste Management Gaseous wastes generated during normal and abnormal operation are collected, controlled, processed, stored, and disposed of utilizing the gaseous waste processing treatment systems. These systems include the offgas system, the standby gas treatment system, and other building ventilation systems.

Various devices and processes, such as radiation monitors, filters, isolation dampers, and fans are used to control airborne radioactive gases. The gaseous waste management systems are designed to meet the requirements of 10 CFR Part 20 and l

10 CFR Part 50, Appendix I.

Results of licensee analyses demonstrate that airborne effluent activity released through building vents is not expected to increase significantly after power uprate.

The licensee also performed an evaluation of the combined effects of power uprate and hydrogen water chemistry (HWC) which revealed that the H concentrationinthefeedwaterwillhavetobeincreasedtomaintai$ desired values. As a result of increased hydrogen addition, the total offgas hydrogen flow will remain less than during non-HWC operation. Also, the recombiner exit temperature will remain less than the non-HWC temperature.

On the basis of its review of available plant data and previous experience j

with other power uprates, the staff concludes that there will not be a significant adverse effect on airborne effluent as a result of the power j

uprate.

3.15.3 Radiation Sources in the Core and Coolant Radioactive materials in the reactor core are produced in direct proportion to the fission rate. Thus, the expected increase in the levels of radioactive materials (for both fission products and activation products) produced are expected to increase by a maximum of 4.1 percent. The licensee noted that experience to date with operation of FitzPatrick indicates that concentrations of fission and activation products in the reactor coolant will not increase significantly above those currently experienced.

Current experience with operation of FitzPatrick indicates that concentrations of fission and activation products in the reactor coolant will not increase significantly above those currently experienced. Current experience with operation of FitzPatrick indicates that the unit operates well below the 0.1 Curie /sec design basis and that current off-site radiological release rates are well below the original design basis, i

Based on the staff's review of available plant data and its experience with other power uprates, the staff concludes that no significant adverse effect on radiation sources in either the core or reactor coolant is expected from the l

proposed power uprate.

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3.16 Eauionent Qualifications (EO)

The licensee evaluated the effcets of plant operations at the proposed power level on qualified equipment including safety-related electrical equipment and mechanical components.

l 3.16.1 Equipment Qualifications (EQ) of Electrical Equipment The licensee performed evaluations of environmental parameters resulting from normal, abnormal, and accident conditions to determine if safety-related

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electrical equipment qualified prior to uprate would be acceptable for use j

under uprate conditions. This evaluation focused on equipment in the drywell (primary containment), reactor building (secondary containment), turbine 3

building, and steam tunnel.

Based on the evaluations performed, the licensee revised qualified life estimates for some components as a result of increased i

normal operating temperatures. With the revision to qualified life estimates, the licensee determined that environmental qualification of all EQ components l

is established for power uprate conditions.

Based on its review and experience with power uprates at other nuclear plants, I

i the staff finds the licensee's conclusions regarding qualification of safety I

j related electrical equipment for power uprate acceptable.

l 3.16.2 Equipment Qualification (EQ) of Mechanical Equipment with 1

Non-Metallic Components i

i The licensee performed evaluations of environmental parameters resulting from l

normal, abnormal, and accident conditions to determine if safety-related mechanical equipment qualified prior to uprate would be acceptable for use i

under uprate conditions. These evaluations determined that increased I

operating temperatures, pressure, and flow conditions due to power uprate l

would not affect the qualification of mechanical equipment.

The licensee also evaluated changes in operating radiation levels due to uprate power levels and other design changes (e.g., increased hydrogen addition rate).

For the changed radiation levels, the licensee evaluated the j

materials of concern which include elastomeric seals, 0-rings, and other non-i metallic pressure retaining parts. The licensee determined that the change in dose due to power uprate had no effect on the qualification of the equipment and the increase due to other design changes, when combined with the power l

uprate, had an insignificant effect on the qualification of equipment.

1 l

The licensee concluded that the power uprate would not adversely affect mechanical equipment and, therefore will not affect the ability of the mechanical equipment to perform its intended function.

Based on its review j

and experience with power uprates at other nuclear plants, the staff finds the j

licensee's conclusions regarding qualification of safety related mechanical equipment for power uprate acceptable.

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1 3.17 Mechanical Comoonent Review and Reactor Vessel Fracture Touahness 4

The staff's review of the safety analysis report provided by the licensee focused on the effects of power uprate on the structural and pressure boundary integrity of the piping systems and components, their supports; reactor vessel and internal components; the control rod drive mechanism (CRDM); and the balance-of-plant (BOP) piping systems.

1 The GE generic guidelines (Reference 3) for BWR power uprate were based on a i

5 percent higher steam flow, an operating temperature increase of 5 'F, and an operating pressure increase of 40 psi or less.

For JAFNPP, the maximum 1

reactor dome pressure increases from 1020 psia to 1055 psia (an increase of j

35 psi),thedometemperatureincreasesfrom547'Fto5p1*F(anincrease of 4 'F) gnd the steam flow rate increases from 10.47x10 lb /hr to 1

10.976x10 lb /hr. The maximum core flow rate remains unchan,ged for the JAFNPP power,uprate conditions.

3.17.1 Reactor Vessel and Internals The licensee evaluated the reactor vessel and internal components by considering load combinations that include reactor internal pressure 1

i difference (RIPD), loss-of-coolant accident (LOCA), and seismic loads.

The j

seismic loads are unaffected by the power uprate.

The licensee determined that the licensing basis LOCA loads such as pool i

swell, condensation oscillation (CD), and chugging remain unchanged because j

the JAFNPP dynamic loads, defined during the Mark I long-term program, are bounding for the power uprate condition with respect to the drywell and i

wetwell pressure, vent mass and energy flow rate, and torus water temperature.

The licensee recalculated RIPDs for the power uprate shown in Table 3-1 of j

Reference 6, for normal, upset and faulted conditions.

The stresses and fatigue usage factors for reactor vessel components were evaluated by the licensee in accordance with the ASME Code,Section III, 1965 Edition and addenda through winter 1966 to assure compliance with the Code of Record. The load combinations for normal, upset and faulted conditions were considered in the evaluation. The maximum stresses for critical components summarized in Table 1 of Reference 7, were found to be within the design allowable.

The fatigue usage factors for the uprated power level were calculated for limiting components such as feedwater nozzle, recirculation inlet nozzle and closure region bolts, and found to be well within the code allowable as shown in Table 3-2 of Reference 5.

No new assumptions were used i

in the analysis for the power uprate condition.

The staff concluded that the maximum stresses and fatigue usage factor as provided by the licensee are within the Code allowable limits and are therefore acceptable.

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3.17.2 Control Rod Drive System The licensee evaluated the adequacy of the control rod drive mechanism (CRDM) in accordance with the code of Record, the ASME Code Section III, 1965 Edition and Addenda through Winter 1966.

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 power uprate (Reference 3). The licensee stated that the nominal steady state operating pressure of the CRDS will be approximately 1075 psig. The design pressure of the CRDM is 1250 psig which bounds the uprated power condition.

Based on its review, the NRC staff concludes that the CRDS will continue to meet its design basis and performance requirements at uprated conditions.

3.17.3 Reactor Coolant Pressure Boundary and Balance of Plant Piping and Components The licensee evaluated the effects of the power uprate conditions, including higher flow rate, temperature and pressure for thermal expansion, fluid transients and vibration effects on the reactor coolant pressure boundary (RCPB) and the balance-of-plant (BOP) piping, systems, and components. The components evaluated included equipment nozzles, anchors, guides, penetrations, pumps, valves, flange connections, and pipe supports. The original Code of Record, ANSI B31.1, Power Piping Code, 1967 Edition with Winter 1969 addenda was used.

No new assumptions were introduced that were not in the original analyses.

The RCPB piping systems evaluated include main steam and associated extraction and drains system, reactor recirculation line, reactor water clean-up (RWCU),

reactor core isolation cooling (RCIC), condensate and feedwater system, high pressure coolant injection (HPCI), residual heat removal (RHR) including containment spray system, core spray system and CRDS. The licensee's evaluation of the RCPB piping systems consisted of comparing the maximum increase in stress for the power uprate (due to increase in pressure, temperature and fluid transient loads) with the design margins available in the original design basis analyses.

If necessary, the licensee performed stress analyses in accordance with requirements of the Code and the Code addenda of record under the power uprate conditions. The licensee concluded that the Code requirements are satisfied for the evaluated piping systems and that power uprate will not have an adverse effect on the reactor coolant piping system design.

The adequacy of the BOP systems was determined from the uprated reactor and BOP heat balances. These systems include lines affected by the power uprate, I

such as the main steam relief valve discharge, main steam and feedwater systems outside the primary containment. The licensee evaluated the stress l

1evels for B0P piping based on increases in temperature and pressure of the design basis analysis input. The licensee concluded that, for B0P piping

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systems, there are sufficient margins between the original design stresses and j

the Code limits to accommodate the stress increase due to the power uprate.

The licensee evaluated pipe supports including anchorage, equipment nozzles, and penetrations by comparing the increased piping interface loads on the j

system components due to the power uprate thermal expansion, with the margin in the original design basis calculations, and performing detailed analyses i

using exact load combinations at the uprated conditions. The licensee found that with the exception of some nozzles and supports, there is sufficient margin between the original design stresses and the Code limits to accommodate the stress increase for all service levels at the uprated power. The licensee i

has completed the evaluation and resolution of pipe supports and nozzles involving the RHR system, HPCI system, and the reactor building closed loop

.j cooling water system.

