ML20151S381

Feedwater Heater(S) Out-of-Service Analysis for River Bend Station

ML20151S381
Person / Time
Site:	River Bend
Issue date:	05/31/1988
From:	Kim H, Nichols E, Sozzi G GENERAL ELECTRIC CO.
To:
Shared Package
ML20151S354	List: ML20151S360 ML20151S381 RBG-28399, Application for Amend to License NPF-47,clarifying License Condition 2.C(13) & Revising Tech Spec Table 3.3.6-2, Item 1.b, High Power Setpoint. Fee Paid
References
NEDO-31583, NUDOCS 8808150178
Download: ML20151S381 (29)

Text

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l NE00 31583 DRF A00 03243 Class I May 1988 Feedwater Heater (s) Out-of-Service Analysis For River Bend Station GE NuMr Energy a ".;! , .

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NEDO-31583 DRF A00-03243 Class I May 1988 FEEDWATER HEATER (S) OUT-OF SERVICE ANALYSIS FOR RIVER BEND STATION H. T. Kim Technical Project Engineer i

Approved: Reviewed : M

,G. L. Sod /I, $t/ager E. E. Nichols Plant PetformMnce Engineering Plant Licensing Services 1

I GENuclearEnergy 175 Cunner Arnue SanJose. CA 95125 l

t NEDO-31583 IMPORTAIIT NOTICE REGARDING

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CONTENTS OF THIS REPORT Please Read Carefully The only undertaking of General Electric Company respecting information in this document are contained in the contract between Gulf States Utility (GSU) and General Electric Company, as identified in the purchasing agreement for this work and nothing contained in this document shall be construed as chang-ing the contract. The use of this information by anyone other than GSU or for any purpose other than that for which it is intended, is not authorized; and with respect to any unauthorized use, General Electric Company makes no representation or warranty, and assumes no liability as to the completeness,-

accuracy, or usefulness of the information contained in this document, 11

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-NEDO 31583 CONTENTS fA&ft

1. INTRODUCTION AND

SUMMARY

1-1 1.1 Introduction 1-1 1.2 Summary 1-2

2. TRANSIENT EVENT ANALYSIS '21 2.1 Abnormal Operating Transients 2-1 2.2 Rod Withdrawal Error 2-2 2.3 Operating Limit MCPR 2-2 2.4 Anticipated Transient Without Scram 2-3 2.5 Overpressurization Analysis 2-3

3. THERMAL-HYDRAULIC STABILITY ANALYSIS 3-1

4. IMPACT ON LOSS OF COO 1 ANT ACCIDENT AND REIATED ANALYSES 4-1 4.1 ECCS Thermal-Hydraulic Performance 4-1 4.2 Acoustic and Flow Induced Loads on Reactor Vessel Internals 41 4.3 Annulus Pressurization Loads 42 4.4 Containment Response 42

5. FEEDWATER N0ZZLE, SPARGER AND PIPING FATICUE USAGE 5-1 5.1 Feedwater Nozzle 5-1 5.2 Feedwater Sparger 5-1 5.3 Feedwater System Piping 5-1

6. REACTOR PROTECTION SYSTEM LOW POWER SETPOINT 6-1

7. REFERENCES 7-1 l

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NEDO 31583 TABLES Number Title Eagt 1-l' Items Potentially Affected by Operation with FWHOS 13 1-2 Operating Limit Changes 14 2-1 Initial Conditions for FWHOS Transient Analysis 2-4 2-2 Summary of Transient Peak Values Results 2-5 2-3 Summary of MCPR Results 26 ILLUSTRATIONS Figure Title Eagg 2-1 Cenerator Load Rejection With Bypass Failure,100% Power / 2-7 100% Core Flow, 320'F Rated Feedwater Temperature 2-2 Feedwater Controller Failure, 100% Power /100% Core Flow, 2-8 320*F Rated Feedwater Temperature iv l

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.O NEDO-31583 U

l. INTRODUCTION AND

1.1 INTRODUCTION

This report presents the results of a safety and impact evaluation which was performed to support continued operation of River Bend Station (RBS) with a number of feedwater heater (s) out of service (FWHOS). Continued operation with FWHOS is desirable in the event that certain feedwater heat-er(s) or string (s) of heater (s) become inoperable during a reactor fuel cycle.

Operational flexibility and plant capacity factor are improved if the plant is able to continue operating until full heating can be restored or until the next convenient outage occurs.

