ML20041B130

Supplemental Reload Licensing Submittal for Peach Bottom Atomic Power Station Unit 2,Reload 5.

ML20041B130
Person / Time
Site:	Peach Bottom
Issue date:	12/31/1981
From:	GENERAL ELECTRIC CO.
To:
Shared Package
ML20041B124	List: ML20041B123 ML20041B126 ML20041B128 ML20041B130
References
Y1003J01A34, Y1003J1A34, NUDOCS 8202230304
Download: ML20041B130 (38)

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. _ . . .. . .. . . . ... .

~'O'#1 DECEMBER 1981 I

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM ATOMIC POWER STATION UNIT 2, RELOAD NO. 5

!!#"!!8!#4ggggg GEN ER AL h ELECTRIC

Y1003J01A34 l

Revision 0 Class I December 1981 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM ATOMIC POWER STATION UNIT 2, RELOAD No. 5 Prepared: 0[

J./D. Leaser Senior Licensing gineer Verified: s J./. Darnley Senior Licensing Engine Approved: ((

R. E. EngeY, ManagW Reload Fuel Licensing NUCLEAR POWER SYSTEMS DIVISION e GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNI A 95125 GENERAL $ ELECTRIC

Y1003J01A34 R;v. O IMPORTANT NOTICE REGARDING CONTENTS OF THIS REPORT j PLEASE READ CAREFULLY This report was prepared by General Electric solely for Philadelphia Electric Company (PECo) for PECo's use with the U.S. Nuclear Regulatory Commission (USNRC) for amending PECo's operating license of the Peach Bottom Atomic Power l Station Unit 2. The information contained in this report is believed by General Electric to be an accurate and true representation of the f acts known, obtained or provided to General Electric at the time this report was prepared.

The only undertakings of the General Electric Company respecting information in this document are contaicad in the contract between Philadelphia Electric Company and General Electric Company for nuclear fuel and related services for the nuclear system for Peach Bottom Atomic Power Station Units 2 and 3, dated j October 3,1973, and nothing contained in this document shall be construed as l changing said contract. The use of this information except as defined by said I

contract, or for any purpose other than that for which it is intended, is not authorized; and with respect to any such unauthorized use, neither General Electric Company nor any of the contributors to this document makes any repre-sentation or warranty (express or implied) as to the completeness, accuracy or usefulness of the information contained in this document or that such use of such information may not infringe privately owned rights; nor do they assume any responsibility for liability or damage of any kind which may result from such use of such information.

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Y1003J01A34 Rev. 0

1. PLANT-UNIQUE ITEMS (l.0)*

Rotated Bundle Loading Error Analysis for P8DRB285: Appendix A Fuel Assembly Rod Replacement: Appendix B I

Lead Test Assemblies Extended Exposure: Appendix C 8x8R Fuel Extended Exposure: Appendix D Developmental Channels: Appendix E Transient Analysis Code Revision: Appendix F

2. RELOAD FUEL BUNDLES (1.0, 2.7, 3.3.I and 4.0)

Fuel Cycle Designation Loaded Number Number Drilled Irradiated 8DRB284L 4 210 210 P8DRB284H 5 236 236 P8DRB285 5 40 40 LTA 2 2 2 New P8DRB284H 6 136 136 P8DRB285 6 16 16 P8DRB299 6 124 124 Total 764 764

3. REFERENCE CORE IDADING PATTERN (3.3.1) l Nominal previous cycle core average exposure at end of cycle: 17.9 GWd/T Minimum previous cycle core average exposure at end of cycle from cold shutdown considerations: 17.7 GWd/T

( Assumed reload cycle core average exposure at end of cycle: 18.3 GWd/T Core loading pattern: Figure 1

4. CALCULATED CORE EFFECTIVE MULTIPLICATION AND CONTROL SYSTEM WORTH - NO VOIDS, 20 C (3.3.2.1.1 and 3.3.2.1.2)

BOC k eff Uncontrolled 1.118 Fully Controlled 0.960 Strongest Control Rod Out 0.986 R, Maximum Increase in Cold Core Reactivity 0.003 with Exposure Into Cycle, Ak

• ( ) refers to areas of discussion in " General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-24011-P-A-2 and NEDO-24011-A-2, July 1981.

