ML19284A870

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Supplemental Reload Licensing Submittal for Reload 4.
ML19284A870
Person / Time
Site: Peach Bottom Constellation icon.png
Issue date: 02/29/1980
From: Engel R, Henrikson P
GENERAL ELECTRIC CO.
To:
Shared Package
ML19284A871 List:
References
8ONED256, NEDO-24237, NUDOCS 8003100227
Download: ML19284A870 (45)
v • d • e


Text

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80 ED2 6 eenuS^**j l

SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM ATOMIC POWER STATION UNIT 2 RELOAD NO. 4 P. H. HENRIKSON l

c 80 0310 0 '2 E 7

NEDO-24237 80NED256 Class I February 1980 SUPPLEMENTAL RELOAD LICENSING SUBMITTAL FOR PEACH BOTTOM ATOMIC POWER STATION UNIT 2 RELOAD NO. 4 Prepared: /httt/pacu P. H. Henrikson Sr. Licensing Engineer Approved: . m/

E. Engel, anager Reload Fuel Licensing NUCLEAR POWER SYSTEMS OlVISION e GENER AL ELECTRIC COMPANY SAN JOSE. CALIFORNI A 95125 GENER AL h ELECTRIC

NEDO-24237 DTORTANT NOTICE REGARDING CONTENTS OF THIS REPORT PLEASE READ n:AREFULLY Thie report uas prepared by General Electric colcly for Philadclphia Electric Company (PECo) for PECo 's une uith the U.S. Nuclear Regulatory Commiccion (USNRC) for amending PECo 's operating license of the Peach Bottom Atomic Pouer Station Unit 2. The informtion contained in this report is believed by General Electric to be an accurate and truc represcntation of the facto known, obtained or provided to General Electric at the time this report was prepared, The only undertakings of the General Electric Company reapecting infomation in this document are contained in the contract between Philadelphia Electric Company and General Electric Company for nuclear fuct and related services for the nuclear system for Peach Bottom Atomic Poucr Station Unita 2 and 3, dated October 3, 1973, and nothing contained in this document shall be construed ao changing caid contract. The use of this infomation except ao defined by caid contract, or for any purpose other than that for chich it is intended, is not authorized; and with respcot to any such unauthorized use, neither General Electric Company nor any of the contributora to this document makes any representation or varranty (c press or implied) as to the completeness, accuracy or uccfulness of the information contained in this document or that such use of auch information may not infringe privately owned rights; nor do they ar.ncne any responsibility for liability or damage of any kind which may result from auch use of such information.

NED0-24237 PEACH BOTTOM ATOMIC POWER STATION UNIT 2 RELOAD 4

1. PLANT-UNIQUE ITEMS (1.0)*

Appendix A: New Bundle Loading Error Analyses Procedures Appendix B: Developmental Channels

.' Appendix C: Extended Exposure Lead Test Assembly Appendix D: GETAB Transient Analysis Initial Conditions Appendix E: Lead Test Assembly Reconstitution P8DRB285: See Reference 2 P8DRB284H: See Reference 3

2. RELOAD FUEL BUNDLES (1.0, 3.3.1, AND 4.0)

Fuel Type Number Number Drilled Irradiated Reload 1, 8DB274L 16 16 Reload 1, 8DB274H 20 20 Reload 1, LTA( } 4 4 Reload 2, 8DB274H 172 172 Reload 3, 8DRB284 260 260 New 56 56 P8DRB285(

P8DRB284H( 236 236 764 764 Total

References:

( )" Lead Test Assembly Supplemental Infornation for Reload 1 Licensing 2," NEDO-21172 Submittal for Peach Bottom Atomic Power Station Unit Rev 1, Supplement 1, March 1976.

( )R.E. Engel, letter to Tom A. Ippolito, "Ceneral Electric Licensing Topical Report NEDE-24011-P-A, " Generic Reload Fuel Application," Appendix D Sub-mittal", December 14, 1979.

J.F. Quirk, letter to Olan D. Parr, " General Electric Licensing Topical Report NEDE-24011-P-A, Generic Reload Fuel Application - Amendment 3",

May 8, 1979.

  • Refers to areas of discussion in " General Electric Boiling Water Reactor Generic Reload Fuel Application", NEDE-240ll-P-A-1, August 1979.

