ML20153C982

Safety Evaluation of TR NEDC-32523P, Generic Evaluations of GE BWR Extended Power Uprate, Including Suppl 1,Vols 1 & I. Rept Acceptable

ML20153C982
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Issue date:	09/14/1998
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NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20066-4001

. . . . . ,o SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION RELATED TO GENERAL ELECTRIC LICENSING TOPICAL REPORT NEDC-32523P GENERIC EVALUATIONS OF GENERAL ELECTRIC BOILING WATER REACTOR EXTENDED POWER UPRATE

1.0 INTRODUCTION

General Electric Company (GE) submitted licensing topical report NEDC-32523P, " Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate,"(Ref.1)in March 1996. This licensing topical report, known as ELTR2, contains generic bounding analyses and equipment evaluations in support of the proposed extended power uprate program for boiling water reactors (BWRs). GE submitted a nonproprietary version of ELTR2 in April 1996, at the staffs request. GE later submitted Supplement 1 to ELTR2 in June 1996 (Ref. 2). ELTR2, Supplement 1, contains an evaluation of the NRC rules and regulations, NRC-issued generic correspondences, and the industry-issued generic correspondences for applicability to the BWR extended power uprate program. In response to the staffs request on March 17,1997, GE provided additional information in a letter dated July 2,1997 (Ref. 3).

The BWR extended power uprate program was initially introduced in GE licensing topical report NEDC-32424P, " Generic Guidelines for General Electric Boiling Water Reactor Power Uprate,"

(Ref. 4)in February 1995. This licensing topical report is known as ELTR1. The NRC staff has previously reviewed and approved ELTR1 in a staff position paper dated February 8,1996 (Ref. 5). References 4 and 5 provide guidance to licensees on the scope and content of information to be submitted as part of a plant-specific power uprate submittal. ELTR2, with the staffs endorsement, is intended for use in conjunction with ELTR1 and requisite plant-specific information in the assessment of a licensee's request for an extended power uprate.

2.0 EVALUATION This staff position documents the staffs review of the generic analyses and evaluations provided in ELTR2. In many cases, the staff has determined that additionalinformation will be required from plant-spocific submittals in order to complete the staffs evaluation of the subjects discussed in ELTR2. The additional requirements for these subject areas are delineated in the applicable sections of this staff position. A detailed evaluation of each subject area discussed in ELTR2 is provided in the following sections.

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2 2.1 Review of Generic Communications GE has evaluated the generic communications that were issued prior to December 1995 for their potential applicability to extended power uprate. GE's review scope included such generic communications as NRC-issued generic letters, bulletins, information notices, circulars, the industry-issued Institute for Nuclear Power Operations'significant operating experience reports, and GE service information letters and rapid information communication service information letters. The staff has previously reviewed and approved GE's evaluation of generic communications for power uprate in the staff safety evaluation dated July 31,1992 (Ref 6).

The staff noted that GE has applied the same generic communication review process that was described in Reference 6 for extended power uprate. Therefore, the staff concludes that licensees applying for an extended power uprate should review and address the generic communications listed in Table 2-1 of the ELTR2 as " Plant Dependent," as well as generic l communications that were published since December 1995. '

2.2 GE Set Point Methodoloav GE stated that the set point methodology discussed in topical report, NEDC-31336P, " General Electric Instrumentation Setpoint Methodology," dated October 1986 (Ref. 7), will be used to determine instrument set points for the extended power uprate. The methodology in NEDC-31336P has been previously approved by the staff and is documented in a safety evaluation dated February 9,1993 (Ref. 8). Since the methodology described in NEDC-31336P is not power level dependent, the staff finds that use of this methodology for the BWR extended power uprate program as described in ELTR2 is acceptable. However, licensees electing to use a set point methodology that is different from the methodology in NEDC-31336P should provide justifications for changes to the instrument set points on a plant-specific basis. l l

The staff also notes that ELTR2 states that in some cases, extended power uprates may result in an increase in instrument span and, as a consequence, recalibration of these instruments will be required because of higher pressure or temperature conditions associated with the instrument functions. Since this condition is plant dependent, each applicant for extended power uprate should identify all such instruments (for which an increase in instrument span has resulted in the need for recalibration) in its plant-specific application for extended power uprate.

2.3 Emeraencv Ooeratina Procedures (EOPs) -

As a result of power uprate, a number of variables and limits utilized in plant-specific EOPs may be affected. In particular, the increase in rated reactor power will directly or indirectly affect many of the variables and limit curves contained in the existing plant EOPs. Although the conditions that require operator actions in the EOPs may change, the operator actions described in the EOPs will remain the same after power uprate.

The ELTR2 includes a table that lists certain sets of variables and limits in EOPs that will require recalculations as a result of power uprate. The table is based upon Revision 4 of the Emergency Procedure Guidelines (EPGs) and draft EPGs/ Severe Accident Guidelines that have been developed by the Boiling Water Reactor Owners' Group (BWROG). Each applicant

3 for extended power uprate will be required to recalculate its plant-specific EOP variables and limit curves using the guidance provided in ELTR2. Each applicant for extended power uprate will also be required to ensure that the EOPs are modified to address changes to other process variables that are not addressed in ELTR2.

3.0 ANALYTICAL EVALUATIONS GE has provided generic bounding analyses and equipment evaluations for selected nuclear steam supply system equipment. Staff review of GE's bounding analyses and equipment evaluation is presented below.

3.1 Loss of Feedwater Flow Transient 4

GE provided the results of loss of feedwater (LOFW) flow transient analyses for all classes of operating reactors (BWR/3 through BWR/6) at power uprate conditions. The purpose of the analyses was to demonstrate that the original design bases of the reactor core isolation cooling (RCIC) or isolation condenser (IC) systems are preserved during an LOFW flow transient. Both RCIC and IC systems were designed to maintain the reactor vessel water level above the top of the active fuel (TAF) during an LOFW flow transient when the other higher capacity high pressure water supply system was assumed to have failed. The bounding analyses for plants in BWR/3, BWR/4, BWR/5, and BWR/6 product lines were presented by each product line and uprated power level. Two criteria were applied to the LOFW event: (1) the limiting criterion requires that the RCIC or IC will maintain the water level inside the core shroud such that the TAF remains covered throughout the event, and (2) the operational criterica requires that the RCIC will maintain the downcomer wide-range water level high enough such that the low-level trip set point (Level 1) for BWR/4s, BWR/5s, and BWR/6s is not activated, and that the RCIC or IC will maintain the fuel-zone water level above the level at which the EPGs direct the operator to initiate the automatic depressurization system for BWR/2s and BWR/3s. The minimum water levels for tnis transient with uprated conditions are presented for different power levels and RCIC and IC capacities. These results may be used as an input in choosing a target maximum power level that will ensure maintaining an adequate water level for LOFW transients. The plant-specific submittal should address the impact of the reduced water level on operator action times for the LOFW transient with additional failures. This may be treated in the plant-specific probabilistic risk assessment or in the individual plant examination.