The effect of power uprate conditions on thermal and vibration displacement j

limits was also evaluated by the licensee for struts, springs and pipe snubbers, and found to be acceptable. The licensee reviewed the original postulated pipe break analysis and concluded that the existing pipe break i

locations were not affected by the power uprate, and no new pipe break j

locations were identified.

Based on the above review, the staff concludes that the design of piping, components and their supports will be adequate to maintain the structural and i

pressure boundary integrity of the rector coolant piping and supports in the

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power uprate conditions 3.17.4 Equipment Seismic and Dynamic Qualification Based on it's review of the proposed power uprate amendment, the staff finds j

that the original seismic and dynamic qualification of the safety-related j

mechanical and electrical equipment is not affected by the power uprate conditions for the following reasons:

1.

Seismic loads are unchanged for power uprate; 2.

The original LOCA load conditions and jet impingement bound the power uprate conditions for FitzPatrick; i

J 3.

The slight increase in SRV loads is insignificant; and 4.

No new pipe break locations resulted from the uprated conditions.

3.17.5 Conclusion for Mechanical Component Review l

On the basis of its review, the staff finds that the licensee's proposed power uprate amendment has no adverse effect on the structural and pressure boundary integrity of piping systems, components, and their supports, reactor internals, core support structure, or the CRD system, and is, therefore, l

acceptable.

, 3.17.6 Reactor Pressure Vessel Fracture (RPV) Toughness Operation with the approximately 5 percent power uprate will result in a higher neutron flux at the reactor vessel wall, which would increase the integrated fluence over the period of plant life.

i The licensee presented an assessment of the impact of power uprate on the RPV

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in Section 3.3.1 of Reference 5.

The current design basis for end of life j

(EOL) fluence for JAFNPP is 32 effective full power years (EFPY) based upon i

40 years of power operation at 80 percent capacity factor. Based upon future operation, FitzPatrick will not reach 32 EFPY, therefore previous E0L i

evaluations are still valid. The applicability of GE Topical Report NED0-32205-A, Revision 1, February 1994, for the James A. FitzPatrick Nuclear Power Plant, was reviewed by the staff. The licensee submitted an equivalent margins analysis to demonstrate compliance with Appendix G, 10 CFR Part 50.

The staff determined the analyses in the GE Topical Report NED0-32205-A, Revision 1, are applicable to the FitzPatrick reactor vessel, and that the FitzPatrick reactor vessel will maintain margins of safety against fracture equivalent to those required by 10 CFR Part 50, Appendix G and the ASME code (in letter dated March 30,1995). The GE equivalent margins analyses are still applicable with power uprate.

Reactor Material Surveillance Schedule 10 CFR Part 50 Appendix H, " Reactor Vessel Material Surveillance Program Requirements," requires a material surveillance program to monitor changes in the fracture toughness of RPV ferritic materials.

Reactor licensees are required to meet 10 CFR part 50, Appendix H unless otherwise approved by the NRC staff.

10 CFR part 50, Appendix H requires that reactor beltline materials surveillance programs must comply with ASTM E185-73, -79, or -82, as modified by Appendix H.

The JAFNPP TS commit to a surveillance program in accordance with ASTM E185-82 and 10 CFR Part 50, Appendix H.

The current FitzPatrick pressure-temperature curves were generated based on Regulatory Guide 1.99, Revision 2.

The first surveillance capsule was removed in April 1985 after 6 EFPY.

The Charpy impact test results from the surveillance program are equivalent to those predicted by Regulatory Guide 1.99, Rev. 2.

The revised limits that l

restricted operating pressures and temperatures to assure that brittle fracture of the reactor vessel cannot occur and that vessel integrity is maintained, staff issued Amendments No. 113 October 22, 1987 and again revised by Amendment No. 158 April 26, 1990. The schedule for subsequent capsule removal was approved by the NRC.

NYPA plans to remove the second capsule during the 1996 refueling outage.

This schedule is designed to support operation following the 1998 refueling outage.

Specimens will be tested in accordance with ASTM E185-82 and 10 CFR Part 50, Appendix H.

NYPA plans to remove the third capsule sometime late in plant life.

The schedule for removal will be based on data from the second capsule, and will

_ _ _ _ be discussed and approved by the NRC in accordance with 10 CFR Part 50, Appendix H.

The licensee has a fourth capsule that is a spare and has not been scheduled for withdrawal. This fourth capsule contains reconstituted specimens and was installed one cycle after the first capsule was removed.

It will be considered for removal, based on test results of previous capsules. The schedule for the fourth capsule will be discussed with and approved by the NRC in accordance with 10 CFR Part 50, Appendix H.

10 CFR Part 50, Appendix H, Section IV (Report of Test Results) specifies that the results of surveillance capsule tests must be submitted to the NRC withip one year of capsule withdrawal and that a schedule must be provided for the l

submittal of changes to the TS P-T curves, if required.

Thus, based on this l

requirement, NYPA must submit the JAFNPP surveillance capsule test results l

within the required time frame along with a schedule for the submittal of any changes that may be needed to the TS P-T curves.

Pressure-Temperature Limit Curves and Upper Shelf Energy 10 CFR Part 50, Appendix G, Section IV, Item A.1.a states that reactor beltline materials must have Charpy USE in the transverse direction for the base material and along the weld for weld material, of no less than 75 ft-lb initially and must maintain Charpy USE throughout the life of the vessel of no less than 50 ft-lb.

This section also states that lower values of Charpy USE are acceptable if demonstrated to provide margins of safety against fracture equivalent to those required by Appendix G of Section XI of the ASNE Code.

The current pressure-temperature curves were developed from General Electric Report DRF 137-0010, " Implementation of Regulatory Guide 1.99, Revision 2 for the James A. FitzPatrick Nuclear Power Plant," dated June 1989. These curves l

are valid through 16 EFPY.

General Electric has evaluated the applicability of the current curves under power uprate conditions (Reference 10). After uprate, but prior to reaching 16 EFPY, the incremental exposure due to uprate will have an insignificant l

effect on the existing P-T curves, and they are still valid. The basis for this conclusion was comparing the changes in adjusted reference temperature (ART) and upper shelf energy (USE) assuming that the licensee operate at uprated power for the 32 EFPY life of the RPV. The change in ART was only 2.5 'F and USE remained above 50 ft-lbs at end of life for all beltline materials except for welds 2-233 and 1-240.

For these welds, the licensee performed an equivalent margin analysis. The equivalent margin analysis l

performed by GE are applicable to EOL at the uprated power fluence.

l Therefore, the staff concludes that the change in ART and USE is not significant.

The P-T curn s will be revised based upon the second capsule data and j

Regulatory Guide 1.99, Revision 2.

The results from this capsule will be submitted to the NRC in late 1997 as required by 10 CFR Part 50, Appendix H.

The licensee estimates that the new curves will not be used until 1999, i

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The current design basis for end of life (E0L) fluence for JAFNPP is 32 EFPY based upon 40 years of power operation at 80 percent capacity factor. The predicted fluence for 32 EFPY was used to determine that vessel material i

properties meet the upper shelf energy (USE) requirements of 10 CFR Part 50.

i At the end of cycle 12, JAFNPP will have operated for less than 13.5 EFPY.

Future operation of FitzPatrick will not reach 32 EFPY, therefore previous EOL evaluations are still valid. The current estimate for E0L conservatively assumes 1.05 EFPY for each future year of operation (due to uprate),

i 100 percent capacity factor, and a 45-day refueling outage every two years.

l The NEDO-32205-A, Revision 1, is still applicable.

5 The staff finds the existing P-T curves valid and conservative for JAFNPP power uprate implementation. Ongoing plant specific analyses bound the i

effects of power uprate and thereby demonstrate compliance with 10 CFR Part 50, Appendix G.

3.18 Instrumentation and Control l

The TS changes proposed in the licensee's application (Reference 1) involve i

changes to the Reactor System trip and inteelock setpoints. These changes are intended to maintain the same margin between the new operating conditions and the new trip points as existed before the proposed power uprate.

i The staff's evaluation of setpoint changes associated with power uprate was limited to those setpoint changes for instrumentation identified in the licensee's submittals to the staff. The staff's review of the licensee's submittals indicates that NYPA conducted plant-specific calculations using i

methods recommended by the Instrument Society of America (ISA) as outlined in ISA-RP 67.04 " Methodologies for the Determination of Set Points for Nuclear i

Safety-Related Instrumentation" (Reference 11) within the limits stated in GE Topical Report NEDC-31336 (Reference 12).

l Thefollowingsetpointchangeshavebeenproposedbythelicenlee:

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1. Reactor Vessel Steam Dome Pressure Hiah (TS Table 3.1-1)

Change trip from 1045 psig to 1080 psig Change Allowable Value from 1120 psig to 1195 psig 4

2. Turbine Stoo Valve and Turbine Control Valve Fast Closure

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Scram Bvoass (TS 2.1.A.3 and 2.1.A.4) l The turbine first stage pressure setpoint was changed to reflect the expected pressure at the new 29 percent power point.

3. APRM Rod Block and APRM Simulated Thermal Power Hiah Clamos i

and APRM Neutron Flux Scram (TS 2.1.A.I.a, 2.1.A.1.b, 2.1.A.1.c)

These setpoints were not physically changed.