1

NEDO-24237

3. REFERENCE CORE LOADING PATTERN (3.3.1)

Nominal previous cycle exposure: 17,211 mwd /t. Assumed reload cycle exposure: 16,980 mwd /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 1.119 Fully Controlled 0.959 Strongest Rod Out 0.988 R, Maximum increase in Core Cold Reactivity with Exposure Into Cycle, (AK) 0.000

5. STANDBY LIQUID CONTROL SYSTEM AND SHUTDOWN CAPABILITY (3.3.2.1.3}

Shutdown Margin (AK) ppm (20*C, Xenon Free) 660 0.037

6. RELOAD-UNIQUE TRANSIENT ANALYSIS INPUTS (3.3.2.1.5 AND 5.2),

From BOC5 to From EOC5-1000 mwd /t EOC5-1000 mwd /t** to EOCS Void Coefficient N/A*

(c/% Rg) -8.86 /-11.08 -8.14 /-10.17 Void Fraction (%) 40.18 40.18 Doppler Coefficient N/A*

(c/*F T,y ) -0.227/ -0.216 -0.232/ -0.220 Average Fuel Temperature ('F) 1392 1392 Scram Worth N/A* ($) -34.68 /-27.74 -34.76 /-27.81 Scram Reactivity vs Time: Figure 2a Figure 2b

  • N = Nuclear input Data A = Used in Transient Analysis
    • Mid cycle exposure point 2

NED0-24237

7. RELOAD-UNIQUE GETAB TRANSIENT ANALYSIS INITIAL CONDITION PARAMETERS (5.2) 8x8 8x8R P8x8R EOC5-1000 EOC5-1000 E0C5-1000 mwd /t EOC5 mwd /t EOC5 mwd /t EOC5 Peaking Factors Local 1.22 1.22 1.22 1.20 1.22 1.20 Radial 1.41 1.38 1.53 1.50 1,52 1.48 Axial 1.40 1.40 1.40 1.40 1.40 1.40 R-Factor 1.098 1.098 1.058 1.058 1.058 1.058 Bundle Pcwer, (Mut) 5.810 5.810 6.317 6.317 6.389 6.389 3undle Flow (103 lbm/hr) 111.0 111.0 111.5 111.5 111.1 111.1 Initial MCPR 1.25 1.28 1.25 1.28 1.26 1.30
8. SELECTED MARGIN IMPROVEMENT OPTJS (5.5.2)

Exposure-Dependent Limits: From BOC5 to EOC5-1000 mwd /t and from EOC5-1000 mwd /t to EOC5

9. CORE-WIDE TRANSIENT ANALYSIS RESULTS (5.2.1)

Core ,.

Exposure Power Flow J Q/A 'SL V f.C P R Transient (Wd / t) (!) (*) ( *. ) (%) (psig) (psig) h3/8x6R/P9xeR Response Generator Load Rejec-tion. !!o Dypass E0C5-1000 104.5 100 266 115 1207 1233 0.19/0.18/0,19 Figure 3a Wd/ t Generator Load Rejec-tion. No Bypass E0C 5 104.5 100 290 117 1214 1240 0.21/0.21,0.23 Figure 3b Generator Load Rejec-tion. No Bypass E0C5 100.0 100 265 111 1195 1222 - - - -

Lo ss of 100*F Teedwater BOCS to Heating EOC5 104.5 100 124 124 1013 1069 0.15/0.15/0,15 Figure 4 Feedwater Controller Failure E0C5-1000 104.5 100 178 115 1146 1192 0.13<0.13/0.13 Figure 5a Wd/ r Feedwater Controller Failure E0C5 104.5 100 201 117 1148 1194 0.17/0.17/0,18 Figure 56 3

NEDO-24237

10. LOCAL ROD WITHDRAWAL ERROR (WITH LIMITING INSTRUMENT FAILURE) TRANSIENT SLWiARY (5.2.1)

MLHGR (kW/ft) Limiting Rod Rod Block Rod Position ACPR Pattern Reading (ft Withdrawn) 8x8 8x8R/P8x8R 8g 8x8R/P8x8R 104 4.5 0.13 0.11 11.8 14.6 Figure 6 5.0 0.14 0.13 11.8 14.9 105 106 5.5 0.15 0.15 11.8 15.2 107* 8.5 0.21 0.21 11.8 16.0 108 9.5 0.22 0.23 11.8 17.2 10.0 0.22 0.24 11.8 17.7 109 12.0 0.22 0.27 11.9 18.3 110