3.2 Stability 3.2.1 Stability Interim Corrective Actions The BWROG and the staff have addressed interim corrective actions to minimize the occurrence and potential effects of core power oscillations. Core power oscillations have been observed at a few BWRs during certain operating conditions, particularly during plant operations with low core flow on a high power rod line on a power / flow map. Until this issue is resolved, individual plant-specific submittals must adopt the operational constraints described in the ELTR2 that are consistent with the guidance provided in NRC Bulletin 88-07 (and Supplement 1) (Ref. 9). Specifically, these operational constraints will continue to restrict plant operation in the high power rod line and the low core flow region of the power / flow map. The

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. 4 required operator actions upon entry into the regions of the power / flow map identified as having ,

a low stability margin shall remain as specified in Supplement 1 to Bulletin 88-07. Licensees l 4

should continue to follow the guidance provided in the BWROG Guidelines for Stability Interim

Corrective Action that was submitted to the NRC by a letter dated June 6,1994. The existing i stability corrective actions are applicable or adaptable to extended power uprate operation. l 3 Stability boundaries are kept the same in terms of absolute power and core flow; and power I levels, reported as a percentage of rated power, are adjusted to the new uprated power.

3.2.2 Lono-Term Solution

- Licensees may choose any of the options for the long-term stability solution that are described in ELTR2. These long-term options are capable of supporting extended power uprate conditions. GE states that the effectiveness of the described long-term solution options is not

impacted by the extended uprate operation. The prevention and detection / suppression features of the long-term stability solutions are either demonstrated to be unaffected by power uprate or are modified and validated in accordance with the solution methodology. Each applicant for

l extended power uprate should address the effect of its proposed power uprate on its long-term

stability solution.

I 3.3 Core Sorav Distribution i

l GE addressed the continued applicability of the core spray distribution assumptions utilized in the GE LOCA/ECCS [ loss-of-coolant accident / emergency core cooling system) models for

" extended power uprate. The core spray distribution assumptions, modeling, and application ,

methodology in SAFER /GESTR-LOCA analyses are found to be adequate by GE for BWR/2, j BWR/3, BWR/4, BWR/5, and BWR/6 plants at power uprate conditions.  :

3.3.1 Short-Term Imoact i

i in the short term following a postulated LOCA, no credit is given for core spray flow to high

power fuel bundles until the upper plenum region forms a pool of water covering the upper tie

, plate of all fuel bundles. The drainage flow rate to the high and average power fuel bundles is l

determined by counter current flow limiting (CCFL) characteristics, further reducing the credit l given for core spray. The model allows for CCFL breakdown in the peripheral region of the 4 core after the upper plenum water level rises above the core spray sparger. When the CCFL breakdown occurs in the peripheral bundles, a rapid drainage of water occurs from the upper I plenum to the lower plenum through the peripheral bundlea, supporting a reflooding of the core.

] This evaluation methodology minimizes the credit for spray cooling during the short-term

response to a postulated LOCA. GE concluded, by a bounding analysis on the peripheral bundles in an uprated core, that the effect on peak cladding temperature (PCT) was less than

. 2 'F in the short term. Since the increase in core power for uprate is accomplished by flattening I the radial bundle power profile, assurance must be provided that the assumption of CCFL 2 breakdown in peripheral bundles will remain valid. Each applicant for extended power uprate should adhere to existing radial power shape limitations when designing core reloads for

uprated conditions. Provided that the radial power distribution remains within the bounds of the l LOCA/ECCS assumptions, the effect of power uprate on the short-term response to a j postulated LOCA should be minimal.

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5 3.3.2 Lona-Term Coolina Imoact in the longer term, credit for spray cooling is given while water level in the core shroud remains below TAF. For these conditions, at least one core spray loop is assumed to be operating.

Test data for the verification of core spray distribution is based on the short-term portion of the l

accident analysis when power levels and steam generation from the core (or from reactor depressurization) are much higher than in the long-term portion of the analysis. Therefore, steam generation during the long term portion of the accident will be much less severe than during the short-term portion, and the effect of core spray distribution on the long-term response to a postulated LOCA is bounded by the effect on the short-term response. Any steam generated in the long term is quenched and the fuel is adequately cooled by the core ,

spray flow. l 3.3.3 Conclusion The effects of extended power uprate on short-term response are addressed in the GE's LOCA/ECCS models and the plant-specific submittal will utilize these models to show compliance with the enteria specified in 10 CFR 50.46. Since the increase in core power for uprate is accomplished by flattening the radial bundle power profile, assurance must be provided that the assumption of CCFL breakdown in peripheral bundles will remain valid. Each applicant for extended power uprate should adhere to existing radial power shape limitations when designing core reloads for uprated conditions. Provided that the radial power distribution remains within the bounds of the LOCA/ECCS assumptions, the effect of power uprate on the short-term response to a postulated LOCA should be minimal. The impact of power uprate on the long-term response to a LOCA will continue to be bounded by the short-term response.

3.4 Safetv Limit Minimum Critical Power Ratio (SLMCPR)

A plant-specific power uprate and the reload submittal should contain analyses to confirm that the SLMCPR is appropriate for the average bundle power at the uprated conditions. This should be done by comparing bundle power to the applicable SLMCPR basis in NEDE-24011-P-A-10-US, " General Electric Standard Application for Reactor Fuel (GESTAR),"

Supplement, March 1991 (Ref.10). If a new plant-specific SLMCPR is needed because the core average bundle power exceeds the documented licensing basis, then the SLMCPR should be established using the same NRC- approved procedures, and it should be included in the plant-specific submittal. Where extended power uprate results in a greater number of bundles operating near the limit, the SLMCPR may be increased to provide the same statistical confidence level that fuel rods will avoid boiling transition.