However, the change in power rating and resultant recalibration will result in an increase of approximately 100 MWt to each of these points.

i 1-5 To verify the results of the licensee-sponsored calculations and to better understand the quantitative effects of the assumed errors, the staff audited the calculations for the Turbine First Stage Pressure Scram Bypass setpoint and the Main Steam High Flow Isolation Trip setpoint. The licensee supplied the subject calculations by a [[letter::JPN-92-050, Forwards Response to RAI Re 920612 Application for Amend to License DPR-59 Re Power Uprate & Proposed Changes for GE NEDC-32016P Power Uprate Safety Analysis.Ge Power Uprate Safety Analysis Changes Withheld|letter dated September 17, 1992]] (Reference 13).

The staff review demonstrated that the instrumentation errors assumed in the analyses were conservative with respect to the manufacturer's ratings and the

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industries method of analysis.

The proposed setpoint changes resulting from the power uprate are intended to maintain the existing margins between operating conditions and reactor trip j

setpoints and do not significantly increase the likelihood of.a false trip nor failure to trip upon demand. Therefore, the existing licensing basis is not i

affected by the setpoint changes to accommodate the power uprate. The staff i

finds the setpoint changes, as described by the licensee's submittal, to be jl acceptable for power uprate.

3.19 Radioloaical Effects of Power Vorate i

To support the power uprate the licensee performed reanalyses of the l

Fitzpatrick UFSAR Chapter 14 accidents:

j

1. Large Break LOCA

2. Main Steam Line Break 1

3. Rod Drop i

4. Fuel Handling Accident For all these design basis accidents, the licensee calculated doses which meet the dose acceptance criteria of 10 CFR Part 100, the SRP, and GDC 19 of j

Appendix A to 10 CFR Part 50.

Large Break LOCA The licensee calculated the potential consequences of a postulated LOCA to the 3

control room operators and to individuals located at the EAB and LPZ using the guidance of SRP 15.6.5.

The sources of releases in the event of a LOCA include containment leakage, main steam isolation valve (MSIV) leakage and emergency core cooling system (ECCS) recirculation loop leakage. However, the licensee did not include the contribution of the MSIV leakage. The basis for their exclusion of the MSIV leakage was that it was treated by a leakage control system. Recirculation loop leakage was assumed to be released in the reactor building with credit assumed for removal of halogens by the SGTS.

The staff assessed the potential consequences of a LOCA based upon the assumptions presented in Table 1 (see Attachment). The staff included the MSIV leakage in the dose assessment because the SRP states that the MSIV leakage control system contributions are only excluded for systems which seal the leakage.

For systems which process the leakage, as Fitzpatrick does, the contribution from this pathway is included in the assessment.

The thyroid and whole body doses are presented in Tables 5 and 6 (see Attachment),

4 l 1 2

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respectively. The doses are within the dose acceptance criteria of 10 Part l

100, the SRP, and GDC 19 of Appendix A to 10 CFR part 50.

Main Steam Line Break 3

l The licensee reevaluated the consequences of a postulated main steam line i

break (MSLB) outside containment. The licensee performed their assessment using the guidance of SRP 15.6.4.

Previous licensee evaluations had limited i

the location of the steam line break to the steam tunnel.

For this evaluation i

the licensee included locations outside containment in addition to the locations inside the steam tunnel. Consequently, the licensee found that the most limiting break location occurred in a 16 inch bypass line leaking to the

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turbine bypass steam chest. The licensee indicated that the break of this j

line would result in more reactor coolant being released than if the 24 inch j

main steam line broke.

Rather than assuming that the release from a MSLB resulted in an instantaneous release, the licensee assumed that activity was released at a rate of 3 air changes per hour. The licensee calculated the consequences of the MSLB at a reactor coolant activity level of dose equivalent iodine of 0.2 gCi/g.

The UFSAR indicated that the licensee would j

place the control room in the pressurization mode with flow through the charcoal adsorber and HEPA filters whenever the coolant concentration of *I j

exceeds 0.01 pCi/g. The licensee indicated that 100 cfm of unfiltered inleakage could be assumed under these non-emergency conditions since the isolation procedure ensures that the appropriate valves and dampers are closed j

and there would be sufficient time to complete those follow-up actions.

The staff assessed the potential consequences of a MSLB based upon the 1

assumptions presented in Table 2 (see Attachment). This license amendment revises the limitations on RCS activity levels in the plant Technical i

Specifications. Therefore, the staff performed an assessment of the MSLB accident using these revised RCS activity levels consistent with the guidance of SRP 15.6.4.

The staff assessment resulted in the doses presented in Tables 5 and 6 (see Attachments).

The doses are within the dose acceptance criteria of 10 Part 100, the SRP, and GDC 19 of Appendix A to 10 CFR part 50.

Fuel Handling Accident The fuel handling accident scenario considered by the licensee was the drop of a fuel assembly from the maximum height allowed by the fuel handling equipment. The dropping of a spent fuel assembly onto the spent fuel pool was assumed to result in damage to fuel assemblies and the release of the volatile gaseous fission products to the spent fuel pool, followed by a release to the reactor building and then to the environment through the standby gas treatment system and main stack over a period of 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.

In their analysis, the licensee assumed that 99.8 percent of the radioactivity released to the reactor building would pass through the standby gas treatment system within the 2-hour period. No credit was assumed for decay due to either holdup in the reactor building or for transit time after release to the environment.

The licensee assumed that all releases through the standby gas treatment system removed with a filter efficiency of 99 percent for all forms of iodine.

_ _. _ _ _.. _ Table 3 (see attachec.1t) contains the assumptions utilized by the staff assessment of the potential consequences of a fuel handling accident. The staff assumed a more conservative filter efficiency of 90 percent for iodides.

The mitigation of the consequences of a fuel handling accident at Fitzpatrick is based upon safety grade radiological monitors isolating the containment in the event of a fuel handling accident.

It is assumed that the containment is being purged at a rate of 20,000 cfm, the containment is isolated approximately 41 seconds after the accident, and the release mixes with 20 percent of the reactor building volume. The doses are presented in Tables 5 and 6.

The doses are within the dose acceptance criteria of 10 Part 100,.the SRP, and GDC 19 of Appendix A to 10 CFR part 50.

Control Rod Drop The licensee performed an analysis of a postulated control rod drop accident consistent with the guidance of SRP 15.4.9.

The reactor scram and the MSIV closure functions of the main steam line radiation monitor were eliminated during the December 1994 outage; therefore, the potential exists for the release pathway of this accident to occur from the offgas system rather than the turbine / condenser if the MSIVs are not isolated.

Table 4 (see Attachment) presents the assumptions utilized in the staff assessment.The staff analyzed the consequences of a control rod drop accident using the guidance of SRP 15.6.4.9.

The staff assessment assumes that the MSIVs are isolated such that the release occurs from the turbine / condenser.

The potential consequences are presented in Tables 5 and 6 (see Attachment).

The doses are within the dose acceptance criteria of 10 Part 100, the SRP, and GDC 19 of Appendix A to 10 CFR part 50.

3.20 Human Factors The staff reviewed New York Power Authority's submittals dated June 12, 1992, and November 14, 1996, for power uprate. The staff's evaluation of the licensee's responses to five review topics is provided below.

Topic 1 - Discuss whether the power uprate will change the type and scope of plant emergency and abnormal operating procedures. Will the power uprate change the type, scope, and nature of operator actions needed for accident mitigation and will it require any new operator actions?

By letter dated November 14, 1996, the licensee stated that the power uprate would not change the type and scope of plant emergency and abnormal operating procedures. The licensee also stated that the power uprate would not change the type, scope, or nature of operator actions needed for accident mitigation and that it would not require any new operator actions.

The staff finds that the licensee's responses are satisfactory.

Topic 2 - Provide examples of operator actions potentially sensitive to power uprate and address whether the power uprate will have any effect on operator reliability or performance.

Identify operator actions that would necessitate reduced response times associated with a power uprate.

Please specify the

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i j !

l 1

expected response times before the power uprate and the reduced response j

times. What have simulator observations shown relative to operator response times for operator actions that are potentially sensitive to power uprate?

Please state why reduced operator response times are needed.

Please state whether the reduced time available to the operator as a result of the power uprate will significantly affect the operator's ability to complete manual i

actions in the times required.