  • Indicates setpoint selected
11. OPERATING MCPR LIMIT (5.2)

EOC5-1000 tr4d/t BOC5 to to EOCS EOC5-1000 1rJd/t 1.28 8x8 1.28 8x8 1.28 8x8R* 1.28 8x8R*

1.30 P8x8R 1.30 P8x8R

  • 8x8R Limits are applicable to LTAs.
12. OVERPRESSURIZATION ANALYSIS SU50!ARY (5.3)

Core , ,,

Power Flow 'SL 'V Plaa:

Transient (?.) (~4) (psig) (psig) Responses; 104.5 100 1269 1299 Fi;;u r e 7 MSIV Closure (Flux Scram)

S

NED0-24237

13. STABILITY ANALYSIS RESULTS (5.4)

Decay Ratio: Figure 8 Reactor Core Stability:

~

Decay Ratio, X /X : 0.85 2 0 (105% Rod Line - Natural Circulation Power):

Channel Hydrodynamic Performance Decay Ratio, X2/XO (105% Rod Line -

Natural Circulation Power) 8x8 Channel 0.39 8x8R/P8x8R Channel 0.29

14. LOSS-OF-COOLANT ACCIDENT RESULTS (5.5.2)

See " Loss-of-Coolant Accident Analysis for Peach Bottom APS Unit 2", NED0-24081, December 1977, including Revision 5 of January 1980.

15. LOADING ERROR RESULTS (5.5.4)

See Appendix A of this supplement.

16. CORTROL ROD DROP ANALYSIS RES'ILTS (5.5.1)

Doppler Reactivity Coefficient: Figure 9 Accident Reactivity Shape Functions: Figures 10 and 11 Scram Reactivity Functions: Figures 12 and 13 Plant-Specific Analysis Results:

Parameter not bounded: Scram Reactivity Function (Cold)

Resultant peak enthalpy: 207 cal /gm (Cold) 5

NEDO-24237

_ _ _ _ _ . . _ . - __ _ ____ [o] o[ [o]

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Rt.LTA D* R 2, 8DB274H Figure 1. Reference Core Loading Pattern 6

NED0-24237 100 45 go C - 67A CRO IN PERCENT 1 - NOMIN AL SCRAM CURVE IN (-$) -

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NEDO-24237 45 100 90 - C - 67A CRD IN PERCENT - 40 1 - NOMIN AL SCRAM CURVE IN (-$1 2 - SCRAM CURVE USED IN ANALYSIS 80 - - 35 70 -

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NED0-24237 02 06 10 14 18 22 26 30 59 24 55 18 10 24 51 26 8 47 18 10 8 8 43 26 8 28 39 10 8 44 35 24 8 28 x 31 24 8 44 Notes: 1. Rod Pattern Is 1/4 Core Mirrer Symmetric, Upper Left Quadrant Shown on Map.

2. Numbers Indicate Number of Notches Withdrawn out of 48. Blank Is a Withdrawn Rod.
3. Error Rod Is Indicated by X Figure 6. Limiting RWE Rod Pattern 14

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NEDO-24237 1.2 ULTIMATE PERFORMANCE LIMIT 1.0 -

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NEDO-24237

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NED0-24237 0.020 080VNDING VALUE FOR 280 cal /g dCALCULATED VALUE 0.016 -

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NEDO-24237 0.020 O BOUNDING VALUE FOR 280 cal /g dCALCULATED VALUE 0.016 -

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NEDO-24237 4

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NED0-24237 1

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NEDO-24237 APPENDIX A NEW BUNDLE LOADING ERROR EVENT ANALYSES PROCEDURES The bundle loading error analyses results are based on new analyses procedures for both the rotated bundle and the mislocated bundle loading error events.

The use of these new analyses procedures is discussed below.

NEW ANALYSIS PROCEDURE FOR THE ROTATED BUNDLE LOADING ERROR EVENT The rotated bundle loading error event analysis results presented in +.his supplement are based on the new analysis procedure described a- -pproved in Reference A-1. This new method of performing the analysis is based on a more accurate detailed analytical model.