1

Y1003J01A34 Rev. 0

5. STANDBY LIQUID CONTROL SYSTEM SHUTDOWN CAPABILITY (3.3.2.1.3)

Shutdown Margin ( k) ppm (200C, Xenon Free) 660 0.04

6. REIDAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 and 5.2)a i

EOC-2 EOC GWd/T Void Coefficient N/Ab (C/% Rg) -7.8/-9.7 -8.5/-10.6 Void Fraction (%) 39.8 39.8 Doppler Coefficient N/Ab (gfoF) -0.23/-0.22 -0.22/-0.21 Average Fuel Temperature (OF) 1296 1296 Scram Worth N/A ($)c l

Scram Reactivity vs Timec

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7. REIDAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2)

Peaking Factors Fuel Exposure (Local, Radial, Bundle Power Bundle Flow Initial Design (CWd/T) Axial) R-Factor (MWt) (103 lb/hr) MCPR 8x8R/ EOC 1.20, 1.47, 1.05 6.18 110 1.32 LTA 1.40 EOC-2 1.20, 1.52, 1.05 6.42 108 1.27 )

1.40 P8x8R EOC 1.20, 1.44 1.05 6.06 111 1.35 (

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t EOC-2 1.20, 1.50 1.05 6.31 109 1.29 i 1.40 l

aApplies to REDY analyzed events only b

N = Nuclear Input Data A = Used in Transient Analysis C

Generic, exposure independent values are used as given in " General Electric Boiling Water Reactor Generic Reload Fuel Application," NEDE-24011-P-A-2, July 1981 2

> Y1003J01A34 Rev. 0

) 8. SELECTED MARGIN IMPROVEMENT OPTIONS (5.2.2)

Transient Recategorization: No Recirculation Pump Trip: No Rod Withdrawal Limiter: No Thermal Power Monitor: No Measured Scram Time: No Exposure Dependent Limits: Yes Exposures Analyzed (CWd/T): EOC EOC-2

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

^

Exposure Range $ Q/A ACPR 8x8R Transient (GWd/T) (Z NBR) (%) /LTA P8x8R Figure Load Rejection EOC 721 128 0.25 0.28 3a without Bypass EOC-2 708 124 0.20 0.22 3b Loss of 100 F BOC to 125 124 0.15 0.15 4 Feedwater Heating EOC Feedwater EOC 374 125 0.20 0.22 Sa Controller Failure EOC-2 272 121 0.14 0.16 Sb

10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE)

TRANSIENT

SUMMARY

(5.2.1)

Limiting Rod Pattern: Figure 6 I

Includes 2.2% Power Spiking Penalty: Yes Rod Position Rod Block (Feet ACPR MLHGR (kW/ft)

Reading Withdrawn) 8x8R/P8x8R/LTA 8x8R/P8x8R/LTA 104 3.5 0.09 16.4 105 4.0 0.11 17.0 106 4.5 0.12 17.3 107** 5.5 0.16 17.7

[ 108 6.5 0.19 17.7 109 9.0 0.22 17.7 110 12.0 0.23 17.7

• Indicates set point selected 3

l Y1003J01A34 Rav. 0

11. CYCLE MCPR VALUES (5.2)

Exposure Range (GWd/t) Pressurization Events Option A Option B BOC to EOC-2 (8x8R&LTA/ 8x8R&LTA/

P8x8R) P8x8R Load Rejection w/o Bypass 0.26/0.28 0.05/0.07 Feedwater Controller Failure 0.19/0.21 0.13/0.15 EOC-2 to EOC I Load Rejection w/o Bypass 0.31/0.34 0.19/0.22 Feedwater Controller Failure 0.26/0,28 0.19/0.21 BOC to EOC Non-Pressurization Events 8x8R&LTA/P8x8R/P8DRB285 Loss of Feedwater Heating 0.15/0.15/0.15 Rotated Bundle Error NA/0.14/0.22*