Design evaluations reported in the RBS Updated Safety Analysis Report (USAR) and the RBS Cycle 2 Reload Licensing Submittal (Reference 1) justify operation with full feedvater heating which corresponds to a rated feedwater temperature

• of 420*F. Operation with FWHOS will result in lower feedwater temperatures with increased subcooling in the core downcomer region and at the core inlet. Loss of feedwater heating from the highest pressure heaters would result in the highest temperature reduction. Loss of heating from the low pressure heaters would result in only a slight reduction of feedwater temperature, It is estimated that, at the worst, feedwater temperature loss during an operating cycle due to inoperable, out-of-service or unavailable heater stages is less than 100'F. Therefore, evaluations presented herein assume a 100*F reduction in temperature. Reference 1 has already evaluated the consequence of the transient with a sudden feedwater temperature loss o! 100*F when initiated from the 420*F rated feedwater temperature. This report will justify the continued operetion at the steady-state condition with FWHOS corresponding to a range of rated feedwater temperature from 420*F to 320*F during the operating cycle. Certain design bases were reevaluated to determine

• To simplify discussion on FVHOS operation, the term "rated feedwater tempera-ture" is used in this report to mean "feedwater temperature at 100% core thermal power (100% steam flow) and 100% core flow conditions." The feedwater temperature at any given core power and flow is dependent upon the combination of operable heater (s) in each of the two strings of heaters and the core power level.

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i NEDO-31583 Io the effects of WHOS operation. Table 1-1 lists potentially affected design evaluations along with a brief description of the technical issues that arise for WHOS operation.

1.2

SUMMARY

Evaluations were performed to justify WHOS operating conditions for a rated feedwater temperature range of 420'F to 320*F. Results of these evaluations, as discussed in this report, indicate that RBS is capable of safe operation with partial feedwater heating corresponding to a feedwater temperature reduction of up to 100'F below normal condition, provided that applicable limits are observed. Table 1-2 lists the limits that have changed as a result of this reduced feedwater temperature operation. The scope of this analysis does not include the coastdown portion of the fuel cycle.

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NEDO 31583 Table 1 1 ITEMS POTENTIALLY AFFECTED BY OPERATION WITH FWHOS 1112 Comment Transient Response" Increased core inlet subcooling can change calculated ACPR and peak pressures for limiting trannients.

Stability Margins

• Increased core subcooling can cause an increase in calculated decay ratio.

ECCS Thermal Hydraulic Increased mass release from a design basis Analysis" loss-of-coolant accident (LOCA) can cause changes in the reactor blowdown response.

Acoustic Loads During The increased subcooling in the reactor Postulated LOCA Events downcomer region could affect predicted acoustic loads on the shroud and jet pumps.

Annulus Pressurizatien Loads Increased mass release rates from a design During Postulated LOCA Events basis LOCA can cause increased loads on the reactor vessel.

Containment Responses and Drywell pressurization rate, and peak drywell Loads During Postulated LOCA pressure could be affected by higher blowdown Events release which could impact containment calculations.

Feedwater Nozzle, Sparger and Reduced feedwater temperature can affect Piping Fatigue fatigue usage.

"These items may ba cycle and fuel-type dependent. (The other items are generic and cycle-independent.)

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NEDO 31583 i Table 1-2 OPERATING LIMIT CHANGES 40-year Averago Number of Days Rated W Temperature between of Allowble Usage During an 420*F and 370'F - 256 Days Operating Year for WHOS Rated W Temperature between Operation without Exceeding the 370*F and 320'F - 61 Days Feedwater Sparget' Fatigue Usage Factor Limit 1-4

It s 1 NEDO-31583

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2. TRANSIENT EVENT ANALYSIS 2.1 ABNORMAL OPERATING 'RANSIENTS All core wide transients described in RBS USAR Chapter 15 And RBS Cycle 2 Reload Licensing Submittal (Reference 1) were examined for FWHOS operation.

Three limiting abnormal operating transients reported in Reference 1 were reevaluated in detail for the FWHCS operation:

a. Generator Load Rejection with Byptss Failure (LRBPF)

b. Feedwater Flow Controller Failure (FWCF)

c. Loss of 100*F Feedwater Heating (LFWH} p The reevaluations for the LRBPF and PVCF events were performed at 100%

power /100% core flow condition with a rated feedwater temperature of 320*F for Cycle 2. Reactor heat balance, core coolant hydraulics and nuclear transient parameter data vere generated and used in the transient analysis. The initial conditions for the FVHOS transient analysis are listed in Table 2-1. The GEMINI /0DYN precedure described in Reference 2 was used for analyses of both the LRBPF and FWCF transients. TS transient peak values and minimum critical power ratio (MCPR) results for the two cases analyzed are summarized in Tables 2-2 and 2-3 respectively, with the licensing values included for comparison. The transient responses are presented in Figures 2-1 and 2 2.