The principle dif ference between the previous analysis procedure and the new analysis procedure is the modeling of the water gap along the axial length of the bundle. The previous analysis used a uniform water gap, whereas the new analysis utilizes a variable water gap which is more representative of the actual condition, since the interfacing between the top guide and the fuel spacer buttons, caused by misorientation, causes the bundle to lean.

The ef fect of the variable water gap is to reduce the power peaking and the R-factor in the upper regions of the limiting fuel rod. This results in the calculation of a reduced CPR for the rotated bundle. The calculation was performed using the same analytical models as were previously used. The only change is in the simulation of the water gap, which more accurately represents the actual geometry.

The results of the analysis indicate for the P8x8R bundle a 17.3 kW/f t LHGR and 0.23 ACPR (includes a 0.02 penalty due to variable water gap R-factor uncertainty) with a minimum CPR of 1.07.

NEW ANALYSIS PROCEDURE FOR THE MISLOCATED BUNDLE LOADING ERROR EVENT The mislocated bundle loading error event analyses results presented in this supplement are based on the new analysis procedure described in Reference A-1.

A-1

NEDO-24237 This new method of performing the analysis employs a statistically corrected llaling procedure and analyzes every bundle in the core.

The use of the statistically corrected llaling analyses procedure indicates that the LilGR is 16.4 kW/fL and that the minimum CI'R for mislocated bundles is greater than the safety limit (1.07) tor all exposures throughout Cycle 5.

REFERENCES A-1 Safety Evaluation Report (letter), D. G. Eisenhut (NRC) to R. E. Engel (CE), MFN-200-78, dated May 8,1978.

4 A-2

NEDO-24237 APPENDIX B DEVELOPMENTAL CHANNELS The analyses given in Reference B-1 are applicable to continued use of develop-mental channels. The location and exposure of the developmental channels has changed. However, the thermal-hydraulic, nuclear, and safety analyses presented in the mai body of this submittal are applicable to the continued use of developmental fuel channels.

REFERENCES B-1 " Developmental Channels Supplemental Information for Reload 1 Licensing Submittal for Peach Bottom Atomic Power Station Unit 2", NED0-21172, Rev. 1, Supplement 2, March 1976.

B-1/B-2

NED0-24237 APPENDIX C EXTENDED EXPOSURE LEAD TEST ASSEMBLIES C.l. PROP 03ED PROGRAM The decision by the federal government to delay nuclear fuel reprocessing has encouraged studies by the nuclear fuel vendors of methods to reduce uranium utilization. An increase in discharge fuel exposures is one method which can be used to reduce uranium needs and improve fuel cycle costs. Current licer. sing analyses for General Electric reload fuel are performed for peak-pellet exposures up to 44,000 mwd /MT (40,000 mwd /ST).* The Peach Bottom-2 Lead Test Assembly (LTA) fuel program is one of several anticipated programs whereby lead burnup bundles will be extended to peak-pellet exposures greater than 44,000 mwd /MT (40,000 mwd /ST) . Information to be obtained from these pro-grams 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 LTAs to be operationally extended were first inserted into Peach Bottom-2 at the beginning of Cycle 2 (Reload 1). These assemblies (8DRB260) include the four highest expesure 8x8R fuel assemblies. It is expected that these assemblies will be inspected prior to insertion for Cycle 5 to ascertain f".el bundle integrity. The four bundles have been and will continue to be operated in symmetrical core positions.

Based on calculations, it is expected that peak pellet exposure may slightly exceed 44,000 mwd /MT (40,000 mwd /ST) during Cycle 5.

  • MT indicates metric ton, ST short ton C-1

NED0-24237 C.2. FUEL MECHANICAL DESIGN ANALYSIS Exposure-dependent fuel mechanical design analyses for the extended exposure LTA's have been performed for conditions which meet or exceed expected Cycle 5 operating conditions in Peach Bottom 2. Models, assumptions and material properties used in these analyses are those documented in Reference C l.

Calculated results are given below. These results are applicabic to all four LTA's during Cycle 5, with or without reconstitution.

C.2.1 FUEL ROD THERMAL ANALYSIS Safety evaluations are performed and measured against established safety criteria.