Rod Withdrawal Error 0.16/0.16/0,16 1

12. OVERPRESSURIZATION ANAIASIS

SUMMARY

(5.3)

P,1' Pv Plant Transient (psig) (psig) Response MSIV Closure 1244 1273 Figure 7 (Flux Scram)

13. STABILITY ANALYSIS RESULTS (5.4)

Rod Line Analyzed: 105% Rod Line Decay Ratio': Figure 8 l Reactor Core Stability Decay Ratio, x2/ *0: 0.85 Channel Hydrodynamic Performance Decay Ratio, x2/ *0 8x8R/P8x8R Channel: 0.29 (

)

14. ROTATED BUNDLE ERROR RESULTS (5.5.4)*

Variable Water Gap Misoriented Bundle Analysis: Yes Includes 2.2% Power Spiking Penalty: Yes Initial Resulting Resulting 1 MCPR MCPR LHCR (kW/ft) 1.20 1.08 17.6

• See Appendix A 4

t Y1003J01A34 Rev. 0

15. CONTROL ROD DROP ANALYSIS RESULTS (5.5.1)

Bounding Analysis Results:

Doppler Reactivity Coef ficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 i Scram Reactivity Functions: Figures 12 and 13 r

Plant Specific Analysis Results:

Parameters Not Bounded:

Scram Reactivity Functions: Cold and Hot Standby Resultant Peak Enthalpies (cal /g):

, Cold Hot Standby 165 231 4

16. LOSS-OF-COOLANT ACCIDENT RESULTS, NEW FUEL (5.5.2)

See " Loss-of-Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2," December 1977, NEDO-24081, as amended.

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Y1003J01A34 *tv. 0 02 06 10 14 18 22 26 30 59 36 55 18 14 24 51 26 8 47 18 10 8 8 43 26 8 28 39 14 8 30 35 36 8 28 0 31 24 8 30 Y1003J01 A34 Notes: 1. Rod Pattern Is 1/4 Core Mirror Synsnetric, Upper Left Quadrant Shown on Map.

2. Numbers Indicate Number of Notches Withdrawn out of 48. Blank Is a Withdrawn Rod.

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Y1003J01A34 Rev. 0 APPENDIX A ROTATED BUNDLE LOADING ERROR ANALYSIS FOR P8DRB285 A separate rotated bundle loading error analysis was performed for the P8DRB285 bundle with the results given below. ACPR results for the rotated bundle loading error event are given separately for this bundle type in Section 11. The ACPR for other events is the same as for the P8x8R bundle type, as given in Section 11.

1 Bundle Initial Resulting Resulting 1 Type MCPR MCPR LHCR (kW/ft)

P8 DRB285 1.28 1.08 17.7 P8x8R 1.20 1.08 17.5 21/22

( Y1003J01A34 Rev. 0 l APPENDIX B l

FUEL ASSEMBLY ROD REPLACEMENT B.1 INTRODUCTION During the Reload No. 5 refueling outage, six fuel rods will be removed from each of two previously irradiated fuel assemblies and replaced with fresh rods with U-235 enrichments as shown in Table B-1. The removed rods will be examined and punctured for fission gas pressure measurement. These rods will not be used during future operation. The average enrichment of the replacement rods is less than the initial enrichment of the rods they are replacing to compensate for fuel depletion. They were selected to assure that the reactivity and power peaking of the reconstituted assemblies will be similar to that of a non-reconstituted assembly. Consequently, the nuclear characteristics of the reconstituted assem-blies are essentially identical to non-reconstituted assemblies. The purpose of this appendix is to report the results of the analyses and safety evaluation for operation of the fuel assemblies af ter replacement of the fuel rods.

B.2 EVALUATIONS AND ANALYSES B.2.1 Nuclear and Thermal Parameter Evaluations Standard lattice physics calculations were made for the reconstituted assemblies, including simulation of the fresh rods. Over the exposure range of interest, the computed lattice reactivities of the reconstituted assemblies are on average within 0.03% oK of the non-reconstituted assembly reactivities. The maximum fuel rod power peaking values for the reconstituted assemblies are always less than 1% greater than the values for the non-reconstituted assemblies, and are lower than the power peaking for the non-reconstituted assemblies throughout most of their operation. Based on the small calculated changes to Km and local peaking caused by the replacement fuel rods, there will be a negligible effect on the nuclear and thermal performance of the reconstituted fuel assemblies.