The results show that the FWCF ACPR for the FWHOS operation increases slightly compared to che licensing value, but is still below the Techn!ct ' Specfication OLMCP4.

The RBS plant specific analysis for the 100*F LFWH transient was perforced at 102% power /100% core flow condition with 320*F rated feedwater using the GE Three Dimensional BWR Core Simulator described in Reference 3.

The results show that the ACPP for the 100'F loss initiated from 320'F is 0.09 compered to 0.11 for 420'F initiation case, indicating that the 100'F LFVH is less savere if initiated from 320*F than 420*F. Therefore, the LFWH for FHWOS is adequately bounded by the normal feedwater temperature case.

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6 1 NED0 31583 l 2.2 ROD WITHDRAWAL ERROR l

The rod wichdrawal error (RWE) transient documented in USAR Chapter 15  !

was analyzed using a statistical evaluation of the minimum critical power ratio (MCPR) and Linear Heat Generation Rate (LHGR) response to the uithdrawal of ganged control rods from both rated and off-rated conditions over the entire operating region. Therefore, this analysis covers a wide variety of feedwater temperatures and core subcooling as different off-rated conditions are included in the database. The 95% probability 95% confidence values from this statistical database are used to develop the Rod Withdrawal Limiter (RWL) system setpoints to protect against a rod withdrawal error.

The rod withdrawal error analysis does not need to be evaluated for FWHOS at end of cycle because all control rods will be fully withdrawn. A RWE analysis was performed at 2000 MWD /T before end of equilibrium cycle to examine the effect of the initial feedwater temperature. An initial condition of 250'F was used to bound all FWHOS operation. Results show that ACPR values resulting from the worst two feet of withdrawal for the 420*F and 250*F feedwater temperature are identical. Therefore, the ACPR values initiating from 250*F feedwater temperature condition fall within the statistical data-base used to establish the RWL system setpoints. It is concluded that operating limit MCPR does not need to be increased due to RWE for FWHOS operation.

2.3 OPERATING LIMIT MCER The MCPR results for the FWHOS operation are summarized in Table 2-3.

The Cycle 2 licensing analysis results are included for comparison. The results show that although the FWCF MCPR increases slightly for the FWHOS ,

operation, it is bounded by the RWE MCPR which is not affected by the FWHOS operation. Based on these results, it is concluded that t current RBS Technical Specification (Reference 6) OLMCPR is adequate for FWHOS operation for a rated feedw ~ temperature range of 420'F to 320*F.

The off-rated power-dependent MCPR (MCPR ) limits are not affected by FWHOS operation since the full power MCPR limit is not affected by FWHOS operation.

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( NEDO 31583 The current off-rated flow-dependent MCPR (MCPR f

) limit curve is based on the steepest power / flow rod line to protect against the recirculation flow runout transient. A power / flow rod line was generated for the 320*F rated feedwater temperature. It shows that the slope of this rod line is bounded by the current design basis rod line. Therefore, the current MCPRg limits are valid for FVHOS operation.

2.4 ANTICIPATED TRANSIENT WITHOUT SCRAM (ATWS)

An evaluation was performed which shows that reducing feedwater tempera-ture helps to reduce the consequences of an ATWS event. Reduced feedwater temperature results !n a reduction of steam flow and core averaga void frac-tion. The lower steam flow rate is produced because more of the core heat is needed to heat up the colder coderator in the core. Therefore, less steam is generated at its rated power as feedwater temperature decreases.

It is concluded that if an ATWS event were initiated at RBS from the FWHOS operation conditions, the results would be less severe than if it were initiated from rated feedwater temperature at a20 F.

2.5 OVERPRESSURIZATION ANALYSIS Lower initial operating pressure and steam flow rate provide better overpressure protection for the most limiting Main Steam Line Isolation Valve event during FWHOS operation. Table 2-2 also indicates that the peak vessel pressures for the LRBPF and FWCF events analyzed for FVHOS are below those for all feedwater heaters operating case. Hence, it is concluded that the pressure barrier integrity is maintained under the FWHOS operation.