By letter dated November 14, 1996, the licensee stated that the power uprate i

could decrease the time available for operator actions. Assuming a 5-percent power uprate and a reduction of 5 percent in the time available to operators, the licensee determined which operator actions would probably be sensitive to I

the power uprate. The licensee identified the following operator actions:

1 (1) failure to overrido main steam isolation valve (MSIV) isolation during an anticipated transient without scram (ATWS), (2) failure to initiate standby i

liquid control during an ATWS with MSIVs closed, (3) failure to maintain the level at the top of the active fuel during an ATWS with MSIVs closed, (4) failure to defeat high-pressure coolant injection auto transfer on high l

torus level during an ATWS with MSIVs closed, (5) failure to depressurize the reactor pressure vessel (RPV) during an intermediate loss-of-coolant accident (LOCA) with loss of high-pressure injection, (6) failure to align residual heat removal service water for injection during a small LOCA, and (7) failure to manually open low-pressure coolant injection valves. The licensee reported i

the following:

the percentage of core damage frequency change for the first j

six actions was less than 0.1 percent, the percentage of core damage frequency change for the seventh operator action was 0.65 percent, and the total

{

percentage of core damage frequency change was 0.65 percent.

a The licensee noted that the difference in simulator-derived times from pre-to i

post-uprate conditions (i.e., initial power of 2436 megawatt thermal and 2536 megawatt thermal, respectively) for RPV water level to lower from normal i

operating level to reach the top of the active fuel upon the loss of high-pressure injection systems is a reduction of about 3 percent. The licensee 1

also noted that for the ATWS event with MSIVs closed, the time for the torus i

to heat up from 79 degrees F to the boron injection initiation temperature (110 degrees F) changed from 101 seconds to 96 seconds. The licensee reported no significant change in the operator's ability to complete actions in response to normal operation, transients, or accidents during simulator training under simulated power uprate conditions. On the basis of previously discussed information, the staff finds that the reduction in the time available to the operator as a result of the power uprate is relatively small and it should not significantly affect the operator's ability to complete the j

subject manual actions.

i i

Topic 3 - Discuss any changes the power uprate will have on control room i

instruments, alarms, and displays. Are zone markings on meters changed (e.g.,

normal range, marginal range, and out-of-tolerance range)?

c By letter dated November 14, 1996, the licensee stated that there are seven i

new meters that have been added in the control room because the new operating points resulting from the power uprate are beyond the upper limits of the old j

l

2 1

meters. The licensee noted that a new scale will be added to the recorder for main steam flow because the upper limit for the recorder would not be j

consistent with the new steam flow because of the power uprate. Discussions J

with licensee personnel, after its November 14, 1996 letter, indicated that f

the new scale has been added to the recorder for main steam flow. The licensee indicated that there are several scales in the control room that have j

normal operating bands or trip setpoint indications that have been adjusted j

for the new normal and high RPV pressure trip setpoints. The staff finds that i

the licensee's responses regarding control room meters and scales that have i

been changed are satisfactory.

Topic 4 - Discuss any changes the power uprate will have on the Safety Parameter Display System (SPDS).

)

i By letter dated November 14, 1996, the licensee stated that the changes to the j

SPDS would be associated with setpoint changes. The licensee explained that the information presented on the top-level SPDS displays and the method of j

presentation would remain the same as before the power uprate. The licensee stated that changes to the lower level displays that support the SPDS include the following:

flags for safety relief valve lift pressure will be changed from 1,110 psig to 1,145 psig and a flag for high-pressure RPV scram vill be changed from 1,045 psig to 1,080 psig. Discussions with licensee personnel, j

after its November 14, 1996 letter, indicated that the previously discussed changes would be made prior to startup from the licensee's current refueling 1

outage.

Further the licensee indicated that changes will be made to three lower level limit displays that support the SPDS (i.e., the limit displays for i

boron injection initiation temperature, heat capacity temperature, and pressure suppression pressure and SRV tail pipe level). Discussions with i

licensee personnel, after its November 14, 1996 letter, indicated that the previously mentioned changes have been completed. The staff finds that the i

licensee's responses relative to SPDS changes that will be made or have been made are satisfactory.

3 Topic 5 - Describe any changes the power uprate will have on the operator training program and the plant simulator.

l By letter dated November 14, 1996, the licensee provided the following information:

(1) training content has been revised to include the power j

uprate modification; (2) simulator training on the power uprate modification has been conducted; (3) classroom training has been completed and involved an overview of various aspects of the power uprate modification, including parameter value changes and setpoint changes; and (4) plant system parameter 2

i changes have been implemented on the simulator, including changes made to j

control room instrumentation and the plant process computer system.

By letter dated November 14, 1996, the licensee committed to the following license j

condition (license paragraph 2.E.(3), Human Factors) associated with the i

proposed power uprate:

"The licensee will review the results of the startup test program to determine any potential effects on operator training.

Training issues identified will be incorporated in Licensed Operator training during 1997. Simulator discrepancies identified will be addressed in accordance with simulator Configuration Management procedural requirements."

_ The facility licensee bases portions of the rationale for the power uprate application on pre-startup operator training using the FitzPatrick simulation facility. The simulator has been modified to reflect uprate operating conditions using best estimate engineering data. Discussions with the facility licensee indicate that some post-modification testing was performed but the testing cannot readily be confirmed to have been consistent with the re-validation requirements of ANSI /ANS 3.5.

A post-modification test report was not prepared in accordance with the Standard because the facility licensee did not consider the modifications to be " major" or "significant."

l l

The facility licensee's position is contrary to previously provided regulatory i

guidance as to "significant." The term "significant" is defined in NUREG-1262 by operationally establishing the threshold as "...any change to the simulation facility, its models or software that might cause the results of 1

performance tests to fall outside the acceptable performance criteria set i

within the standard." The staff agrees with the facility licensee assessment J

that a need exists to incorporate actual performance data when it becomes available after startup, but further considers that the post-startup simulator update is required by the Standard to be a complete re-validation of the affected portions of the simulator.

The following simulator perspectives regarding facility licensee responses by letter dated November 14, 1996 to power uprate questions should be considered while preparing the simulator re-validation program:

Question 7 - Specific tests are required for the reference plant startup test program.

Similar tests will be required for the simulator re-validation.

Question 11 - SPDS changes in the reference plant have not been reflected in the simulator changes described by the facility licensee.

Question 12 - The facility licensee refers to the startup test program for the reference plant but does not indicate applicability of the startup test procedures for the simulator.

Question 15 - The facility licensee discusses " systems minimally impacted by the uprated power level." To the extent that these systems are already included in the scope of simulation, they should be included in the simulator re-validation. They should not be excluded from re-validation.

Question 17 - The facility licensee has evaluated the environmental effects of operation at the uprated power level on various equipment in the drywell, reactor building, turbine building, and steam tunnel.

These effects should be included in the simulator re-validation.

Discussions with the facility licensee indicated that simulator fidelity will be re-validated in accordance with ANSI /ANS 3.5-1985, Section 5.4.1, Simulator Performance Testing.

Simulator re-validation will include comparison of individual simulated systems and components and simulated integrated plant

i

. steady state and transient performance with reference plant response using similar sta'rtup test procedures.

On the basis of the information provided by the licensee and licensing condition, the staff concludes that the power uprate should not adversely affect operator performance or operator reliability.

3.21 Startuo Testina The licensee will follow a startup testing program as described in GE Licensing Topical Report NEDC-31897P-1, " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate" (Reference 3). The startup test program includes system testing of such process control systems as the 1

feedwater flow and main steam pressure control systems. The licensee will l

collect steady-state operational data during various portions of the power ascension to the higher licensed power level so that predicted equipment performance characteristics can be verified. The licensee will do the startup testing program in accordance with its procedures. At the staff's request, the licensee proposed a license condition to follow a startup test program as described in GE Licensing Topical Report NEDC-31897P-1, " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate" (Reference 3). The staff finds the licensee's approach in conformance with the test guidelines of Reference 3, and is therefore, acceptable.

i 3.22 Evaluation of Effect on Resoonses to Generic Communications. Plant i

Unioue Safety Evaluations. and Temoorary Modificationi In Reference 4 GE submitted a generic assessment of the effect of power uprate on responses to generic NRC and industry communications. The NRC staff approved the assessment in a letter dated July 31, 1992 (Reference 17).

Additionally, the licensee and GE performed a similar review of the effects of power uprate on plant-specific responses to NRC and Industry communications and performed an evaluation of the effects of power uprate on safety evaluations for work in progress and on temporary modifications. Any items found unacceptable were revised. The staff may audit these activities after plant startup following the implementation of power uprate. The staff finds this approach consistent with the NRC-approved generic guidelines in Reference 3 and is, therefore, acceptable.

3.23 Flow-Accelerated Corrosion (FAC) in Pinina Erosion / Corrosion (E/C also referred to as flow accelerated corrosion or FAC) in piping may be affected by the increased flow rates, higher operating temperatures, and change in moisture content of two-phase flow streams.

In Reference 15 the licensee presented it's evaluation of the impact of power i,

uprate on piping FAC. The licensee stated that the differences in operating conditions due to power uprate are small and should not cause a new FAC concern. NYPA has a formal FAC program for monitoring pipe wall thickness, i

and the licensee stated that any increase in the rate of pipe wall thinning i

that occurs as a result of the uprate will be identified through this program.

I.

The licensee uses industry experience, plant experience, and the Electric Power Research Institute (EPRI) CHECMATE/CHECWORKS model as the primary bases for NYPA FAC program. JAFNPP used the CHECMATE/CHECWORKS model to rank systems and components with respect to FAC susceptibility. The NRC staff requested that NYPA address whether JAFNPP will experience a higher steam moisture content as a result of power uprate and whether the higher flow (and, if applicable, higher moisture content) could change the FAC monitoring / susceptibility points as determined by the CHECMATE/CHECWORKS model. The licensee responded that the potential affects of increased pressure, temperatures, and flowrates on the FAC models were reviewed in Reference 16 as part of the power uprate program. These analyses conclude that, based on the relatively small increase in these parameters under uprate conditions, there will be a negligible effect on E/C wear rates.

The negligible increase in E/C rates will not affect any end-of-life predictions for components previously inspected. Highly susceptible FAC points will be inspected based on CHECMATE/CHECW9RKS analyses during the current Refuel 12/ Cycle 13 refueling outage, #.ch commenced on October 26, 1996. The inspection points were selectec' based on pre-uprate parameters.

The licensee stated that the FAC models will be updated and computer analyses will be l

performed to include the inspection data from the Refuel 12/ Cycle 13 FAC inspections. These updates are required as part of plant procedures and will be performed prior to the Refuel 13/ Cycle 14 refueling outage.