These procedures ensure that 99.9 percent of the fuel rods will avoid boiling transition during any abnormal operational occurrence. The same safety margin between the onset of transition

! boiling and the point at which fuel damage would occur is maintained, and the same transition l-boiling avoidance probability is maintained. Transient events will continue to be analyzed against this SLMCPR, using NRC-approved procedures, when establishing the operating limit MCPR. This operating limit MCPR will be documented in each plant-specific power uprate submittal and confirmed for each cycle of operation in the cycle-specific reload analysis.

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6 3.5 Containment Atmosobere Combustibility The Commission's regulation as specified in 10 CFR 50.44 requires licensees to install means to control hydrogen gas that may be generated following a postulated LOCA. Because extended power uprate will not require a new fuel design, the ELTR2 assumes that fuel assemblies utilized for extended power uprate would have no significant difference in the amount of cladding material, and hence in the total amount of hydrogen gas generation during a postulated LOCA. Based on this assumption, the ELTR2 concNdes that extended power uprate would not affect the design-basis metal water hydrogen generation source-term in the design of installed combustible gas control equipment. Should a new fuel design be introduced for any core, independent of power uprate, it should be evaluated for potential change in the amount of cladding material.

The post-LOCA containment atmosphere combustibility is also affected by the concentrations of hydrogen and oxygen produced by radiolytic decomposition of water in the reactor. As a result of power uprate, the post-LOCA production of hydrogen and oxygen from radiolysis will increase proportionally with reactor power level. Mark Ill type containments implement both igniters and recombiners to control containment atmosphere combustibility. Mark I and ll containments generally use either containment atmosphere dilution systems, which inject nitrogen to dilute the combustible gas concentration, or hydrogen recombiners, which thermally or catalytically recombine hydrogen and oxygen to maintain combustible gas concentrations below established limits. The ELTR2 coacludes that this smallincrease in hydrogen and oxygen concentration is within the capacity of currently installed combustible gas control systems. Plant specific submittals should confirm the capability of the combustible gas control system to comply with Regulatory Guide 1.7, " Control of Combustible Gas Concentrations in Containment Following a Loss-of-Coolant Accident," and should address any procedural or equipment set point changes that may be required to assure adequate containment atmosphere combustible gas control.

3.6 Material and Coolant Chemicals GE concluded that extended power uprate will result in a slight increase in a plant's susceptibility to erosion / corrosion of components including piping, but there should be no adverse impact of extended power uprate on erosion / corrosion susceptibility orovided that licensees reexamine their existing erosion / corrosion inspection programs. The staff concurs with GE's conclusion. Special attention should be paid to the plant-specific basis for selecting the locations for erosion /corresion inspections to determine whether additional locations need to be included in the erosion / corrosion inspection program.

3.7 Anticioated Transient Without Scram (ATWS) Evaluations GE has evaluated the acceptability of the consequences of ATWS for the BWR/3 product lines for extended power uprate conditions. The transients that were analyzed include (1) main steam isolation valve closure (MSIVC), (2) pressure regulator failure open (PRFO), (3) loss of offsite power (LOOP), and (4) inadvertent opening of a relief valve (IORV). The initial conditions of 120 percent of the currently rated power level and a steam flow of 124 percent of the currently rated value were assumed for these evaluations. The initial conditions used in the

7 l analysis are presented in Table 3-3 of ELTR2. The ATWS mitigation features required by 10 l CFR 50.62, such as the recirculation pump trip, alternate rod insertion (ARI), and the standby liquid control system (SLCS) flow rate of 86 gpm are also taken into account in the analysis.

l The ARI is assumed to fail in the analysis of ATWS events. Operator actions needed to control ATWS mitigating systems, in addition to the water supply systems of feedwater, high-pressure l coolant injection (HPCI), and RCIC, are assumed in accordance with the BWROG EPGs. The sequence of events for the MSIVC, PRFO, LOOP, and ICAV are presented in Tables 3-4,3-5, 3-6, and 3-7 of ELTR2. The analysis results of the ATWS events and the acceptance criteria are presented in Tables 3-8 and 3-9 of ELTR2.

The potential effect of power uprate on shutdown capability and injection capability of the SLCS should be evaluated on a plant-specific basis. Power uprate may require a slight increase in -

L shutdown concentration or boron enrichment requirements, due to higher fuel batch fractions or enrichments. Plant-specific evaluations should also be performed to assure that the SLCS has high enough operating pressure to inject adequate boron solution at higher reactor pressure.

l The differential pressure margin between the SLCS pump discharge relief valve setting and the maximum discharge pressure must be sufficient to reliably meet its injection flow rate requirements.

l The ODYN computer code has been reviewed by the staff for application to BWR/3 ATWS analyses. The staff considered as part of its review the benchmarking and model description provided by GE in NEDC-24154P (Ref.11). The staff also performed a series of audit calculations-to confirm the GE boron reactivity model and the adequacy of ODYN predictions.

Based upon this review, the staff concluded that ODYN is acceptable for application to BWR/3 ATWS calculations provided that the code is used within the restrictions discussed in Section 5.1.3 of N,EDC-24154P.

3.8 Overoressure Protection 3.8.1 Introduction The reactor vessel overpressure analysis is performed to show compliance with the American i Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code Section ill overpressure criteria. There are two main overpressure transients: all MSIVC, and l turbine / generator trip coincident with failed bypass valves (T/GTWOBP). These represent the most severe abnormal operational transients resulting in a nuclear system pressure rise. The l evaluation of these transients assumed failure of the direct primary scram and that only the l indirect backup scram would be available. The MSIVC event is more severe than the T/GTWOBP at extended uprate power levels when credit is taken for the first backup scram.

Therefore, it is used as the ASME overpressure protection basis event. A pressure relief system is provided to protect the nuclear system from overpressure. The system consists of safety / relief valves (SRVs) located on the main steam lines between the reactor vessel and the first isolation valve within the drywell.

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8 3.8.2 Puroose In the extended power uprate overpressure studies it was confirmed that the MSlVC with backup high neutron flux scram is the most limiting event. The study was used to determine the effects of different valve types and configurations on the overpressure analyses by evaluating overpressure over a sampling of representative plants. Analyses were done for plants of various size, power density, SRV capady, and valve type.

3.8.3 Initial Conditions and Analvsis Assumotions ASME requirements are found in the ASME Boiler and Pressure Vessel Code Section 111 and are summarized in ELTR2. The Code requirements for overpressure protection include the i following: (1) the maximum vessel pressure must not exceed 110 percent of vessel design pressure, typically 1375 psig for GE BWRs; (2) the analysis shall be performed at least at 102 percent of the maximum licensed power; (3) the initial pressure shall be at least equal to the technical specification (TS) limiting condition for operation (LCO); and (4) the conservative scram, void, and Doppler reactivity for power increase events must be used. The Code allows a credit for the dual purpose SRVs in the ASME-qualified mode of safety operation. The Code also allows credit for the first backup scram, the high neutron flux scram, and the MSIV fast closure within the minimum specified time, which is normally 3 seconds.