B.2.2 Mechanical Design Evaluation The six replacement fuet rods in each of the two reconstituted assemblies are mechanically similar to the fuel rods which they are replacing and also to the standard fuel rods in the Reload No. 5 fuel bundles. The only mechanical dif fer-ence is a longer upper end plug on each replacement rod to accommodate the irradiation growth of the rods in the reconstituted assemblies. An analysis of differential rod growth in the reconstituted assemblies shows that the replace-ment fuel rods are mechanically compatible with the irradiated rods and thus will have no adverse effect on the safety analyses for Cycle 6 or subsequent cycles for Peach Bottom 2 (PB 2). The peak linear heat generation rates of the recon-stituted assemblies are still within the operating limit of 13.4 kW/f t which was used in evaluating the mechanical performance of the maximum duty fuel rod in Reload No. 5. Therefore, the results of the fuel rod thermal and mechanical design evaluations in NEDE 24011-P-A-2 are conservatively applicable to the recon-stituted assemblies.

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Y1003J01A34 Rev. 0 l

l B.2.3 Evaluation of the Effect of the Fresh Fuel Rods on PCT /MAPLHGR l The effect on MAPLHCR of the replacement of six exposed fuel rods with fresh rods in two PB 2 bundles has been evaluated. The reconstitution is conserva-tively estimated to increase peak clad temperature (PCT) by 14 F. Since the current maximum PCT is 1958, ti.is increase will result in a PCT of 1972 F, well below the 2200 F limit. Thus, there will be no change in MAPLHGR for the recon-stituted bundles.

The estimated PCT increase is based on:

1. an increase in average stored energy of the bundle due to the introduc-tion of fresh rods which have a higher stored energy, and

2. a slight shift in power to the center 16 rods due to the change in local peaking for the reconrtituted bundles.

B.2.4 Transient Analysis for Cycle 6 Based on the analysis results described in Section B.2.1 above, the transient analysis results contained in this submittal are unaf fected by fuel rod replace-ment of the two fuel assemblies.

B.3

SUMMARY

AND CONCLUSIONS It is concluded, based on the results of the evaluations and analysis described l

in Section B.2, that the accident and transient analyses of Cycle 6 are insigni-ficantly affected and the operating limits of Cycle 6 are unaffected by the recon-stitution of the two fuel assemblies. The operating MCPR limit is given in Section 11 of this submittal.

Table B-1 RECONSTITUTION ENRICHMENTS (wt % U235)

Rod Location Original Replacement A-1 1.3 1.3 H-1 2.0 1.7 B-2 2.4 2.0 C-2 2.4 2.0 D-4 3.95 3.3 F-6 3.95 3.8 24

Y1003J01A34 R;v. O t

APPENDIX C I

LEAD TEST ASSEMBLIES EXTENDED EXPOSURE C.1 PROPOSED PROGRAM The Peach Bottan 2 Lead Test Assenbly (LTA) fuel program is one of several pro-grams in the U.S. whereby lead burnup bundles are being extended to peak-pellet exposures greater than 44,000 mwd /MT (40,000 mwd /ST) but not to exceed 50,000 mwd /MT. Information from these programs will be used to systematically determine the impact on fuel reliability and weigh the advantages of extended exposures relative to other uranium utilization improvement methods.

The four LTAs (8DRB260) inserted into Peach Bottom 2 at the beginning of Cycle 2 (Reload 1) are currently the highest exposure 8x8R fuel assemblies in operation in any operating reactor. Previous to Cycle 5, these LTAs were licensed for operation during cycle 5 (Reference C-1). Inspections perfonned during the EOC 4 refueling outage confirmed the mechanical integrity of the LTAs for continued operation in Cycle 5. Two of the four LTAs have been proposed for further extended operation in Cycle 6. The program plan again includes the inspection of these assemblies to ascertain their mechanical integrity before Cycle 6 i operat ion.