The consequence of calculating values which exceed such criteria is that fuel failure 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 transients are: (1) rupture of the fuel rod cladding due to strain ceused by relative expansion of the UO2 Pellet, and (2) severe overheating of the fuel rod cladding caused by inadequate cooling.

A value of 1% plastic strain of the Zircaloy cladding has been established as the safety limit below which fuel damage due to overstraining of the fuel cladding is not expected to occur. The Fuel Cladding Integrity Safety Limit ensures that fuel damage resulting from severe overheating of the fuel rod cladding caused by inadequate cooling, is avoided. Of these criteria, only the linear heat generation rate associated with the 1% plastic strain safety limit is af fected by increased fuel exposures. Analyses performed for the extendad exposure fuel bundles resulted in 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 (45,200 mwd /ST) f or urania-gadolinia rods. Both values include the 2.2% power spiking penalty documented in Reference C-1. These results assure that the same minimum margin to 1% plastic strain (175% of minimum steady-state power) reported in Reference C-1 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

NEDO-24237 C.2.1.1 Fuel Cladding Temperatures The cladding surface temperature is calculated using the cladding surface heat flux at a given axial position on a fuel element in conjunction with the over-all cladding-to-coolant film coefficient. The models used are noted in Reference C-1. The inside, average, and outside cladding temperature during normal operation at the end of Cycle 5 are calculated not to exceed 791 F, 753 F, and 716*F, respectively.

C.2.1.2 Fission Gas Release The amount of fission gas released during a time increment is calculated base 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 most limiting peaking factors is less than the 25% noted in Reference C-1.

C.2.1.3 Incipient Center Melting The fuel is designed so that fuel melting is not ex'pected to accur during normal steady-state fuel power operation which remains valid even at extended exposures.

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 structural integrity, resulting stress and/or strain levels are compared to the associated mechanical limits documented in Reference C-1. Results of the fuel rod mechanical analyses of the normal and transient loads for the extended exposure fuel are given below. The results of the combined LOCA and seismic evaluation documented in Reference C-1 do not change.

C.2.2.1 Cladding Creep Collapse Cladding creep collapse evaluation was performed with the models documented in Reference C-1. This calculation demonstrated that cladding creep collapse is not expected to occur in the event of a maximum overpressure transient throughout Cycle 5.

C-3

NEDO-24237 C.2.2.2 Stress Evaluations Fuel rod stress analyses of the extended exposure LTA's were performed with the model documented in Reference C-1 for operation through Cycle 5. These analyses showed that the fuel design ratios were well below 1.0.

C.2.2.3 Deflection Evaluation The operational fuel rod deflections considered are a result of manufacturing tolerances, flow-induced vibration, thermal ef fects, and axial load. Deflec-tions of the extended exposure LTAs were combined and compared to the fuel rod-to-fuel rod and fuel rod-to-channel spacing deflection limits given in Reference C-1. This comparison demonstrated that the fuel rod clearance criterion was met.

C.2.2.4 Fatigue Evaluation The cyclic loads considered in cladding f atigue analysis are coolant pressure and therral gradients. The analysis performed for the higher exposure LTAs was based on previous and projected operating cycles through the end of Cycle 5, 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-1 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 opetational transients because of the thermal margins at which the f uel is operated and the high fuel burnups. It can, there-fore, be concluded that the energy release and potential for a chemical reaction C-4

NED0-24237 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-1 does not change because of the extended fuel exposures.

C.2.3.3 Fretting Wear and Corrosion As discussed in Reference C-1, 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 results. It is expected that the LTAs will be visually examined before loading in Cycle 5.

9 C-5

NEDO-24237 C.3. IMPACT ON RELOAD ANALYSES All of the models documented in Reference C-1 are applicable for use with higher fuel exposures. However, some inputs into these models are exposure-dependent 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 reactivities, relative rod powers, and few group cross-sections as functions of instantaneous void, exposure, exposure-void history, control state, and f uel and moderator temperature. Because of the exposure dependence of these results, the libraries were expanded to include higher burnups as noted below. The core analysis is unique for each reload. It is performed using the above lattice " libraries" 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 Traditionally, bundle reactivities have been expressed in terms of k, (i.e. ,

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 extended exposure LTAs decreases by 0.05 Ak, from a lattice exposure of 38,000 to 50,000 mwd /EE (35,000 to 45,000 mwd /ST).