3.8.4 Analvsis Sample plant configurations by plant model, SRV manufacturer and type, vessel size, and 120 percent of original licensed power is provided in Table 3-11 of ELTR2. An example of the SRV set points used for each sample configuration is provided in Table 3-12 of ELTR2. The ODYN computer code was used to calculate the vessel pressure response to the overpressure event.

The initial power is assumed to be 1.02 times the uprated power level and the initial pressure is the peak reactor operating pressure allowed by the TS LCO for reactor operating pressure.

The SRV opening set points plus a 3-percent upper limit on the actual set point is also assumed in the analysis.

3.8.5 Results Figure 3-16 in ELTR2 shows the effect of increased initial dome pressure on overpressure study results, with sensitivity to the number of valves considered operable, and the relieving capacity of each SRV as a percent of initial steam flow. The results show that the TS limit on the number of operable valves may need to be assessed in order to provide acceptable overpressure protection. The results also show that the peak vessel pressure is a strong function of initial operating dome pressure, and that the reactor has the potential to increase power output by maintaining an initial dome pressure of less than 1100 psia. This may require a steam turbine modification. The effect of a 20-percent uprate increase in initial steam flow causes about a 20-percent increase in peak pressure.

Table 3-13 in ELTR2 provides the results of a sensitivity study between BWR types, initial power level, initial dome pressure, number of SRVs out of service, lowest SRV set point, simmer margin, and the calculated maximum pressure. The results from Table 3-13 show that

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!* 9 l an increased dome pressure of 50 psi will still maintain margin to the ASME overpressure limit for power level increases up to 120 percent of original rated power. However, the number of I allowed out-of-service SRVs may need to be reduced in some cases.

3.8.6 Conclusions Overpressurization is a strong function of initial operating dome pressure rather than power.

The results indicate that, for the plants considered, a margin to the overpressure limit of 1375 i psig is available at the 120 percent power uprate conditions with an initial dome pressure less i than 1100 psia. This implies that from an overpressure perspective, it is feasible for power i uprate to be as high as 120 percent power with an increase in operating dome pressurs by 4

about 75 psig.

j in the plant-specific power uprate, the SRV set points must be reevaluated to ensure that the ASME mechanicallimits and simmer margin are satisfied. The sensitivity of any unpiped (discharge directly to containment) valves on the overpressure results must be considered.

Plant-specific overpressure analyses must be submitted with each power uprate request, as well as with each core reload design submittal.

4.0 HARDWARE CAPABILITY EVALUATIONS GE has evaluated the effect on hardware capability due to power uprate assuming the following primary operating condition changes: (1) increased poser level of 20 percent (i.e., heat flux, stored heat, fission products, neutron fluence), (2) increased reactor pressure (to 1095 psia),

(3) increased reactor temperature (to 556 *F), and (4) increased steam and feedwater flow by approximately 24 percent. The other operating conditions assumed in the hardware evaluations are given in Table 5-1 of ELTR1.

4.1 Low Pressure Emeraenev Core Coolina Systems GE has performed an assessment of the impact of power uprate of up to 120 percent on the performance and design requirements of the low pressure systems that provide emergency core cooling following a DBA (design-basis accident]-LOCA. Typically, the performance requirements and TS requirements (e.g., flow rates and start times) for these systems will not be changed due to power uprate, and plant-specific evaluations are performed to assure that the existing acceptance criteria can be met at the higher power.

4.1.1 Low Pressure Core Sorav (LPCS)

The purpose of the LPCS, in conjunction with other ECCS, is to provide core cooling during LOCA conditions by spraying water into the reactor upper plenum. The LPCS will be evaluated to demonstrate that the system, in conjunction with other ECCS, is capable of providing sufficient core cooling to comply with the 10 CFR 50.46 criteria as part of the plant-specific analyses as specified in Section 5.3.1 and Appendix D of ELTR1. Section 3.3 of ELTR2 discusses the core spray distribution analysis assumptions at power uprate conditions.

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4.1.2 Low Pressure Coolant Iniection (LPCI)  !

l The purpose tne LPCI mode of the residual heat removal (RHR) system, in conjunction with j other ECCS, is to provide core cooling during LOCA conditions by injecting water into the I recirculation pump discharge piping or inside the reactor core shroud. The LPCI will be evaluated to demonstrate that the system, in conjunction with other ECCS, is capable of providing sufficient core cooling to comply with the 10 CFR 50.46 criteria as part of the

,,lant-specific analyses as specified in Section 5.3.1 and Appendix D of ELTR1.

4.1.3 Reactor Shutdown Coolina (SDC)

The !nitial cooldown of the reactor is via discharge of steam to the main condenser until the reactor pressure is reduced to the SDC interlock pressure. The SDC was originally designed to cool the reactor from the saturation temperature associated with the SDC interlock pressure to ,

125 *F within 20 hours2.314815e-4 days <br />0.00556 hours <br />3.306878e-5 weeks <br />7.61e-6 months <br /> following "all control rods in." A plant-specific analysis to determine the i time to achieve the 125 *F will be made as part of the analyses described in Section 5.4.6 of ELTR1. The SDC usually establishes the heat exchanger heat removal capacity requirements for BWR/3, BWR/4, and BWR/S reactors.

For BWR/6 plants, the RHR heat removal capacity is established by the suppression pool cooling requirements. This heat removal capability exceeds the 20-hour cooldown requirement. l Therefore, the SDC requirement is expected to be met at the 120-percent power uprate l condition. A plant-specific confirmatory analysis will be made to verify the 20-hour cooldown requirement as described in Section 5.6.4 of ELTR1.

4.1.4 Reactor Steam Condensino (SCl The RHR SC mode of operation is designed to allow the reactor to be maintained in the {

hot-standby condition. Some plant designs also take credit for the SC mode following a reactor I isolation to terminate SRV cycling and thereby reduce the number of SRV cycles considered in  !

the containment design.

The uprated power condition will result in increased steam generation and, therefore, an ,

increase in the SC-mode required heat removal capacity. A plant-specific analysis must be i performed to establish that the design requirements of the SC mode are met. It is also possible that the RHR SC mode may not be needed and can be eliminated.