C' . 2 FUEL MLCHANICAL DESIGN ANALYSIS Exposure-dependent fuel mechanical design analyses for the extended exposure LTAs have been performed for conditions which meet or exceed expected Cycle 6 operating conditions in Peach Bottom 2. Models, assumptions, and material properties used in these analyses are those documented in Reference C-2. Cal-culated results are given below.

C.2.1 Fuel Rod Thermal Analysis I

Safety evaluations are performed and measured against established safety criteria.

The consequence of calculating values which exceed such criteria is that fuel f ailure must be assumed to occur. For plant normal and abnormal operation, this is not permissible. Fuel failure is defined as a perforation of the cladding which would permit the release of fission products to the reactor coolant. The mechanisms which could cause fuel damage in reactor abnormal operational tran-sients are (1) rupture of tb tuel rod cladding due to strain caused by relative expansion of the UO2 Pelle' ai (2) severe overheating of the fuel rod cladding caused by inadequate cool' A value of 1% plastic stram on the Zircaloy claddng has been established as the safety limit below which fuel damage due to overstraining of the fuel clad-ding is not expected to occur. The fuel cladding integrity safety limit ensures j that fuel damage resulting fran severe overheating of the fuel rod cladding caused

\

by inadequate cooling is avoided. Of these criteria, only the linear heat gene-ration rate associated with the 1% plastic strain safety limit is affected by increased fuel exposures. Analyses performed for the extended exposure fuel bundles resulted in 1% plastic strain values of 16.1 kW/ft at a peak-pellet exposure of 55,000 mwd /MT (50,000 mwd /ST) for UO2 fuel rods and 16.8 kW/ft at 49,700 mwd /MT (45,100 Wd/ST) for urania gadolinia rods. Both values include the 2.2% power spiking penalty documented in Reference C-2. These results 25

Y1003J01A34 Rev. 0 assure that the same minimum margin to 1% plastic strain (175% of minimum steady-state power) reported in Reference C-2 is maintained. These linear heat generation rate values are used during specific evaluations of transients due to single operator error or equipment malfunction to ensure that the safety limit is not exceeded.

C.2.1.1 Fuel Cladding Temperatures The cladding surface temperature is caluculated using the cladding surface heat flux at a given axial position on a fuel element in conjunction with the overall cladding-to-coolant film coefficient. The models used are noted in Reference C-2. The inside, average, and outside cladding temperature during normal oper-ation at the end of Cycle 6 are calculated not to exceed 809 0F, 7730F, and 7380 F, respectively.

1 C.2.1.2 Fission Gas Release i

The amount of fission gas released during a time increment is calculated based i on the fission gas generated and fission gas release fraction. The calculated <

maximum fission gas release fraction in the extended exposure fuel rod with the m3st limiting peaking factors is less than the 25% noted in Reference C-2.

C.2.1.3 Incipient Center Melting i The fuel is designed so that . fuel melting is not expected to occur during normal steady-state full power operation which remains valid even at extended exposures.

Linear heat generation rates associated with incipient fuel center melting are greater than 125% of normal steady-state full power operation at EOC 6.

C.2.2 Fuel Assembly Mechanical Evaluations The fuel assembly is evaluated by analyses, tests, and experience to demonstrate fuel assembly structural integrity. . When analyses are used to demonstrate struc-tural integrity, resulting stress and/or strain levels are compared to the associ-ated mechanical limits documented in Reference C-2. Results of the fuel rod mechanical analyses of the n'rmal and transient loads .for the extended exposure fuel are given below. The results of the combined LOCA and seismic evaluation documented in Reference C-2 do not change.

C.2.2.1 Cladding Creep Collapse A cladding creep collapse evaluation was performed with the models documented in Reference C-2. Results of this evaluation demonstrate that cladding creep collapse is not expected to occur in the event of a maximum overpressure tran-sient throughout Cycle 6.

C.2.2.2 Stress Evaluations 1 1

Fuel rod stress analyses of the extended exposure LTAs were performed with the model documented in Reference C-2 for operation through Cycle 6. These analyses showed that the fuel design ratios were well below 1.0.