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l NEDO-31583 Table 2-1 i

INITIAL CONDITIONS FOR ,

r- WHOS TRANSIENT ANALYSIS Rated FW - 320*F

1. Thermal Power Level (!CJt) 2894

2. Steam Flow (1b/sec) 3049

3. Core Flow (M1b/hr) 84.5

4. Feedwater Flow Rate (1b/sec) 3049

5. Feedwater Temperature (*F) 320

6. Vessel Dome Pressure (psig) 1005

7. Core Exit Pressure (psig) 1015

8. Core Coolant Inlet Enthalpy (Btu /lb) 514.2

9. Turbine Inlet Presaure (psig) 979

• The other input data used for the transient analysis are the same as those used in Reference 1.

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NEDO-31583 Table 2-2 1

SUMMARY

OF TRANSIENT PEAK VALUE RESULTS h

Peak Peak Peak Rated W Neutron Heat Vessel Temperature Flux Flux Pressure Transient ('F) (% NBR) (osir) ACPR b)

(% NBA).

.IRBPF 420(*) 286.2 107,8 1213 0.07 320 284.9 107.6 1192 0.07 WCF 420(*) 230.2 108.2 1202 0.06 320 253.8 110.6 1177 0.08

(# Licensing Basis (Reference 1)

( ) CPR based on an initial CPR which yields a safety limit MCPR of 1.07; uncorrected for ODYN adjustment.

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[ , NEDO 31583' Table 2-3

SUMMARY

OF MCPR RESULTS Cycle 2 Transient Licensine Basis EHHQ1 LRBPF 1.15 1.15 FWCF 1.14 1.16 LFWH 1.18 1.16 RWE 1.18* 1.18

• The RBS Cycle 2 Licensing Basis OLMCPR 2-6 l l

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NEDO-31583 I NEUTRON FLU ( l YESSEL PRE 5h RISE (PSI) 2 AVE $URFACE HEAT FLyX 2 SAFETY VAL){ FLQw 3 CCRE INLET TLOW 3 RELIEF VALVE Flow 13 0. 0 3 50.9 i e<*Me v "; E e r_ ~

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Figure 2-1. Plant Response to Generator Load Rejection with Bypass Failure, 100% Power /100% Core Flow, 320'F Rated Feedwater Temperature 2-7

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Figure 2-2. Plant Response to Feedwater Controller Failure, 100% Power /100% Core Flow, 320*F Rated Feedwater Tersperature i

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3. THERMAL HYDRAULIC STABILITY ANALYSIS l

-General Design Criterion 12 (10CFR50, Appendix A) states that power f oscillations which result in exceeding specified acceptable fuel design limits be either not possible or can be readily and reliably detected and suppressed.

Historically, compliance :o GDC-12 was demonstrated by assuring that neutron flux oscillations would r at occur. This eliminated the need to perform fuel integrity calculations under limit cycle conditions. As a result of stability tests at operating BWRs and extensive development and qualification of GE analytical models, stability criteria have been developed which also demon-strate compliance to GDC-12. Reference 4 provides these stability compliance criteria for GE fueled BWRs operating in the vicinity of limit cycles. The NRC has reviewed and approved this in Reference 5; therefore, a specific analysis for each cycle is not required.

Operation in the FWHOS mode is bounded by the fuel integrity analyses in Reference 4. In general, the effect of reduced feedwater temperature results in a higher initial CPR, which yields even larger margins than those reported in Reference 4. The analyses are independent of the stability margin, since the reactor is already assumed in limit cycle oscillations. Reference 4 also demonstrates that for neutron flux limit cycle oscillations just below the 120% neutron flux scram setpoint, fuel design limits are not exceeded for those GE BWR fuel designs contained in General Electric Standard Application for Reactor Fuel (GESTAR, Reference 5). These evaluations demonstrate that substantial thermal / mechanical margin is available for the GE BWR fuel designs even in the unlikely event of very large oscillation.

To provide assurance that acceptable plant performance is achieved during operation in the least stable region of the power / flow map, as well as during all plant maneuvering and operating states, a generic set of operator recom-mendations has been developed and communicated to all CE BWRs. These recom-mendations instruct the operator on how to reliably detect and suppress' limit cycle neutron flux oscillations should they occur. The recommendations were developed to conservatively bound the expected performance of all current ,

I product lines.