4.1.5 Low Pressure ECCS Eauioment Qualification Desian Considerations GE indicated that some of the equipment and associated equipment qualification may be l impacted by the increased temperatures and pressures resulting from the power uprate. The j following parameters were evaluated by GE to ensure that the impact of power uprate is i acceptable: the design temperature and pressure of the reactor coolant pressure boundary (RCPB), the design temperature of the components that are exposed to suppression pool water  ;

and are not a part of the RCPB. the ambient temperature of the surrounding equipment, the l pump mechanical seals, and the pump net positive suction head (NPSH). These parameters  !

and their effect on the system components are discussed in the following paragraphs. f 1

1 11 The power uprate is expected to result in an increase in the peak pressure and temperature conditions in the RCPB during transients. The components that are part of the RCPB will be evaluated in the plant-specific submittal to establish that these higher pressures and temperatures are acceptable.

The power uprate will result in higher suppression pool temperatures following a LOCA, during the alternate SDC mode, and during certain plant transients. Components exposed to these higher suppression pool temperatures will be evaluated to demonstrate that the resulting l increase in thermal stress in the components is within acceptable limits and that the impact on l the component design temperature is acceptable. These evaluations are part of the l plant-specific submittal as specified in Appendix J of ELTR1, The low pressure portions of the '

RHR and LPCS systems are not subjected to higher pressures from power uprate because the low pressure injection permissive set points will not be changed for power uprate.

The ambient temperature of the surrounding system equipment will be increased by the higher suppression pool temperatures due to power uprate. The RHR and LPCS systems equipment will have increased temperatures due to convective heating in the equipment rooms.

GE indicated that the pump mechanical s'eals are generally acceptable for suppression pool temperatures up to 212 *F. If the suppression pool temperature exceeds the seal design limits, the acceptability of the seal to operate at the higher temperatures will be evaluated. It is possible to replace the seals with seals designed to operate at higher temperatures, if necessary. A plant-specific evaluation should be performed to demonstrate that the pump seals are acceptable at the power uprate conditions.

The effect of higher suppression pool temperature on pump NPSH will also be evaluated in the plant-specific submittal as discussed in Appendix G in ELTR1. This evaluation will consider the current plant configuration as well as including any effects of changes to the suction strainers.

The impact en the available RHR and LPCS pump NPSH will be evaluated at the higher suppression pool temperatures associated with power uprate.

4.2 Hiah Pressure Coolant Iniection (HPCI) and Reactor Core Isolation Coolina (RCIC)

Systems The design bases for the HPCI and RCIC systems are to provide reactor vessel inventory control during (1) small and intermediate break size LOCAs (HPCI with other ECCS as backup),

and (2) transients involving loss-of-feedwater flow (RCIC with HPCI as backup). These systems are designed to provide their rated flows over a reactor vessel pressure range from 150 psig to a maximum pressure based on the lowest opening set point for the SRVs. The opening set points for the SRVs will be increased to maintain an adequate simmer margin above the increased reactor operating pressure associated with power uprate. Increasing the GRV pressure set points will have a potential impact on the maximum operating pressure (75 psi for the 20-percent power uprate) for the HPCI and RCIC systems for reactor isolation events. The assumed increase in reactor operating pressure of 75 psi results in an upper analytical set point increase for the lowest group of valves of about 1240 psig. Increasing the SRV set point pressure increases the required operating pressure of the HPCI and RCIC

12 systems. The plant-specific submittal will determine the maximum operating pressure for the HPCI and RCIC systems.

The required HPCI and RCIC water flow rates will remain unchanged after uprate. However, the pump and turbine operational requirements are increased due to the increased SRV pressure set points (as described above). The required increase in HPCI and RCIC discharge pressures will require the turbine speeds to be increased slightly (about 3 percent at the new operating condition). This change in turbine speed was evaluated by the vendor against the manufacturer performance curves and found not to significantly affect the operation of the HPCI and RCIC systems. Both the HPCI and RCIC systems should be capable of meeting the design flow requirements at the increased pressures associated with uprate.

The ELTR2 assessed the impact of increased reactor pressure on the potential for turbine overspeea during startup of the HPCI and RCIC systems. The increased ree': tor operating pressure associated with uprated conditions has the potential to result in increased turbine overspeeding during system startup, increasing the probability of the system to trip. The ELTR2 stated that modifications to HPCI systems that use Terry Corporation turbine assemblies will be made as described in GE Services information Letter (SIL) No. 480.

Likewise, modifications to the RCIC system will be made as described in GE Sll No. 377. In order to reduce the possibility of turbine overspeed trips, plant-specific submittals must address the modifications described in GE SIL No. 480 and GE SIL No. 377 (or equivalent modifications).

l Plant-specific submittals must provide assurance that the HPCI and RCIC systems will be capable of injecting their design flow rates at the higher reactor operating pressures associated with power uprate. The higher pump discharge peak pressure must also be evaluated for compliance with the design pressure for the as-built system piping. Additionally, each licensee must also provide assurance that the reliability of these systems will not be decreased by the higher loads placed on the systems or because of any modifications made to these systems to compensate for these increased loads.

4.3 Hiah Pressure Core Sorav (HPCS)

The HPCS system on BWR/5 and BWR/6 plants consists of a single, motor-driven centrifugal pump located outside of the primary containment, a peripheral ring spray sparger located in the reactor vessel above the core region, and associated piping, valves, controls, and l instrumentation. The system is designed to operate from normal offsite auxiliary power or from l an emergency diesel generator if offsite power is not available. The primary purpose of the i HPCS system is to maintain reactor vessel water level inventory during a postulated

! small-break LOCA that does not immediately depressurize the vessel. The HPCS also serves

! as a backup to the RCIC system for the loss of feedwater transient for BWR/5 and BWR/6 product line plants. The HPCS system was designed to provide makeup water over the entire range of reactor operating pressures, and the physical equipment design of the HPCS system is compatible with a reactor maximum design pressure of 1250 psig, which bounds the potential range of HPCS system operating pressures.

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13 l Increased reactor vessel operating pressures have little impact on HPCS system effectiveness i

during a postulated large-break LOCA since the primary contribution of HPCS to core cooling would occur during and following reactor depressurization. For postulated small- and intermediate-break LOCAs, peak fuel cladding temperatures w H increase slightly due to lower HPCS flow rates caused by increased reactor vessel pressures. The effect of uprate on HPCS flow and resultant PCTs for small- and intermediate-break LOCAs will be verified on a plant-and fuel-bundle-specific basis and will be documented in the plant-specific power uprate submittal.