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) Y1003J01A34 Rev. O h

C.2.2.3 Deflection Evaluation The operational fuel rod deflections considered are a result of manufacturing t ole ranc e s , flow-induced vibration, thermal ef fects, and axial load. Deflections of the extended exposure LTAs were evaluated and compared to the fuel rod-to-r fuel rod and fuel rod-to-channel spacing deflection limits given in Reference l C-2. This comparison dr.n ,nstrated that the fuel rod clearance criterion was met.

C.2.2.4 Fatigue Evaluation The cyclic loads considered in cladding fatigue analysis are coolant pressure and thermal gradients. The analysis performed for the higher exposure LTAs was based on previous and projected operating cycles through the end of Cycle 6, maximum and minimum pressures, and the stresses determined in Subsection C.2.2.2. The cumulative fatigue damage was calculated to be less than the allowable fatigue limit.

C.2.3 Fuel Rod Corrosion, Hydriding and Fretting Wear Considerations C.2.3.1 Potential for Hydriding The potential for hydriding is discussed in Reference C-2 and is not affected by higher fuel exposures.

C.2.3.2 Fuel Element Energy Release Significant boiling transition is not possible at normal operating conditions or under conditions of abnormal operational transients because of the thermal margins at which the fuel is operated and the high fuel burnups. It can, there-fore, be concluded that the energy release and potential for a metal-water reac-tion is not an important consideration during normal operation or abnormal transients. The insignificant energy released in the event of boiling transition reported in Reference C-2 does not change because of the extended fuel exposures.

C.2.3.3 Fretting Wear and Corrosion As discussed in Reference C-2, no significant fretting wear or corrosion has been observed throughout a continuing fuel surveillance program. Increased exposures are not expected to significantly change the observed result s. It is expected that the LTAs will be visually examined before loading in Cycle 6.

C.3. IMPACT ON REIDAD ANALYSES All of the models documented in Reference C-2 are applicable for use with higher fuel exposures. However, some inputs into these models are exposure-depende,t and are reflected in calculated results. A description of these exposure-dependent changes is given below.

C.3.1 Nuclear Evaluations The nuclear evaluations are comprised of two analyses: lattice and core. Most of the lattice analysis is performed during the bundle design process. The results of these single bundle calculations are reduced to " libraries" of lattice reactivi-27

l Y1003J01A34 Rev. O ties, relative rod powers, and few group cross-sections as functions of instantan-cous void, exposure, exposure-void history, control state, and fuel and moderator i t empe ra t ur e. Because of the exposure dependence of these results, the libraries were expanded to include higher burnups as noted below. The core analysis is l unique for each reload. It is performed using the above lattice " libraries" l to demonstrate that the core meets all applicable safety limits. The effects )

of higher fuel exposures are thus reflected in the core analysis results through '

use of the expanded " libraries."

C.3.1.1 Reactivity 3

)

Traditionally, bundle reactivities have been expressed in terms of km (i.e., 3 the neutron multiplication of an infinite array of like bundles). This lattice reactivity is a function of lattice average enrichment, gadolinia loading, void fraction, hydrogen-to-uranium ratio, and exposure. Hot reactivity of the exten-ded exposure LTAs decreases by 0.05 Ak. from a lattice exposure of 38,000 to 50,000 mwd /MT (35,000 to 45,000 mwd /ST).

i C.3.1.2 Local Peaking Factors Por a given lattice at a given void fraction, the maximum local peaking factor will occur at different fuel rods as the exposure increases. This is due to the dif ferent depletion and generation rate of the various fissile nuclides in each fuel rod. Calculated maximum local peaking factor for the extended expo-sure LTAs increases by 0.06 from a lattice exposure of 38,000 to 50,000 mwd /MT (35,000 to 45,000 mwd /ST).

The local peaking factor does vary with void fraction, and this dependence is taken into account in the calculations used to assign local peaking factors to each axial segment of the fuel. The above values are for 0.40 void, as this is the typical average bundle void fraction.