The small changes in parameters would not result in a change to the monitoring points.

l Therefore, NYPA does not expect power uprate to have an impact on the JAFNPP l

FAC program.

Because the operating conditions under power uprate will not affect the monitoring points for FAC, the expected increases in pipe wall wear rates are l

small, and the FAC program will continue to monitor for increases in pipe wall wear rates, the NRC staff finds that the effects of power uprate on FAC have been adequately considered and are not significant.

l 3.24 Hazardous Chemicals The NRC staff has reviewed the portion of the submittal dealing with the licensee's analysis of hazardous chemicals. The licensee identified several chemicals which were either stored on the plant site or transported in the i

vicinity of the plant.

Most of these chemicals did not need to be evaluated l

because they either were non-toxic, of low volatility or stored in sufficiently small quantities to pose any safety hazards.

Few chemicals which, when accidentally released, could pose safety hazard to the control l

room operators were evaluated by the licensee using the VAPOR computer code, developed by Stone and Webster.

The results of this analysis have indicated that, for the conditions existing at the plant site, none of these chemicals could produce conditions which would incapacitate control room operators. The

(

staff has reviewed the licensee's analysis and performed its independent verification using the NRC's HABIT computer code. On the basis of this i

I evaluation it concludes that the licensee's analysis is acceptable, i

l l

,m

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1 j l 3.25 Technical Specification (TS) Chanaes The analyses and evaluations supporting the proposed TS changes were completed using the guidelines in GE Topical Report NEDC-31897P-A " Generic Guidelines j

for General Electric Boiling water Reactor Power Uprate" (Reference 3). The 3

purpose of the proposed changes is to revise the TS to permit operation of the l

James A. FitzPatrick Nuclear Power Plant at an uprated power of 2,536 Mt.

i l

Engineering analyses and evaluations confirm that the plant can be operated safely at an uprated power. The increase in rated power remains below the plant design power level of 2,550 Nt which was the basis for the original l

plant safety evaluation.

i

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The following TS changes related to power uprate have been proposed by the l

licensee:

I

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Reactor Parameters:

(Pages 2, 5, 15, 16, 187, 134) The affect on reactor i

parameters is limited. Higher power is achieved by control rod pattern adjustments to increase reactor thermal power in a more uniform (flattened) power distribution to increase steam flow without increasing core j

recirculation flow. This requires an increased reactor dome pressure for adequate turbine inlet pressure.

Operational Limits: The increased thermal power requires a change to the limitation on operation in the high power, low flow portion of the power / flow map to limit thermal hydraulic instabilities and power oscillations.

Setpoints:

(Pages 11, 19, 27, 29, 34, 41a, 42) The increased reactor pressure had a direct impact on the high pressure scram setpoint and the safety relief j

valve setpoint. Additionally, the bypass for the turbine stop valve closure and control valve fast closure scram was changed in proportion to the increase in thermal power.

4 i

Analysis Results:

(Pages 139, 149, 188, 193) Analyses of uprated power transients and accidents required changes to various technical specifications and their bases. Operational parameters and assumptions used in analyses were revised to reflect their use as initial conditions.

Revised radiological analyses changed dose results.

The results of the accident analyses required 4

revisions to properly reflect plant capabilities.

t l

Testing:

(Pages 117, 121a, 147, 172) A number of changes to testing requirements resulted from power uprate.

The increase to reactor pressure had a direct effect on hydrostatic leakage testing pressure.

The test pressure for HPCI and RCIC pumps was revised to reflect SRV setpoints assumed in analyses.

1 Administrative:

(Pages 6a, 14, 20, 29, 35, 41a, 100, 102, 125, 152, 166, 188, 193, 254c, 285, 285a) Administrative changes (i.e. adding references, revising references, correcting associated errors and editorial changes) were also i

made.

i

l l Paae 2 Specification 1.0.D - Definition of Hot Standby Condition Replace the value "1,005 psig" with the value "1,040 psig."

The change revises the hot standby condition to reflect the operating pressure

(

of the reactor at uprated power conditions.

Paae 5 Specification 1.0.N-Definition of Rated Power Replace the value "2436 MWt" with the value "2536 MWt" and add "(Reference 1)"

i to the end of the sentence.

The change revises the definition of rated power to reflect the increased thermal power at uprated power conditions and provides a reference to the safety evaluation submitted in support of power uprate.

Pace 6a Specification AE - References Add a new specification that reads as follows "AE.

References 1.

General Electric Report NEDC-32016P-1, " Power Uprate Safety Analysis for the James A. FitzPatrick Nuclear Power Plant,"

April 1993 (proprietary), including Errata and Addenda Sheet No. 1, dated January 1994."

The change adds the current revision of the Power Uprate Safety Analysis Report (NEDC-32016P-1, Reference 1). This change supersedes the change originally proposed in the licensee's submittal dated June 12, 1992.

Paae 11 Specification 2.1.A.3 - Turbine Stop Valve Closure Scram Trip Setting.

Replace "above 217 psig turbine first stage pressure" with the phrase "the reactor is at or above 29% of rated power" The change replaces the turbine stop valve closure scram bypass setpoint pressure with a reference power level.

Pace 14 Bases Section 1.1.E.1 - References Replace the following:

"1.

' General Electric Standard Application for Reactor Fuel,' NEDE-24011-P, l

latest approved revision and amendments."

With:

"1.

' General Electric Standard Application for Reactor Fuel,' NEDf-24011-P-A-13, August 1996."

_ _ _ _ _ _ _ _. This change adds the current revision of the General Electric Standard Application for Reactor Fuel. This change is necessary to reflect the change proposed to TS page 254c.

Pace 15 Bases Section 2.1 - Fuel Cladding Integrity In the first paragraph replace the value "2,436 Mt" with the value "2,536 MWt" in two locations.

In the first sentence of the second paragraph, replace the word "given" with the word " described" At the end of the second paragraph, add the following two sentences:

" Reference 1 evaluates the safety significance of uprated power operation at 2,536 MWt. This evaluation is consistent with and demonstrates the acceptability of the transient analyses required by Reference 2."

The changes revise the Bases to reflect the increased rated thermal power at uprated power conditions and the associated supporting references. The first change identifies the new thermal power level and the reference, added on TS page 20, is the Power Uprate Safety Analysis Report (PUSAR).

Paae 16 Bases Section 2.1 - Licensed Maximum Power Level Replace the value "2,436 MWt" with the value "2,536 MWt."

The change revises the Bases to reflect the new maximum'11 censed power level at uprated power conditions.

Paae 19 Bases Section 2.1.A.3 - Turbine Stop Valve Closure Scram Trip Setting In the last sentence of Section 2.1.A.3, replace the phrase, " turbine steam flow is below 30%" with the phrase " reactor power is below 29%."

In the last sentence of Section 2.1.A.3, add the phrase " consistent with the safety analysis discussed in Reference 1" at the end of the sentence.

The changes revise the Bases to reflect the change proposed to TS 2.1.A.3 and 3.1.A (Table 3.1-1, footnote 4) to the value at which the turbine stop valve closure scram is bypassed and provide a reference to the PUSAR to identify a safety discussion of the supporting analyses.

l Paae 20 Bases Section 2.1.C.1 - References Replace "1.

Deleted" with "1.

General Electric Report, NEDC-32018P-1, ' Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant,' April 1993 (proprietary), including Errata and Addenda Sheet No. 1, dated January 1994."

This change adds the current revision of the PUSAR (NEDC-32016P-1, Reference 1).

Bases Section 2.1.C.2 - References i

Replace:

"2.

' General Electric Standard Application for Reactor Fuel', NEDE-24011-P-A (Approved revision number applicable at time that reload fuel analyses l

are performed)."

with:

"2.

' General Electric Standard Application for Reactor Fuel', NEDE-24011-P-A-13, August 1996."

l i

This change adds the current revision of the General Electric Standard Application for Reactor Fuel. This change is necessary to reflect the change proposed on TS page 254c, Section 6.9.(A).4.t>.1.

i Paae 27-Technical Specification 2.2.1.A - Reactor Coolant System Replace the value "1,045 psig" with the value "1,080 psig."

The reactor high pressure scram setpoint was revised to reflect changes in plant operating conditions during power uprate. The current reactor high pressure scram limiting safety system setting of 1,045 psig is increased by 35 psig to 1,080 psig to reflect the 35 psig increase in the steam dome during operation.

Technical Specification 2.2.1.B - Reactor Coolant System Replace the value "1,110 psig" with the value "1,145 psig."

l The revised reactor safety / relief valve (SRV) setpoints reflect changes in the plant operating conditions during power uprate.

j Pace 29 l

Replace the following change originally proposed in submittal dated June 12,

(

1992, and superseded by the change proposed in submittal dated August 15, 1996:

l l

"...NEDC-31697P, ' Updated SRV Performance Requirements for the JAFNPP,'

assuming 9 of 11 SRVs were operable with opening pressures less than or equal to 1195 psig. The resultant peak vessel pressure for the event was shown to be less than the vessel pressure code limit of 1375 psig.

(See current reload analysis for the reactor response to the main steam isolation valve closure with flux scram event).

The value of 1195 psig is the SRV opening pressure up to which plant performance has been analyzed, assuming 2 SRVs are inoperable.