The HPCS flow rate at a vessel pressure of 1000 psig is approximately 3 times larger than the RCIC flow rate for all BWR/5 and BWR/6 plants. For events in which MSIV closure is coupled with the LOFW flow transient, the reactor pressure will be maintained by the SRVs, and the HPCS flow rate, when the reactor pressure is above its design operating pressure, will be reduced. A plant-specific LOFW analysis specified in ELTR1 will consider the flow decrease '

with increased SRV set point in determining whether RCIC or HPCS failure is the worst case for the LOFW transient. Therefore, its impact will be evaluated on a plant-specific basis to assure that adequate system flow is available to meet the design basis of the plant. The HPCS flow capability will be evaluated at the maximum operating pressure.

In summary, the HPCS design flow rate versus pressure requirements will be evaluated to confirm that they are adequate to meet the original design basis for the LOFW flow transient for power uprate conditions. The PCT will be confirmed for the LOCA cases on a fuel-design-specific basis for each plant specific power uprate submittal. The flow capability at the power uprate pressures will be evaluated to confirm that they are adequate for all design-basis LOCA and transient conditions. The HPCS system is designed for pressure and temperature conditions compatible with those for the reactor vessel that bounds the power uprate conditions and should be acceptable for power uprate conditions.

4.4 Control Rod Drives (CRDs) and Scram Performance The CRD system controls changes in core reactivity by positioning neutron absorbing control rods within the reactor it is also required to scram the reactor in emergency conditions by rapidly inserting withdrawn control rods into the core. The CRD system was evaluated at the uprated steam flow and dome pressure conditions. As discussed below, the current CRD system will continue to carry out all its functions within current performance requirements at extended power uprate.

4.4.1 Control Rod Scram The increased reactor vessel dome pressure associated with power uprate produces a corresponding increase in the bottom head pressure. This increase in reactor pressure has an effect on the insertion times for control rods during a scram. For pre-BWR/6 product line l reactors, initial control rod scram insertion speeds are slowed due to the increased reactor l pressure. However, near the end of the control rod stroke, the increased reactor pressure will

! speed up the control rod insertion rate, resulting in a slightly shorter overall scram time for the l control rods. Pre-BWR/S plants will be required to provide assurance that the scram time

performance indicated in the current plant TSs will be maintained. If needed, the minimum

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14 pressure in the hydraulic control unit (HCU) accumulator could be increased. The CRD system pumps can pressurize the HCU accumulators to a higher value, in this case, an evaluation will be performed to assess the adequacy of the system design to operate at the higher pressure.

Due to a difference in CRO system design, BWR/6 plants may experience slightly longer control l rod scram stroke times after uprate. The scram performance of the normal control rod /CRD j should have sufficient margin to meet scram performance requirements. This anticipated change in scram performance may require changes to the plant TS requirements for scram stroke times. The plant specific submittal for BWR/6 plants must provide assurance that the l scram insertion speeds used in the transient analyses are slower than the requirements I contained in the plant TSs.

The generic scram times for ASME overpressure protection and critical power ratio pressurization transient analyses are based on generic pressure versus time envelopes. Each plant-specific power uprate will confirm that the transient pressure remains within the generic envelope, and if not, the scram time will be recalculated.

l 4.4.2 Control Rod Positionina and Coolina Normal control rod insertion and withdrawal functions will not be significantly affected by the increase in reactor bottom head pressure. Normal CRD header pressure is maintained approximately 250 psig above the lower head pressure. The flows required for CRD cooling and driving are assured by automatic opening of the CRD system flow control valve (FCV), thus compensating for an increase in reactor pressure. li the FCV is near the full open position the i ability to insert or withdraw CRDs with excess seal leakage may be reduced for normal operation. The plant-specific submittal will address the ability of the CRD system to maintain an adequate pressure differential for control rod operation.

4.4.3 Control Rod Drive Mechanism (CRDML For the power uprate operation at 120 percent of the original licensed power level, the reactor dome pressure will be 1095 psia which produces an operating pressure of 1135 psia (or 1120 psig) at the reactor bottom head. The reactor bottom head temperature will be 540 'F.

The components of the CRDM that form part of the primary pressure boundary have been designed for a dorne pressure of 1250 psig and a temperature of 575 *F, which bounds the uprated power condition.

The limiting component of the CRDM is the indicator tube which has a calculated primary membrane plus bending stress of 20,790 psi and 23,830 psi for the BWR/2-5 and BWR/6 type CRD, respectively. The allowable stress is conservatively specified as 1.5 Sm or 25,860 psi.

The maximum fatigue usage factor was calculated to be 0.15 for the limiting component of the CRD main flange.

Therefore, the staff concludes that the CRDM will continue to maintain its structural integrity at the extended uprated power condition in accordance with its design basis.

15' 4.5 Recirculation System I

The recirculation system consists of the recirculation pumps, the suction and discharge valves, the recirculation flow control valves or the variable frequency motor-generator sets and adjustable speed drives, the jet pumps, the piping,'and instrumentation and controls. The recirculation system is considered to be nonsafety-related because its function is to vary the core flow and the power level during normal operation.

Plant-specific data will be reviewed to confirm that the existing recirculation system will accommodate the increase in resistance due to an increase in core average void fract:on at the uprated condition when operating at maximum core flow. The results of this review will be documented in the plant specific submittal. As specified in Section 5.6.2 of ELTR1 the evaluation of the recirculation system vibration will also ba documented in the plant-specific submittal.

A plant specific analysis of the loss of one feedwater pump event will be submitted per Appendix E of ELTR1 to assess the effect of a higher flow control line on scram avoidance.

4.5.1 Recirculation Pumos GE indicated that the recirculation pumps were originally designed for pressures of 1500 to 1650 psig. Based on the review of overpressure transients, GE determined that the recirculation pump design pressure and temperature have adequate margin for uprated conditions. The slightly increased drive flow associated with the power uprate will be achieve J by an increased pump speed for plants with variable speed pumps, or by an increased flow control valve opening for plants with constant speed pumps. A plant-specific evaluation should be performed to ensure continued compliance with the design basis and also to ensure that there are no detrimental effects on vessel internals caused by vibration at the higher vane I passing frequencies.