C.3.1.2 Local Peaking Factors For 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 C-6

NED0-24237 each fuel rod. Calculated maximum local peaking factor for the extended exposure 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 f actor does vary with void fraction, and this dependence is taken into account in the calculations used to assign local peaking f actors 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 in the absorption of resonance-energy neutrons caused by a change in the tem-perature of the material in question. The Doppler reactivity coefficient 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-1.

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 from the reduction in neutron slowing down due to the decrease in the water-to-fuel ratio. Beyond 11,000 mwd /MT (10,000 mwd /ST), the void effect is essentially constant.

C.3.2 STEADY-STATE HYDRAULIC ANALYSIS Core steady-state therma l-hydraulic analyses are performed using a model 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-1. The flow distribu-tion to the fuel assemblies and bypass flow paths is calculated on the assumption C-7

NEDO-24237 that the pressure drop across all f uel assemblies and bypass flow paths is the same. An iteration is performed on flow through each flow path (fuel assemblies 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 radioactive release from plants for normal operation, abnormal operational transients, and postulated accidents meet applicable regulations in which conservative limits are docu-mented. This conservatism is augmented by using conservative evaluation models and observing limits which are more restrictive than those documented in the applicable regulations. These observed operating limits and methods used to determine if the limits are met are documented in Reference C-1.

C.3.3.1 Fuel Cladding Integrity Safety Limit The generation of the Minimum Critical Power Ratio (MCPR) limit requires a stapistical 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 f or 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 AMCPR value for the most limiting transient from rated conditions postulated to occur at the plant to the fuel cladding integrity safety limit. Higher fuel C-8

NEDO-24237 exposures are reflected in the nuclear input data. However, due to the high exposure, these fuel assemblies will operate at significantly lower power levels 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 steamline isolation valves (MSIV) with indirect (flux) scram, and the margin at extended exposures will not exceed the nominal E0C margin because of the reduced power levels. The model used to analyze this event is described in Reference C-1.

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 reactiv-ity 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 performance 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 may be encountered in the course of BWR plant operation.

The analysis is performed using the models documented in Reference C-1 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-9

NEDO-24237 C.3.3.5 Accident Evaluations Accidents are events which have a projected frequency of occurrence of less than once in every one hundred years for every operating BWR. The broad spectrum of postulated accidents is covered by six categories of design basis events. These events are the control rod drop, main steamline 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 MAPLHGR values and associated peak cladding temperatures and oxidation fractions were incorporated into Reference C-2.

C-10

NED0-24237 C.4. REFERENCES C-1. "Ceneric Reload Fuel Application," NEDE-240ll-P-A-1, August 1979.

C-2. " Loss of Coolant Accident Analysis for Peach Bottom Atomic Power Station Unit 2," NED0-24081, December 1977, including Revision 5 of January 1980.

C-ll/C-12

NEDO-24237 APPENDIX D GETAB TRANSIENT ANALYSIS INITIAL CONDITIONS Table 5-8 of NEDE-240ll-P-A-1 contains the non-varying plant GETAB analysis initial conditions. However, for Peach Bottom 2 Reload 4 the assumed reactor core pressure and inlet enthalpy are:

Reactor Pressure (psig) 1035.0 Inlet Enthalpy (Btu /lb) 521.4 D-1/D-2

NEDO-24237 APPENDIX E LEAD TEST ASSEMBLY RECONSTITUTION E.1 INTRODUCTION The Lead Test Assemblies (LTA) were described .in " Lead Test Assembly Supple-mental Information for Eeload 1 Licensing Submittal for Peach Bottom Atomic Power Station Unit 2," NED0-21172, Revision 1, Supplement 1, March 1976.

During the Reload No. 3 refueling outage, two fuel rods were removed from assembly LJLTA-2 and replaced by fresh rods with a U-235 enrichment of 3.00 weight percent. The removed rods were destructively examined and were not employed during future Peach Bottom Unit 2 operation. The information con-tained in NEDO-21172, Revision 1, Supplement 1, was unaffected by the recon-stitution of the LTA since the replacement rods were virtually identical to those used during initial fabrication of the LTA. The results of analyses and evaluations of the effect of installing two fresh rods in place of two exposed rods on the LTA itself and on the safety analyses for operation of Peach Bottom Unit 2 following Reload No. 3 were reported in Appendix B of NEDO-24132, Revision 1, September 1978.