Therefore, SRV opening pressures below 1195 psig ensure that the ASME Code limit on peak reactor pressure is satisfied..."

with:

"...NEDC-32016P-1, ' Power Uprate Safety Analysis for James A. FitzPatrick Nuclear Power Plant,' Including Errata and Addenda Sheet No.1, dated January 1994, assuming 9 of the 11 SRVs were operable with opening pressures less than or equal to 1179 psig. The resultant peak vessel pressure for the event was shown to be less than the ASME Code limit of 1375 psig (see current reload analysis for the reactor response to the main steam isolation valve closure with flux scram event)..."

This change adds the current revision of the PUSAR (NEDC-32016P-1).

Pace 34 Bases 3.1 - Turbine Stop Valve Closure Scram Trip Setting Replace "217 psig turbine first stage pressure (30 percent of rated)" with "29% of rat a reactor power" in the second sentence of the last paragraph.

The change revises the Bases to reflect the change to Table 3.1-1 of TS 3.1.A.

Paae 35 Bases Sections 3.1.B, 3.1.C.1, and 3.1.C.2 1.

Replace the following in Bases Section 3.1.B:

...NED0-21662 (Reference 1) and NEDC-31317P (Reference 2) including the latest revision, errata and addenda..."

with:

... Reference 1..."

2.

Replace the following from Bases Section 3.1.C:

"1.

General Electric Topical Report NED0-21662, Revision 2, ' Loss-of-Coolant Accident Analysis Report for James A. FitzPatrick Nuclear Power Plant (Lead Plant)', July 1977 with errata and i

addenda.

2.

General Electric Topical Report NEDC-31317P, ' James A.

FitzPatrick Nuclear Power Plant SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis,' October 1986 with revisions, errata and addenda."

(

with:

i I

__ _ _. _. _ _. ___.._. _..__ _ _._ _. _ _ _ "1.

' James A. FitzPatrick Nuclear Power Plant SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis,' NEDC-31317P, Revision 2, April

• 1993."

This change deletes reference to an outdated LOCA analysis report for FitzPatrick (NED0-21662).

In addition, this change provides the current l

reference LOCA analysis report.

j Paae 41 and Paae 43 (Previously 41a. 42)

Table 3.1-1 and its associated notes were significantly altered subsequent to the submittal on June 12, 1992. The submittal specified the correct setpoints for power uprate (technical changes) and moved a table entry form page 42 to l

41 (editorial change). The revised pages in the submittal of November 20, l

1996 retain the required technical changes while making no editorial changes.

The Trip Level Setting for the Reactor High Pressure Scram is changed from "s 1045 psig" to "s 1080 psig". The change revises the reactor high pressure j

scram setpoint to reflect the reactor operating pressure.

In note 3.C 1

"30 percent" with "29 percent." Note 5 is reworded from " Bypassed when turbine first stage pressure is less than 217 psig or less than 30 percent of rated power" to " Bypassed when reactor power is less than 29 percent of rated power." The changes provide the revised value at which the turbine stop and turbine control valves can have their closure / fast valve closure scrams bypassed.

Paae 100 Section 3.3 and 4.3 of the Bases Item B.3 Repl ace, "...NEDE-24011... "

with, "...NEDE-24011-P-A-13, August li16..."

This change adds the current revision of the General Electric Standard Application for Reactor Fuel. This change reflects the change to TS page 254c, Section 6.9.(A).4.b.l.

Paae 102 L

Section 3.3 and 4.3 of the Bases Item C.

l Repl ace, "...NEDE-24011-P-A... "

with, "...NEDE-24011-P-A-13, August 1996..."

This change adds the current revision of the General Electric Standard Application for Reactor Fuel. This change reflects the change to TS page l

254c, Section 6.9.(A).4.b.1.

i

__ l Paae 117 Specification 4.5.C.1 - HPCI System Surveillance Test Pressure Replace the value "1,120 psig" with the value "1,195 psig."

The change revises the HPCI test pressure to reflect the analyzed value at which the SRV could be set. HPCI must be able to deliver water to the primary system at the highest pressure allowed by the SRV.

Paae 121a Specification 4.5.E.1 - RCIC System Surveillance Test Pressure Replace the value "1,120 psig" with the value "1,195 psig."

The change revises the RCIC test pressure to reflect the analyzed value at which the SRV could be set.

RCIC must be able to deliver water to the primary system at the highest pressure allowed by the SRV.

I Paae 125 Bases Section 3.5.A Replace the following, "...NEDE-240ll-P-A..."

with, "...NEDE-240ll-P-A-13, August 1996..."

This change adds the current revision of the General Electric Standard Application for Reactor Fuel. This change reflects the change to TS page 254c, Section 6.9.(A).4.b.1.

Paae 139 Specification 3.6.C.1 - Coolant Chemistry l

Replace "3.1 pCi/gm" with the value "0.2 pCi/gm" The change reduces the reactor coolant system radioactivity operating limit to be consistent with new accident and transient analyses performed at power uprate conditions. Subsequent to the submittal of June 12, 1992, proposed changes to page 139, amendment 190 replaced abbreviations for ' hour (s)' with the full word and amendment 199 relocated specification 4.6.A.7 requirements for the reactor vessel flux monitoring surveillance program from the TS to a NYPA controlled document.

This revised page retains these previously approved changes.

Paae 142a l

Specification 4.6.E - Safety / Relief Valves i

Replace "1110 psig" in two places in specification 4.6.E with "1145 psig." At

-the time of the June 12, 1996 submittal, TS 4.6.E did not explicitly state the SRV lifting setpoint, rather it provided the setpoints by reference to TS i

{

- ~ _ -.. - __. - - - - - - - -

4 i 4 j

i 2.2.B.

The revised page maintains the explicit statement of SRV setpoint j

which was approved in amendment 217 to the TS, issued September 28, 1994.

j Paae 147 i

Bases 3.6 and 4.6 - Maximum Hydrostatic Test Pressure Replace "1,105 psig" with "1,144 psig" The change increases the peak hydrostatic test pressure to reflect the increased reactor operating pressure. The ASME code allows hydrostatic testing to 1.1 times the operating pressure. The power uprate increase in the operating pressure by 35 psig to 1,040 psig results in a 39 psig increase in i,

allowable test pressure to 1,144 psig. An additional editorial change is made j

to correct an error introduced in amendment 199. Replace the last sentence on the page, "Accordingly, the maximum hydrostatic test pressure will be 1.1 l

times the operating pressure of about 1105 psig" with "Accordingly, the maximum hydrostatic test pressure will be 1.1 times the operating pressure or about 1144 psig."

Paae 149 5

Bases 3.6.C and 4.6.C - Coolant Chemistry and Dose Analysis Replace the sentence, "In the event of a steam line rupture outside the drywell, with this coolant activity level, the resultant radiological dose at l

the site boundary would be 33 rem to the thyroid, under adverse meteorological conditions assuming no more than 3.1 Ci/gm of dose equivalent I-131" is i

replaced with "In the event of a steam line rupture outside the drywell, a more restrictive coolant activity level of 0.2 Ci/gm of dose equivalent I-131 was assumed. With this coolant activity level and adverse meteorological conditions, the calculated radiological dor,e at the site boundary would be less than 30 rem to the thyroid."

d i

The change revises the Bases to make it consistent with the revised dose analysis for main steam line break. This analysis was performed using a revised reactor coolaat specific activity.

The revised specific activity is now smaller (3.1 #Ci/gm changes to 0.2 #C1/gm) than the specific activity allowed by the Radiological Effluent Technical Specification limit.

l Paae 152 Section 3.6 and 4.6 of the Bases Item E.

Replace:

1 3

...The safety bases for a single nominal valve opening pressure of 1110 psig i

are described in NEDC-31397P, ' Updated SRV Performance Requirements for the JAFNPP.' The single nominal setpoint is set below the reactor vessel design pressure (1250 psig) per the requirements of Article 9 of the ASME Code -

Section III, Nuclear Vessels. The setting of 1110 psig..."

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i with:

1 "The safety bases for a single nominal valve opening pressure of 1145 psig are described in NEDC-32016P-1, ' Power Uprate Safety Analysis for James A.

FitzPatrick Nuclear Power Plant,' Including Errata and Addenda Sheet No.1, 1

i dated January 1994. The single nominal setpoint is set below the reactor vessel design pressure (1250 psig) per the requirements of Article 9 of the j

ASME Code - Section III, Nuclear Vessels.

The setting of 1145 psig..."

l Replace, "...The analyses with NEDC-31697P also..."

with, "...The analyses with NEDC-32016P-1, including Errata and Addenda Sheet No. 1, dated January 1994, also..."

i These changes adds the current revision of the PUSAR (NEDC-32016P-1) and adds a revision bar which was omitted in the licensee submittal dated August 15, J

1996.

Paae 166

{

Specification Section 4.7.2.c. - Containment Leakage Test Pressure l

Replace, "...hydrostatically tested at 2 1000 psig..."

l with, "...hydrostatically tested at 2 1035 psig..."

i The change revises the leakage testing criteria to reflect the new operating pressure.

i i

Paae 187 1

Bases 3.7 - Suppression Chamber Blowdown In the second paragraph replace the value "1,020 psig" with the value "1,040" psig."

The change revises the Bases section to reflect the increase operating pressure.

Pace 188 Bases 3.7 - Torus Water Volume 3

In the firs aragraph, replace the value "105,600 ft " with the value "105,900 ft The change revises the discussion in the Bases of the suppression chamber water volume to reflect the uprated power containment analyses.