GE indicated that the mechanical seals should be capable of withstanding the higher pressure, since each of the two seal stages has the capability to withotand reactor pressure. The l Increased pressure should have no detrimental effect on the motor bearings, heat exchangers,

! or running clearances. The increased recirculation flow may have a more severe service duty l

on the pump shafts on plants that do not have upgraded shaft designs. Pump vibration monitoring and inservice inspections should be adequate to detect and limit the extent of pump shaft cracking. An increase of approximately 20 percent in power will be accompanied by an increase of 75 psig and 9 'F. These increases are small when compared to operating conditions of approximately 1000 psig and 540 *F.

Each applicant for extended power uprate will be expected to review plant-specific operating data to ensure that the recirculation system, including the recirculation pumps and its associated components, will accommodate the increase in system pressure as well as the increase in flow resistance that is expected due to the increase in core average void fraction due to uprate. The results of this review will be documented in the plant-specific uprate i submittal. An evaluation of recirculation system vibration will also be included in the plant-i f

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specific submittal. The licensee must ensure that the recirculation system, as well as other i

pressure boundary components or systems, continue to meet ASME Code requirements. )

4.6 Safetv/ Relief Valves (SRVs) l 1

The performance of SRVs was evaluated under uprated power conditions for the impact of higher steam flow (124 percent), a temperature increase of approximately 10 *F, and an 1 operating pressure increase of 75 psi. The plant-specific submittal will be required to confirm I the capability of the SRVs to meet the ASME Code requirements for reactor vessel overpressure protection. The ability of the SRVs to accommodate pressure increase transients  ;

under power uprated conditions has been discussed in Section 3.8 of this staff position.  ;

increasing SRV set points will involve increases in the peak reactor coolant system pressure f and temperature. These increased parameters will result in higher differential pressures and flow rates for the operation of safety- related valves in systems connected directly to the RCPB. '

These systems include the HPCI, HPCS, RCIC and RWCU (reactor water cleanup) systems.

Licensees will need to demonstrate that the safety-related valves in these systems are capable of operating against the worst-case design-basis conditions in light of the more severe l pressures and temperatures. Because of the inability to test these valves under differential pressure and flow conditions in the plant, sufficient margin must be provided in evaluations of the capability of these valves. {

1 The SRVs have design pressures of 1250 to 1375 psig (plant dependent). A 75 psiincrease in normal operating pressure should not impact their analyzed structural integrity. Similarly, the increase in saturated steam temperature should also be within their analyzed design conditions.

The increase in valve opening pressure may increase discharge line and containment loads during valve actuation greater than previously analyzed for some plants.

l The SRVs were evaluated under uprated power conditions for the im,7act of 24-percent higher l steam flow, a ternperature increase of approximately 10 *F, and a pressure increase of 75 psi or less from the plant's original design-basis condition.

An increase in steam flow does not directly affect SRVs, which are normally closed during plant operations. However, the potential for flow-induced vibration may exist because of the increased flow rate which may cause the incidence of seat leakage. The impact of the increased pressure and temperature on the structural integrity of SRVs is minimal because changes in these parameters are very smallin comparison to the original design specifications l of the valves.

To maintain an adequate simmer margin (the difference between the valve opening set pressure and normal operating pressure), the lowest set points are normally increased by the same amount as the operating pressure increase. In most cases, the highest pressure set point of SRVs is not changed. By not changing the highest spring set point, the previous analysis of containment loads remained bounding for the power uprate.

l l For some plants, the increase in SRV opening pressure will increase the discharge line and

! containment loads during the valve actuation that may be greater than those used in the i

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17 design-basis analysis. The effects of increased discharge line and containment loads should be evaluated on a plant-specific basis in accordance with ELTR1. The increase in SRV set points willinvolve increases in the peak reactor coolant system pressures and temperatures.

Therefore, licensees will need to demonstrate that the SRVs are capable of operating in the more severe pressures and temperatures that may exceed the design-basis condition.

4.7 Main Steam Isolation Valves (MSIVs)

MSIVs provide a part of the RCPB and a number of safety functions, including the isolation of the reactor from the environment during postulated accident scenariosJ Hence, the current licensing-basis requirements for the MSIVs must continue to be met under uprated conditions.

The impact of increased pressures and temperatures on the MSIVs must be evaluated with respect to the capability of the MSIV to meet pressure boundary structural requirements, and the capability of the MSIV to meet the required safety functions.

The RCPB requirements of the MSIVs, such as closure time and leakage limits, will continue to be monitored by various surveillance requirements in the plant TSs to ensure that the origina; licensing basis for the MSIVs is preserved.

The increased normal operating steam flow associated with power uprate will increase the force ,

in closing the MSIVs. This increased steam flow could increase the impact loading of the l valves when closing from the uprated flow conditions. However, the MSIVs are designed to withstand the closure impact from 200 percent of the original rated steam flow. This upper flow value is based on the maximum dome operating pressure and the steam flow limiting venturi.

The venturi size is not affected by power uprate. A 120-percent power uprate with a pressure increase of 75 psi will result in 173 percent of steam flow. The increase in maximum hydraulic pressure due to this higher flow rate is estimated to be less than the cylinder design pressure.

This will be confirmed on a plant-specific basis for each extended power uprate. The higher operating pressure at power uprate conditions tends to result in higher seating forces during isolation, thus reducing leakage. MSIVs are under close scrutiny for leakage and closure time from various surveillance requirements in the TSs, and their safety performance is routinely monitored. The Class 1E components such as the MSIV limit switches and the solenoid valves could be affected by the higher operating temperature due to power uprate. The design conditions for these Class 1E components are expected to bound the power uprate conditions, and any potential impact on the environmental qualification of these components should not be significant. This will be confirmed on a plant-specific basis to assure that potential accident conditions remain bounded.

4.8 Pinino Evaluation ELTR2 presented a piping evaluation for a BWR/5 plant at 120 percent of the originally licensed power level. The percent increases in flow rate, temperature, and pressure were aetermined from the plant-specific licensing-basis documents. The analyses were performed for thermal expansion, interface loads, and vibration effects for the main steam and recirculation piping.

The increase in piping stresses, cumulative fatigue usage, interface loads applying to supports and equipment were determined. These increases are presented for BWR/5 main steam and recirculation piping (Ref. 2).

i 18 1

i The components evaluated include equipment nozzles, anchors, guides, penetrations, valves,

! flange connections, and pipe supports. Results of the evaluation were compared to the i- allowable limits in the original code of record such as ASME Code Section Ill. No new 4

assumptions were introduced that were not included in the original analyses. In cases where the Code allowables are not satisfied, detailed analyses or field modifications can be completed

such that Code requirements are met. Pipe break locations and pipe whip restraint hardware capacities are also evaluated to demonstrate acceptability.