As a continuation of this program, during the Reload No. 4 refueling outage, it is expected that up to five fuel rods will be removed from assembly LJLTA-3 and replaced with fresh rods with a U-235 enrichment of 2.00 weight percent.

It is expected that the removed rods will be destructively examined and such rods will not be used during future Peach Bottom Unit 2 operation. The enrich-ment of the replacement rods is less than the initial enrichment of the rods they are replacing to compensate for fuel depletion and selected so that the computed reactivity of the reconstituted LJLTA-3 will never exceed that of a regular LTA bundle. Therefore, the information contained in NEDO-21172, Re-vision 1, Supplement 1 is unaffected by the reconstitution of the LTA since the nuclear characteristics of the reconstituted bundle are virtually iden-tical to a non-reconstituted bundle. The purpose of this appendix is to report the results of the analyses and evaluations of installing five fresh rods in place of five exposed fuel rods in the LTA and on the safety analyses for operation of Peach Bottom Unit 2 following Reload No. 4.

E-1

NEDO-24237 The locations of the planned exchanged rods in assembly LJLTA-3 are shown on Figure E-1. After reconstitution, LJLTA-3 will be loaded into core location 13-14 as shown in Figure 1. The process computer nodal exposure values for LJLTA-3 will be adjusted to account for the removal of the five exposed rods and the insertion of five fresh rods. The nuclear data for the unperturbed LTAs will be used in the process computer evaluations of the nodal powers of LJLTA-3. However, the R factors and local peaking factors for assembly LJLTA-3 will be used in the CPR and linear heat generation rate determinations for the four LTAs. The process computer will use data from the LPRM string situated at location 48-49 (see Figure 1) in the detcrmination of the nodal powers of LJLTA-3. TIP data from location 48-49 are applicable to the LJLTA-3 location.

Further, the determination of the average power of the nodes of the four fuel assemblies surrounding LPRM location 12-13 (pseudo location for LJLTA-3) is not significantly affected by the presence of LJLTA-3 instead of an unperturbed LTA, because of the insignificant change in k, of LJLTA-3 due to reconstitu-tion. The k,of LJLTA-3 is within 0.4% of the k, of the unperturbed LTA.

Therefore, the accuracy of the process computer power allocation to LJLTA-3 is affected by 50.4%.

E.2 EVALUATIONS AND ANALYSES E.2.1 Nuclear and Thermal Parameter Evaluations Standard lattice physics calculations were made for the reconstituted LTA including simulation of the fresh rods. Cycle 5 operation was simulated by

" burning" the reconstituted LTA in ten exposure steps.

Over the exposure range of interest, the computed lattice reactivity of the reconstituted LTA is within 0.4% AK of the unperturbed LTA reactivity. The fuel rod power peaking for the reconstituted LTA remains low, but is up to 2% greater than the value for an unperturbed LTA at the same exposure.

Although the reduced gap conductance in the two fresh rods tends to reduce the change in critical power ratio (ACPR) due to transients, which tends to reduce the MCPR operating limit, the effect of the small increase in fuel rod power peaking results in an overall slight decrease in the steady-state CPR of the reconstituted LTA. To account for the slight decrease in steady-state CPR, and the small increase in local peaking, the R factors and local E-2

NEDO-24237 peaking factors for the LTAs will be increased to account for the 2% greater fuel rod power peaking factor of LJLTA-3. The process computer will compute the steady-state actual CPR and LHGR associated with the reconstituted LTA for all four LTAs, while the MCPR operating limit already calculated for the LTAs will be conservative due to the slight decrease in ACPR for LJLTA-3.

Using the BWR Simulator, power distribution analyses have been performed for Cycle 5 of Peach Bottom 2. The core exposure distribution at beginning of Cycle 5 was obtained by simulating the estimated exposure accumulation to end of Cycle 4 and representing the projected refueling and reconstitution.

The results show that the fuel assemblies surrounding the reconstituted LTA will reach thermal limits significantly sooner than the reconstituted LTA.

Estimated margins between the reconstituted LTA and the surrounding fuel assemblies are:

MCPR 3% to 12%

MAPLHGR 11% to 18%

Linear Heat Generation Rate 7% to 14%

Due to the relatively high exposure, the neighboring fuel assemblies will operate closer to limits, and LJLTA-3 is predicted never to be limiting.