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Bases 3.7 - Primary Containment In the second paragraph delete the phrase " complete condensation of the limit i

for".

In the fifth paragraph replace the value "130 'F" with the value "105 *F." Also, in the fifth paragraph replace the word " form" with the wori i

"from."

1 The changes correct typographical errors and revises the value of the limiting suppression pool temperature. The value of "130 *F" was part of the original TS Bases for restricting the temperature rise in the torus pool during the use i

of the RCIC, HPCI or relief valves.

It reflects the condensation limit for blowdown of 170 'F, based on the Humboldt Bay and Bodega Bay tests, less the 40 *F pool temperature rise associated with blowdown. Amendment 16 changed this value to "105 'F."

The value was inadvertently changed in amendment 36.

This change corrects this error.

)

1 Bases 3.7 - Suppression Chamber Water Temperature Replace:

"Using a 40 *F rise (Section 5.2 FSAR) in the suppression chamber water temperature and a maximum initial temperature of 95 "F, a temperature of 3

145 *F is achieved, which is well below the 170 'F temperature which is 4

i used for complete condensation."

with:

" Containment analyses predict a 46 'F increase in pool water temperature, after complete LOCA blowdown. These analyses assumed an initial suppression pool water temperature of 95 *F and rated reactor power of 2536 MWt.

LOCA analyses in Section 14.6 of the FSAR also assume an initial 95 'F pool temperature. Therefore, complete condensation is assured during a LOCA because the maximum pool temperature (141 *F) is less than the 170 *F temperature seen during the Bodega Bay tests."

This change revises the discussion in the Bases of the calculated temperature rise in the suppression chamber based upon the uprated power analyses and suppression pool temperature Limiting Conditions for Operation.

Bases 3.7 - ECCS Pump NPSH Replace the fourth paragraph which says:

"For an initial maximum suppression chamber water temperature of 95 "F and assuming the normal complement of containment cooling pumps (two LPCI pumps and two RHR service water pumps) containment pressure is not required to maintain adequate net positive suction head (NPSH) for the core spray LPCI and HPCI pumps."

with:

1

- _ - _ - - - "For an initial maximum torus water temperature of 95 'F, assuming the worst case complement of containment cooling pumps (one LPCI pump and two RHR service water pumps), containment pressure is required to maintain adequate net positive suction head (NPSH) for the core spray and LPCI pumps.

This change revises the discussion in the Bases of the NPSH requirements for the ECCS pumps. The design basis and analyses do not consider two trains to be available and the Bases should reflect the design condition.

Bases 3.7 - Torus Temperature Limits

' Delete the last paragraph and replace with:

" Experiments indicate that unacceptably high dynamic containment loads may result from unstable condensation when suppression pool water temperatures are high near SRV discharges. Action statements limit the maximum pool temperature to assure stable condensation. These actions include:

limiting the maximum pool temperature of 95 *F during normal operation; initiating a reactor scram if during a transient (such as a stuck open SRV) pool temperature exceeds 110 'F; and depressurizing the reactor if pool temperature exceeds 120 'F.

T-quenchers diffuse steam discharged from SRVs and promote stable condensation. The presence of T-quenchers and compliance with these action statements assure that stable condensation will occur and containment loads will be acceptable.

Paae 188a Bases 3.7 - Torus Temperature Limits The following paragraph was moved from the proposed page 188 in the licensees submittal dated June 12, 1992 to page 188a:

"NEDC-24361P (August 1981) summarizes analyses performed to predict pool temperatures and containment loads during plant transients using these temperature limits at a power level of 2535 MWt (104% of rated). NEDC-2435P also substantiates the acceptability of the plant design using the local pool limits of NUREG-0661. NED0-30832 (December 1984) shows that SRV condensation loads are low compared to other design loads for plants with T-quenchers. NEDO-30832 describes why local pool temperatures need not be analyzed at a rated power level of 2536 MWt."

The above two paragraphs added to Bases 3.7 - Torus Temperature Limits, eliminates the discussion in the Bases of the peak suppression pool temperature limit of 160 'F used to avoid excessive loads due to steam condensing during blowdown to the torus. This temperature limit was adopted by utilities in 1974 when the phenomenon was initially defined.

The power uprate evaluation has identified plant specific and generic analyses that supersede this temperature limit and establish new justification for the torus

l i i I

I temperature liinits. The purpose of the change is to reconcile these assessments and their relation to the torus temperature limits.

Paae 193 Bases 4.7.A - LOCA Dose Analysis

]

Replace:

The design basis loss-of-coolant accident was evaluated in FSAR section 14.6 incorporating the primary containment... Thus, these doses are the maximum that would be expected in the unlikely event of a design basis loss-of-coolant accident."

with:

j Design basis accidents were evaluated as discussed in Section 14.6 of the FSAR and the power uprate safety evaluation, Reference 18. The whole 4

body and thyroid doses in the control room, low population zone (LPZ) and j

site boundary meet the requirements of 10CFR Part 50 and 100. The technical support center (TSC), not designed to these licensing bases, was also analyzed. The whole body and thyroid dose acceptance criteria used for the main control room are met for the TSC when initial access to the TSC and occupancy of certain areas in the TSC is restricted by administrative control. The LOCA dose evaluations, References 19, 20,

'i and 21, assumed:

the primary containment leak rate was 1.5 volume percent per day; source term releases were in accordance with TID-14844 and Regulatory Guide 1.3, and were consistent with the Standard Review Plan; and the standby gas treatment system filter efficiency was 99% for halogens."

The change eliminates the discussion in the Bases of the specific dose results from the LOCA dose calculation. A change was necessary because the dose i

analysis at uprated power changed the calculation models and results. A i

reference to the uprate power safety evaluation was also necessary. The change discusses the assumptions used in the dose calculations in the same level of detail for consistency. The results of the dose calculations are discussed generally since it is necessary to show compliance with the acceptance criteria and not calculated results. The results of the dose analysis will be included in an FSAR update following power uprate approval.

Paae 254c Administrative Controls Section 6.9.(A)4.b.

The changes provide specific date and version of references for the analytical methods used to determine core operating limits as follows:

1. " General Electric Standard Application for Reactor Fuel," NEDE-24011-t P-A-13, August, 1996.

I

2. " James A. FitzPatrick Nuclear Power Plant SAFER /GESTR-LOCA Loss-of-Coolant Accident Analysis," NEDC-31317P, Revision 2, April 1993.

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' 3. "BWR Owners' Group Long-term Stability Solutions Licensing Methodo1ogy," NED0-31960-A, June 1991.

4. "BWR Owners' Group Long-term Stability Solutions Licensing Methodology," NED0-31960-A, Supplemental 1, March 1992.

The change also deletes a reference that is no longer applicable. The changes ensure that values for cycle specific parameters are determined consistent with applicable design and safety analysis.

Paae 285 and 'Jin Section 7.0 - References The changes of this page add references that were used in making changes to l

other TS pages.

The TS changes reflect the necessary changes to implement the uprated power level to 2536 MWt which is now the rated 100 percent power. The higher i

operating pressure and slightly higher operating temperature necessitated the above TS or TS set point changes. The staff's review of the proposed TS l

changes indicate that they reflect the revised operating conditions and have been accounted for in the review of the safety analyses, thus, they are acceptable.

Paae 3 of Licerlig 2.C.(1) Maximum Power Level The change reflects the uprated Maximum Power Level. The change replaces 2436 megawatts (thermal) with 2536 megawatts (thermal).

Pace 4 of Licente_

2.E. Power Uprate License Amendment Implementation The addition of license conditions to support power uprate monitoring and testing.

l

4.0 STATE CONSULTATION

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

The State official had no comment.

5.0 ENVIRONMENTAL CONSIDERATION

Pursuant to 10 CFR 51.21, 51.32, and 51.35, an Environmental Assessment and Finding of No Significant Impact have previously been prepared and published in the Federal Register on December 2, 1996 (61 FR 64004). Accordingly, based I

upon the environmental assesspent, the Commission has determined that the

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issuance of this amendment will not have a significant effect on the quality of the human environment.

1 6.0. CONCLUSION The Commission 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 l

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.

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7.0 REFERENCES

1.

Letter from R.E. Beedle, NYPA, to NRC, " Proposed Changes to the Technical Specifications Regarding Power Uprate (JPTS-91-025)," June 12, 1992.

2.

Letter from William T. Russell, NRC, to Patrick W. Marriott, GE, " Staff Position Concerning General Electric Boiling Water Reactor Power Uprate Program," September 30, 1991.

3.

GE Licensing Topical Report NEDC-31897P-1, " Generic Guidelines for General Electric Boiling Water Reactor (BWR) Power Uprate," June 1991.

(Proprietary information. Not publicly available) and GE Nuclear Energy,

" Generic Guidelines for General Electric Boiling Water Reactor Power Uprate," Licensing Topical Report NED0-31897, Class I (nonproprietary),

l February 1992; and NEDC-31897P-A, Class III (proprietary), May 1992.

4.

GE Licensing Topical Report NEDC-31984P, " Generic Evaluations of General Electric Boiling Water Reactor Power Uprate," Class III (proprietary),

July 1991, NED0-31984, Class I (nonproprietary), March 1992; and l

supplements 1 and 2.

5.

GE Nuclear Energy, " Power Uprate Safety Analysis for the James A.

FitzPatrick Nuclear Power Plant," Licensing Topical Report NEDC-32016P-1, Class III (proprietary), Revision 1, April 1993.

6.