C.3.1.3 Doppler Reactivity The Doppler coefficient is of prime importance in reactor safety. The Doppler coefficient is a measure of the reactivity change associated with an increase j in the absorption of resonance-energy neutrons caused by a change in the tem- q perature of the material in question. The Doppler reactivity coef ficient pro-vides instantaneous negative reactivity feedback to any rise in fuel temperature, on either a gross or local basis. Maximum and minimum calculated Doppler coef-ficients at several exposures are shown in Reference C-2.

C.3.1.4 Void Effect The most important of these effects is void reactivity. The overall void coef-ficient is always negative over the complete operating range, since the BWR design is undermoderated. The reactivity change due to the formation of voids results f rom the reduction in the number of neutrons slowing down due to the decrease in the watet-to-fuel ratio. Beyond 11,000 mwd /MT (10,000 mwd /ST), the void effect is essentially constant.

28

Y1003J01A34 Rev. O C.3.2 Steady-State Hydraulic Analysis Core steady-state thermal-hydraulic analyses are performed using a nodel of the reactor core, which includes hydraulic descriptions of orifices, lower tieplates, fuel rods, fuel rod spacers, upper tieplates, the fuel channel, and core bypass flow paths. Model details are documented in Reference C-2. The flow distribu-tion to the fuel assemblies and bypass flow paths is calculated on the assumption

! that the pressure drop across all fuel assemblies and bypass flow paths is the l same. An iteration is performed on flow through each flow path (fuel assemblies l and bypass paths), which equates the total differential pressure (plenum to plenum) across each path and matches the sum of the flows through each path to the total core flow. This analysis is insignificant 1y affected by extended exposure fuel.

C.3.3 Reactor Limits Determination Limits on plant operation are established to assure that the plant can be safely operated and not pose any undue risk to the health and safety of the public.

This is accomplished by demonstrating that the radioactive release from plants for normal operation, abnormal operational transients, and postulated accidents meets applicable regulations in which conservative limits are documented. This conservatism is augmented by using conservative evaluation models and observing limits which are more restrictive than those documented in the applicable regula-tions. These observed operating limits and methods used to determine if the limits are met are documented in Reference C-2.

C.3.3.1 Fuel Cladding Integrity Safety Limit The generation of the Minimum Critical Power Ratio (MCPR) limit requires a statis-tical analysis of the core near the limiting MCPR condition. Bounding statistical analyses have been performed which provide conservative safety limit MCPRs for operating BWR plants. These safety limit MCPRs conservatively apply for all reload cycles including equilibrium cycle. Insertion of low powered extended exposure LTAs does not change the conclusions of these bounding analyses.

C.3.3.2 MCPR Operating Limits The MCPR operating limit is established to ensure that the fuel cladding integ-rity safety limit is not exceeded for any moderate frequency transient. This operating requirement is obtained by addition of the absolute, maximum MCPR value for the most limiting transient from rated conditions postulated to occur at the plant to the fuel cladding integrity safety limit. Higher fuel exposures are reflected in the nuclear input data. However, due to the high exposure, these fuel assemblies will operate at significantly lower power levels than other 8x8R bundles and will not be near MCPR operating limits.

C.3.3.3 Vessel Pressure ASME Code Compliance To assure that the peak allowable pressure of 110% of the vessel design pressure is not exceeded, the most severe isolation event with indirect scram and credit for subsequent valve operation is evaluated. The event which satisfies this specification is the closure of all Main Steam Line Isolation Valves (MSLIVs) with indirect (flux) scrom, and the margin at extended exposures will not exceed the nominal end-of-cycle margin because of the reduced power levels. The model 29

l Y1003J01A34 Rev. O used to analyze this event is described in Reference C-2. The results of this analysis are not significantly affected by the LTA bundles.

C.3.3.4 Stability Analysis Two types of stability are examined utilizing a linearized analytical model. .

First, is the hydrodynamic channel stability of one or more types of channels operating in parallel with other channels in the core. Second, is the reacti-vity feedback stability of the entire reactor core which also involves power oscillations. The assurance that the total plant is stable and, therefore, has significant design margin is demonstrated analytically when the acceptable per-formance limit of a decay ratio less than 1.0 or a damping coefficient greater than 0.0 is met for each type of stability. These criteria must be satisfied for both usual and unusual operating conditions of the reactor that mcy be en-countered in the course of BWR plant operation.