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When operating in FWHOS mode during a cycle, the colder feedwater flow increases the core inlet subcooling and will also result in power distribution changes. These changes result in reduced stability margin when operating in the high-power / low-flow region of the operating domain. Tests performed at an overseas BWR/6 in October 1984 evaluated the effects of reduced feedwater temperature during a cycle on stability margins. It was determined that the

! reduction in stability margin is within the conservative basis of the operator recommendations and, therefore, the recommendations are applicable for FWHOS during the cycle.

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1. NEDO-31583

4. IMPACT ON LOSS-OF-C001 ANT ACCIDENT AND REIATED ANALYSES 4.1 ECCS THERMAL-HYDRAULIC PERFORMANCE l

A Loss-of-Coolant Accident Analysis (LOCA) was performed for RBS with FWHOS operation. Reduction of feedwater temperature results in increased subcooling in the vessel, thus increasing the mass flow rate out of a LOCA break. However, an increase in initial total system mass and a delay in lower plenum flashing also occur. They act together to decrease the impact of increased flow out of the recirculation line break. As a result of this offsetting effect, the peak cladding temperature was shown to be lower than the 2144*F value reported for RBS and below the 2200'F 10CFR50.46 cladding temperature limit.

4.2 ACOUSTIC AND FLOW-INDUCED LOADS ON REACTOR VESSEL INTERNALS The acoustic loads are lateral loads on the vessel internais that result from propagation of the decompression wave created by a postulated recircula-tion suction line break. The acoustic loading on the vessel internal is proportional to the total pressure wave amplitude in the vessel recirculation outlet nozzle. FWHOS increases subcooling in the downcorer. This results in a lower saturation pressure, thereby having a larger total pressure amplitude and resulting in larger acoustic loads.

The flow-induced loads are additional lateral loads on the vessel inter-nals that result from high velocity flow in the downcomer in a postulated recirculation line break. These loads are proportional to the square of the critical mass flux rate out of the break. Higher subcooling in the downcomer under FWHOS increases the critical flow and flow-induced loads.

l The reactor internals most impacted by acoustic and flow-induced loads under FWHOS operation are the shroud, shroud support and jet pump. The impact l on these components was evaluated throughout the FWHOS operating power flow region. The analyses concluded that these components have enough design margin to handle the loading during FWHOS.

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1. hEDO 31583 0

4.3 ANNULUS PRESSURIZATION (AP) IDADS A study has been perfoirsd to assess the impact of FWHOS operation on the annulus pressurization (AP) loads for River Bend. A review of RBS USAR Figures 6.2-39 through 6.2-55 indicates that the feedwater line break results in the l greatest forces upon the RPV and the greatest pressure differentials across the biological shield wall. Therefore, an evaluation of the feedwater line break flow has been performed in this study. The break flow for the feedwater line break with FWHOS were determined to be less than those presented in the USAR during the inventory depletion period of the feedwater line when the peak AP loads occur. Therefore, the normal operation AP loads calculated in the RBS USAR bound those expected to result under FWHOS operation.

4.4 CONTAINMENT RESPONSE The impact of FWHOS on the containment LOCA response was evaluated. Both main steamline break and recirculation line break were analyzed over the FWHOS operation power / flow region. The peak drywell and wetwell pressure and temperature, pool swell, condensation oscillation and chugging load during FWHOS operation were evaluated.

The peak drywell-to-contaiment differential pressure during the FWHOS operation occurred under recirculation line break at the maximum vessel subcooling condition in the power / flow map. This peak differential pressure increased by 0.2 psi compared to the design basis accident main steamline break; however, this differential pressure (18.8 psid) is still below the design differential pressure of 25 psid reported in USAR Section 6. Also, the peak pool swell, condensation oscillation and chugging loads evaluated during FWHOS operation vary slightly over the peak values presented in USAR Section

6. However, the analysis concluded that the variation is insignificant and there is enough design margin to handle these loads during FWHOS operation, i

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5. FEEDWATER N0ZZLE, SPAD.GER AND PIPING FATIGUE USAGE 5,1 FEEDWATER N0ZZL2 An evaluation was performed on the feedwater nozzle in RBS for FWHOS j operation. Assuming a full, single 18 month cycle operation with feedwater heater out of service based on an 80% capacity factor would result in 438 full power days operation per cycle. This will result in an additional 0.0214 fatigue usage factor over 40 years of continuous FWHOS operation. Thus, the fatigue usage factor; will still be less than 0.8, which is below the limit of 1.0, 5.2 FEEDWATER SPARGER An evaluation was performed to examine the impact.of FWHOS operation on the feedwater sparger for RBS. Two cases were anaiyzed to determine the number of days allowable per year (for 40 years) for FWHOS operation without exceeding the feedwater sparger fatigue usage factor limit of 1.0. The results show that the 40-year average number of days allowable during an operating year for FWHOS operation decreases with lower feedwater temperature; 256 days and 61 days for rated feedwater temperatures of 370*F and 320*F, respectively.