MCPR limits during transients are not affected because LJLTA-3 is predicted never to be limiting.

E.2.2 Mechanical Design Evaluation Tt i five replacement fuel rods in LJLTA-3 are mechanically similar to the fuel rods which they are replacing and also to the standard fuel rods in the Reload 4 fuel bundles. The only mechanical difference is a longer upper end plug on each replacement rod to accommodate the irradiation growth of the rods in LJLTA-3. An analysis of differential rod growth in the modified LJLTA-3 was performed. The results of the analysis show that the replace-ment fuel rods are mechanically compatible with the irradiated rods and thus will have no adverse effect on the existing safety analyses for Cycle 5 or subsequent cycles for Peach Bottom 2.

E-3

NEDO-24237 The peak linear heat generation rate of the reconstituted LTA, although slightly higher than the other LTAs at the same exposure, is still within the operating limit of 13.4 kv/f t which was usad in evaluating the mechanical performance of the maximum duty fuel rod in Reload 3. Therefore, the results of the fuel rod thermal and mechanical design evaluations in the " Lead Test Assembly Supplemental Information Reload 1 Licensing Submittal for Peach Bottom Atomic Power Station Unit 2," NEDO-21172, Revision 1, Supplement 1, March 1976, are conservatively applicable to the reconstituted LJLTA-3.

E.2.3 Evaluation of the Effect of the Fresh Fuel Rods on PCT /MAPLHGR Reconstitution of LJLTA-3 will result in a reduction of the planar average exposure of the assembly compared to that assumed in the analysis of the original LTA (NEDO-21172, Revision 1, Supplement 1, March 1976).

The following effects of the change in the effective exposure on LOCA analysis provide the basis for assessing the effect on Peak Clad Temperature:

1) The local power distribution in the center 16 rods is decreased.

Thia occurs because the fresh rods redistribute the power to the-periphery of the bundle.

2) The calculated total stored energy is increased. This occurs because the calculated gap conductance at low exposures is gen-erally smaller, and since the reconstituted bundle has a lower effective exposure compared to the unperturbed LTA, the calculated stored energy would be higher compared to that calculated in the LOCA analysis for the unperturbed LTA. The maximum increase in total planar stored energy is approximately 1.% for all planar exposures.

The resulting decrease in PCT due to the above effects is less than 20"F.

Since the PCT of the unperturted LTA is well below 2200*F, the PCT of the reconstituted LTA will also remain below 2200*F. Thus, the previously cal-culated MAPLHGRs for the LTA given in NEDO-24081, " Loss-of-Coolant Accident E-4

NEDO-24237 Analysis for Peach Bottom Atomic Power Station Unit 2," December 1977, are valid for the reconstituted LTA.

E.2.4 Transient Analysis for Cycle 5 Based on the analysis results described in Section E.2.1 above, the transient analysis results contained in this submittal are unaffected by reconstitution of the LTA.

E.3 Summary and Conclusions It is concluded, based on the results of the evaluations and analysis described in Section E.2 that the accident and transient analyses of Cycle 5 are insigni-ficantly affected and the operating limits of Cycle 5 are unaffected by the introduction of the reconstituted LTA. The operat'ng MCPR limit is given in Section 11 of this submittal.

E-5

NEDO-24237 WIDE-WIDE CORNER T T 4 3 2 2 2 2 2 3 G

3 2 1 5 1 1 1 2 T

~

G T 2 1 1 1 1 1 5 1 G

2 5 1 1 WS 1 1 1 2 1 1 W 1 1 1 1 T T 1 1 1 2 1 1 1 1 G G 2 1 5 1 1 1 5 1 T T 3 2 1 1 1 1 1 2 ROD ENRICHMENT NUMBER TYPE wt % U-235 OF RODS 1 3.01 38 WS - SPACE POSITIONING / 2 2.22 14 WATER ROD T - TIE RODS 3 127 4 G - GADOLINIUM RODS W - WATER ROD 4 1.45 1 RECONSTITUTED RODS 5 3m 5 LJLTA 3 WS/W _. 2 Figure E-1. Location of Fuel Rod Types Within 8xfEBundle E-6

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