The analysis is performed using the models documented in Reference C-2 at the most limiting condition, which usually occurs near the end of cycle, with power peaking toward the bottom of the core. The most sensitive reactor operating condition is that corresponding to natural circulation flow and a power level equal to or greater than the rated rod pattern power level. Extended exposures for the LTAs are reflected in the nuclear characteristics used in the analysis.

C.3.3.5 Accident Evaluations Accidents are events which have a projected frequency of occurrence of less than once in every 100 years for every operating BWR. The broad spectrum of postulated acciderts is covered by six categories of design basis events. These events are the control rod drop, main steam line break, loss-of-coolant, refueling, recirculation pump seizure, and fuel assembly loading accidents. Consequences of these events with the low-powered extended exposure LTAs are not as great .

as lower burnup bundles. However, the MAPLHGR values for the test bundles have been extended to an average planar exposure of 55,000 mwd /MT (50,000 mwd /ST).

These new MAPLHCR values and associated peak cladding temperatures and oxidation fractions were incorporated into Reference C-3.

C.4 REFERENCES l C-1. Supplemental Reload Licensing Submittal for Peach Botton Atomic Power Station Unit 2, Reload No. 4, NEDO-24237, February 1980.

C-2. Generic Reload Fuel Application, NEDE-24011-P-A-2, August 1981.

C-3. Loss of Coolant Accident Analysis for Peach Botton Atomic Power Station Unit 2, NEDO-24081. December 1977, including E&A sheet 6 of November 1981.

30

f Y1003J01A34 R v. O

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APPEND 1X D 8x8R FUEL EXTENDED EXPOSURE h

During Cycle 6, the 8DRB284L fuel (which was inserted at BOC4) is expected to attain a peak pellet exposure in excess of 40,000 mwd /ST (44,000 mwd /MT) lut not to exceed 50,000 mwd /MT. Thermal and mechanical analyses have been performed for this fuel type in accordance with the NRC approved methods described g

in Reference D-1 to an exposure of 50,000 mwd /MT. Results of those analyses cre within the applicable criteria of Reference D-1. Results of the analyses for the linear heat generation rates associated with 1% plastic strain and incipient center melting indicate that the LHGR values for P8x8R fuel in Table 2-3a and Table 2-4 of Reference D-1 for 50,000 mwd /MT peak pellet exposure cre applicable to the 8DBR284L fuel.

i l

t t

I

)

REFERENCE D-1. General Electric Boiling Water Reactor Generic Reload Fuel Applicat. ion, NEDE 24011 - P-A-2, July 1981.

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l 1

i APPENDIX E I

DEVEIAPMENTAL CHANNELS E.1 ANALYSES The analyses given in Reference E-1 are applicable to continued use of developmen-tal channels. The location and exposure of the developmental channels has changed.

Ilowever, the thermal-hydraulic, nuclear, and safety analyses presented in the main body of this submittal are applicable to the continued use of developmental luel channels.

E.2 REFERENCES E-l. Developmental Channels Supplemental Information for Reload 1 Licensing Submittal for Peach Bottom Atomic Power Station Unit 2, NEDO-21172, l

Rev. 1, Supplement 2, March 1976.

I r

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Y1003J01A34 RIv. 0 l APPENDIX F TRANSIENT ANALYSIS CODE REVISION f F.1 CODE The pressurization transient events reported in this submittal were analyzed with the ODYN M04 transient analysis code, which is a revision of the ODYN code discribed in Reference F-1. A description of this revised code and a comparison to the previous code are given in References F-2 and F-3.

F.2 REFERENCES F-1. General Electric Boiling Water Reactor Generic Reload Fuel Applica-tion, NEDE-24011-P-A-2, July 1981.

F-2. Letter, J. F. Quirk (GE) to P. S. Check (NRC), ODYN Improvements, September 25, 1981.

, F-3. Letter, J. F. Quirk (GE) to T. P. Speis (NRC), ODYN Improvements, l October 13, 1981.

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