5.3 FEEDWATER SYSTEM PIPING A standard stress analysis was performed on the feedwater system piping up to che first feedwater guide lug outside the containment for feedwater temperature at 250*F to bound the 320*F rated feedwater temperature caso.

Results of the study show that with the additional FWHOS operations, the feedwater piping fatigue usage factor still meets the allowable limit of 1.0.

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6. REACTOR PROTECTION SYSTEM LOW POWER SETPOINT I At reactor power levels where significant amounts of steam are being

-generated the fast closure of turbine stop or control valves will result in rapid reactor vessel pressurization. When pressure increases, power increases especially if the bypass valves fail to open. For this reason, scram occurs on turbine stop valve position and control valve fast closure to provide margin to the core thermal-hydraulic safety limit.

However, at sufficiently low initial thermal power levels, steam flow is within the turbine bypass system capability and only a mild core transient occurs without a need for automatic shutdown. Therefore, automatic shutdown on stop valve closure and fast control valve closure is bypassed at low power.

On BWR/6 systems the required lower bound for stop valve and control valve fast closure scram is 40% of rated thermal power. Turbine first stage pressure (TFSP) is the parameter used to initiate the turbine stop valve closure scram bypass functions. At normal feedwater heating operating conditions, this 40%

power is equivalent to approximately 30% of the TFSP (in psia) that would exist at turbine valves wide-open steam flow conditions. Below 40% power, the turbine stop valve or control valve scram functions are disabled. At these low power levels, high neutron flux scra , and vessel pressure scram and other scram functions are sufficient to provide the safety limit margin even with stop valve or control valve sudden closures.

! Under operation with reduced feedwater temperature the relationship between vessel steam flow (and therefore TFSP) and core thermal power changes.

Less steam flow is generated at the same thermal power and TFSP is reduced.

Therefore the effect of reduced feedwater temperature is to raise the scram bypass power level (by approximately three percent for 320*F rated feedwater

! temperature). Thus, it is necessary to review the turbine stop and control valve scram bypass setpoints to determine if adjustments are necessary for FWHOS operation to maintain the required 40% bypass setpoint.

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6 NEDO-31583 Conservatism in current RBS Technical Specification scram bypass TFSP nominal setpoint was assessed by comparing it to the RBS startup test data for "TFSP vs Reactor Power". The nominal TFSP setpoint corresponding to 40%

rated power (using measured plant data where TFSP is 754.6 psia at turbine control valves wide open) is estimated to be 235 psia for normal feedwater temperature case and 218 psia for 320'F rated feedwater temperature case. The RBS Technical Specification nominal setpoint is 191 psia. This means that the current Technical Specification setpoint is conservative in the scram bypass power level by approximately 4% of rated power compared to that actually required for 320'F rated feedwater temperature operation including all required uncertainties and allowances. Therefore, the existing conse-vatism in the Technical Specification setpoint justifies no adjustment of currer.t TFSP setpoint for the FWHOS operation.

It is concluded that the current TFSP setpoint has enough margin to accommodate effects of the FWHOS operation. Therefore, the RPS TFSP setpoints will remain adequate for safe operation of RBS with FWHOS operation.

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7. REFERENCES

1. Supplemental Reload Licensing Submittal for River Bend Station Reload 1, 23A5819, Rev. O, Class I, July 1987.

2. J.S. Charnley to H.N. Berkow (NRC), "Revised Supplementary Information Regarding Amendment 11 to GE Licensing Topical Report NEDE 240ll-P-A",

. January 16, 1986.

3. "Three-Dimensional BWR Core Simulator", NEDO-20953-A, January 1977.

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4. G.A. Watford, "Compliance of the General Electric Boiling Water Reactor

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Fuel Designs to Stability Licensing Criteria", General Electric Company, October 1984 (NEDE-22277 P-1).

5. "General Electric Standard Application for Reactor Fuel", General Electric Company, May 1986 (NEDE-240ll-P A-8).

6. Technical Specification for River Bend, NUREG 1172, Docket No. 50-458, Appendix A to License No. NPF-47, Amendment #12.

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l GE Nuclear Energy 175 Cunner Avenue San Jose. CA 95125