3 On the basis of the above review, the staff concludes that although the method for the evaluation is consistent with Appendix K of ELTR1, the adequacy of affected piping, piping

, components, and their supports will be dependent on the plant-specific design and as-built information, to demonstrate the structural and pressure boundary integrity of the reactor coolant

piping and supports for the power uprate condition.

5.0 IMPACT ON SAFETY MARGINS l _ELTR2 provides a generic assessment of the impact of extended power uprate on plant safety j margins. The impact of power uprate on plant safety margins is assessed generically in the

following categories

fuel thermal limits, design-basis accidents, transient analyses, and 1

environmental consequences. Although this generic discussion is not sufficiently detailed to exclude the need to assess the impact of power uprate on a plant-specific basis. it does provide insight on the impact that the expected changes will have on the current safety margins. The underlying philosophy of the generic BWR power uprate program has been that any changes to plant safety margins resulting from power uprate must be acceptably small. Plant-specific uprate submittals will also address the impact of power uprate on plant safety margins.

5.1 Fuel Thermal Limits ELTR2 indicates that no change is required to the basic fuel design to achieve the uprated power level or to maintain the margins. No increase in the allowable peak bundle power has been requested. A slightly flatter radial power distribution may be utilized to supply the additional power and maintain limiting fuel bundles within current licensing constraints. The fuel operating limits, such as maximum average planar linear heat generation rate and operating limit MCPR, will still be met at the uprated power level. The plant-specific submittal will confirm the acceptability of these operating limits as determined for uprated power conditions. Reload analyses will continue to meet acceptable NRC criteria as specified in GESTAR. New fuel designs, if needed in the future, will need to meet NRC-approved acceptance criteria. GE fuel will continue to meet the criteria accepted by the NRC as specified in NEDO-31908, " Licensing Criteria for Fuel Designs," January 1991.

5.2 Desian-Basis Accidents Plant-specific analyses will continue to demonstrate the ability of each plant to cope with the full spectrum of hypothetical pipe break sizes, from breaks as small as instrument lines to breaks in the largest recirculation, steam, feedwater and ECCS lines. These analyses will address both high and low energy line breaks, and the success of plant systems in dealing with the breaks while accommodating a single active equipment failure, in addition to the break. Challenges to

. l l

l 19 the fuel and containment, as well as potential radiological releases to the environment, will be assessed on a plant-specific basis using NRC-approved methods.

5.3 Transient Evaluations The effects of plant transients will be analyzed against the safety limit minimum critical power ratio (SLMCPR) which will be established using NRC-approved procedures, as described in Section 3.4 of this staff position. The SLMCPR will be confirmed for each plant requesting a power uprate. Where extended power uprate results in a greater number of bundles operating near the limit, the SLMCPR may be increased to provide the same statistical confidence level that fuel rods will avoid boiling transition. When establishing the operating limit MCPR, transient events will continue to be analyzed against the SLMCPR using NRC-approved procedures.

This operating limit MCPR will be documented in each plant-specific uprate submittal and I l confirmed for each cycle of operation in the cycle-specific reload analysis.

l

6.0 CONCLUSION

l GE has provided analytical evaluations, evaluations of components and equipment, evaluation

( of generic communications applicable to power uprate, and a discussion of the impact of power uprate on licensing criteria to demonstrate a feasibility of extended power uprate (up to 20 percent increased core thermal power) in ELTR2, and its supplement. Based on its review, the staff has determined that ELTR2 in conjunction with the generic guidelines provided in ELTR1 ,

and a plant-specific licensing evaluation would provide the information necessary for staff review of individual application for extended powar uprate. Many topic areas discussed in the L ELTR2 will require either additional plant-specific information or verification before the staff will be able to complete its evaluation of those topics. Requirements for additional plant-specific information needed to complete staff review are documented throughout this staff position.

Those topic areas will be resolved in individual plant specific safety evaluations supporting extended power uprate.

l Principal Contributors: R. Frahm, Sr.

A. Ulses C. l. Wu H. Garg l C. E. Carpenter L

R. Goel j Date: September 14,1998 i

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7.0 REFERENCES

1. General Electric Company (GE), Licensing Topical Report No. NEDC-32523P, " Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate" (ELTR2),

dated March 1996. (Proprietary and nonproprietary reports available.)

2. GE, Licensing Topical Report No. NEDC-32523P, " Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate - Supplement 1," Volumes 1 and 2, dated June 1996. (Proprietary information. Not publicly available.)

3. Letter from W. Marquino (GE) to T. J. Kim (NRC), "NRC Request for Additional Information (RAls) on GE Licensing Topical Report NEDC-32523P, Generic Evaluations of General Electric Boiling Water Reactor Extended Power Uprate," dated July 2,1997.

(Proprietary information. Not publicly available.)

4. GE, Licensing Topical Report NEDC-32424P, " Generic Guidelines of General Electric Boiling Water Reactor Extended Power Uprate," dated February 1995. (Proprietary and nonproprietary reports available.)

5. Letter from D. M. Crutchfield (NRC) to G. L. Sozzi (GE), " Staff Position Concerning GE BWR Extended Power Uprate Program," TAC No. M91680, dated February 8,1996.

6. Letter from W. T. Russell (NRC) to P. W. Marriott (GE), " Staff Safety Evaluation of General Electric Boiling Water Reactor Power Uprate Generic Analyses," TAC No.

M81253, dated July 31,1992.

7. GE, Licensing Topical Report NEDC-31336P, "GE Instrument Setpoint Methodology,"

dated October 1986. (Proprietary information. Not publicly available.)

8. Letter from B. Boger (NRC) to D. J. Robare (GE), " General Electric (GE) Tropical Report NEDC-31336, ' General Electric Instrument Setpoint Methodology'," dated February 9, 1993.

9. NRC Bulletin 88-07 (and Supplement 1), " Power Oscillations in Boiling Water Reactors,"

June 15,1988 (December 30,1988).

10. GE, Licensing Topical Report NEDE-24011-P-A-10-US, " General Electric Standard Application for Reactor Fuel (GESTAR)," U.S. Supplement, dated March 1991.

(Proprietary Information. Not publicly available.)

11. GE, Licensing Topical Report NEDC-24154P," Qualification of the One-Dimensional Core Transient Model (ODYN) for Boiling Water Reactors," dated December 1997.

(Proprietary information. Not publicly available.)