GE Nuclear Energy, " James A. FitzPatrick Nuclear Power Plant SAFER /GESTER-LOCA, Loss-of-Coolant Accident Analysis," Licensing Topical Report NEDC-31317P, Class III (proprietary), Revision 2, April 1993.

7.

Letter from NYPA, to NRC, JPN-92-089, " Response to Request for Additional Information-Power Uprate Submittal," December 29, 1993.

8.

Letter, from R.L. Gridley, GE, to D.G. Eisenhut, NRC, " Review of Low Core Flow Effects f n LOCA Analysis for Operating BWRs, Revision 2," May 28, 1978.

i 9.

Letter, from D.G. Eisenhut, NRC, to R.L. Gridley, GE, " Safety Evaluation Report on Revision of Previously Imposed MAPLHGR (ECCS-LOCA) Restriction for BWRs at Less Than Rated Core flow," May 19, 1978.

10. General Electric Report No. GE-NE-B1301805-05R1, " Reactor Vessel Fracture Toughness Engineering Evaluation for the James A. FitzPatrick 104% Power Uprate," dated March 1996 (proprietary).

11.

ISA-RP 67.04, Draft 7, " Methodologies for the determination of Setpoints for Nuclear Safety-Related Instrumentation," September 14, 1990.

l

12. GE Topical Report NEDC-31336P, " General Electric Instrument Setpoint Methodology," October 1986 (Proprietary information. Not publicly available).

l

, 13.

Letter, from R.E. Beedle, NYPA, to NRC, " Response to Request for Additional Information Regarding Proposed Technical Specification Change l

for Power Uprate," September 17, 1992,

14. GE Engineering Report, NE-187-59-1191P, " Anticipated Transients Without Scram (ATWS) Analyses for the James A. FitzPatrick Nuclear Power Plant,"

November 1991.

l 15.

Letter, from W.J. Cahill, Jr., to NRC, " Response to Request for Additional Information Regarding Power Uprate," November 14, 1996.

I 16.

Stone & Webster Engineering Corporation, " Core Power Uprate Engineering Report for James A. FitzPatrick Nuclear Power Plant," December 1991.

17.

W.T. Russell, U.S. Nuclear Regulatory Commission, letter to P.W.

Marriott, General Electric Company, " Staff Safety Evaluation of General i

Electric Boiling Water Reactor Power Uprate Generic Analysis," July 31, 1992.

Attachment:

Tables Principal Contributors:

J. Raval R. Emch J. Hayes F. Collins G. Carpenter B. Elliot N. Trehan G. Hubbard M. Sykes R. Goel B. Marcus R. Scholl, Jr.

G. West C. Wu B. Whitacre L. Lois C. Gratton Date:

December 6, 1996 l

l l

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Attachment Table 1 Assumptions for LOCA Analysis l

Core Thermal Power (MWt) 2587 l

Activity Released to the Reactor i

Building Airborne (fraction of core) l Iodine 0.5 Noble Gases 1.0 t

Iodine Plateout Factor 0.5 l

Iodine Species (fraction) l Elemental 0.91 Particulate 0.05 Organic 0.04 Activity Released to Sump (fraction) l l

Iodine t

0.5 l

Noble Gases 0.0 1

Suppression Pool Scrubbing DF Particulate 5

Elemental 5

Reactor Building Leakage Rate (%/ day) 0.5 Sump Liquid Mass (lbs) 1.13E5 Standby Gas Treatment Adsorber 90 i

4 Efficiency (%)

Recirculation Loop Leakage Rate (gpm) 5 l

Minimum Time to Recirculation (sec) 600 l

i

- _. = _

Table 1 Assumptions for LOCA Analysis (continued)

Control Room 3

Free Volume (ft )

1.49E5 Makeup Air Filtration Rate (cfm) 900 Makeup Air Treatment Adsorber 90 l

Efficiency (%)

Isolation Time (min) 30 i

Unfiltered Air Infiltration Rate (cfm)

Prior to isolation 15,010 Following isolation 2,110 Occupancy Factors 0-1 day 1.0 l

1-4 days 0.6 i

4-30 days 0.4 3

Atmospheric Dispersion Factors (sec/m )

EAB 5.18E-5 LPZ 0-4 hours 2.6E-5 4-8 hours 6.2E-6 8-24 hours 4.7E-6 1-4 days 1.7E-6 4-30 days 5.8E-7 Control Room 0-2 hours

• 8.1E-5 3

Breathing Rates (m /sec)

Offsite 0-8 hours 3.47E-4 8-24 hours 1.75E-4 1-30 days 2.32E-4 Control Room 3.47E-4

• Calculated control room doses beyond first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> were not significant compared to first 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> t

l l

Table 2 Assumptions for Main Steam Line Break Accident t

Steam and H O Released 2

Initially steam only (lbs) 1.16E4 Water and steam mixture 9.37E4 Time to Isolate MSIVs (sec) 10.5 Reactor Coolant Activity Level Equilibrium V,alue Dose 0.2 Equivalent '3 I (gCi/g)

MaximumIodineAgivityof 2.0 Dose Equivalent I (gCi/g) 3 Atmospheric Dispersion Factors (sec/m )

l EAB 1.79E-4 I

LPZ 0-8 hours 2.00E-5 8-24 hours 1.34E-5 1-4 days 4.27E-6 4-30 days 1.60E-6 Control Room 0-8 hours 3.29E-3 8-24 hours 2.81E-3 1-4 days 2.00E-3 4-30 days 1.22E-3 Control Room 3

Free Volume (ft )

1.49E5 Makeup Air Filtration Rate 900 (cfm)

Makeup Air Treatment Adsorber 90 l

Efficiency (%)

Isolation Time (min) 0 Unfiltered Air Infiltration 100 Rate (cfm) l i

f

~ -. _ _.

Table 3 Assumptions for Fuel Handling Accidents

(

Core Power (MWt) 2587 Number of fuel rods in full core 35,784 Number of damaged fuel rods 125 Highest Power Discharged Assembly i

Axial Peak to Average Ratio 1.5 Radial Peak to Average Ratio 1.5 Occurrence of Accident (hours after shutdown) 24 Activity released from the gap as Noble gases except Kr 0.10 asKr 0.30 gdineexcept*I 0.10 I

0.12 Iodine Gap Inventory organic (%)

0.25 inorganic (%)

99.75 Fuel Pool Decontamination Factor organic (%)

1 inorganic (%)

133 Standby Gas Treatment Adsorber Efficiency (%)

90 l

.-. -.. -. - -... _. ~ - -.. =. -. -

e i

d Table 3 Assumptions for Fuel Handling Accidents (Continued) 3 Atmospheric Dispersion Factors (sec/m )

l EAB 5.18E-5 l

LPZ 2.6E-5 i

i Control Room 8.lE-5 J

4 F

i

Table 4 Assumptions for Control Rod Drop Accident Core Thermal Power (MWt) 2587 Number of Fuel Rods Expected to Fail 850 Peaking Factor 1.5 Fraction of Failed Fuel which reaches the turbine and condenser from the reactor vessel (%)

i Noble gases 100 t

Halogens 10 i

Plateout Fraction of Halogens in 0.9 Turbine / Condenser Leak Rate for Condenser (%/ day)

I l

Control Room 3

Free Volume '(ft )

1.49ES l

Makeup Air Filtration Rate 900 (cfm)

Makeup Air Treatment Adsorber 90 Efficiency (%)

Isolation Time (min) 30 Unfiltered Air Infiltration Rate (cfm)

Prior to isolation 15,010 Following isolation 2,110 Occupancy Factors 0-1 day 1.0 1-4 days 0.6 4-30 days 0.4

... ~. -... - _. -.. -.... -.- -..-_.

Table 4 Assumptions for Control Rod Drop Accident (Continued) 3 Atmospheric Dispersion Factors (sec/m )

EAB 1.79E-4' LPZ 0-8 hours 2.00E-5 8-24 hours 1.34E-5 Control Room 0-8 hours 3.29E-3 8-24 hours 2.81E-3

(

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Table 5 Thyroid Doses from Postulated Accidents (Rem)

Accident EAR LPl 1.

Large Break LOCA MSIV and Containment Leakage 7

14 ECCS Leakage 6

15 Total 13 29 Control Room MSIV & Containment 7

ECCS Leakage 7

Total 14-2.

Main Steam Line Break 0 Equilibrium TS Value 0.7 0.08 010 times Equilibrium TS Value 7.

0.8 Control Room 9 Equilibrium TS Value 3

9 10 times Equilibrium 30 TS Value 3.

Control Rod Drop 0.2 0.09 Control Room 19 4.

Fuel Handling 0.93 0.48 Control Room 1.5 i

1 4

l Table 6 Whole Body Doses from Postulated Accidents (Rem)

Accident EAR L2Z 1.

Large Break LOCA HSIV and Containment Leakage 3

2.9 ECCS Leakage 0.02 0.02 Total 3.02 2.92 Control Room l

HSIV & Containment 4.8 ECCS leakage 0.03 l

Total 4.83 2.

Main Steam f.ina Break 0.084 0.01 9 Equilibrium IS Value 0.005 0.0006 0 10 times Equilibrium TS Value 0.05 0.006 Control Room 0 Equilibrium TS Value 0.2 9 10 times Equilibrium 2

TS Value 3.

Control Rod Drop 0.37 0.046 l

Control Roon 1.9 l_

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Fuel ihn<1 ling 0.21 0.11 l

Control Room 0.34 l

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