ML20154H249

Large Break Loca/Eccs Analysis W/Increased Radial Peaking

ML20154H249
Person / Time
Site:	Palisades
Issue date:	08/02/1988
From:	Gottula R SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML18053A574	List: ML20154H249 ML20154H260
References
ANF-88-107, NUDOCS 8809210240
Download: ML20154H249 (52)

Text

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. ADVANCED NUCLEAR FUELS CORPORATION PALISADES LARGE BREAK LOCA/ECCS ANALYSIS WITH INCREASED RADIAL PEAKING 1

AUGUST 1988 hhk kD C 55 p PDC

ADVANCED NUCLEARFUELS CORPORATION ANF-88-107 Issue Date:8/2/88 PALISADES LARGE BREAK LOCA/ECCS ANALYSIS WITH INCREASED RADIAL PEAKING

k. O- b R. C. Gottula Team Leader PWR Safety Analysis Licensing and Safety Engineering Fuel Engineering and Technical Service Contributors:

N. F. Fausz B. E. Schmitt (Intermoutain Technologies Inc.)

Calvin Slater Ross Jensen l

August 1988

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CUSTCMER OtSCLAlWER OfPOftTANT NOTICS ROGAMlHNO CONTENTS ANO USE OF TNts _

DOCVesENT _

PLEAes READ CAREP%LY A&ensed Numeer Fuese Corporecon's worrences and reoroesntatoes con-commg tre suetect metter of this occument are those est be e M Agreernent Donocen Adveneed Numeet Puses Corporeuen and the Cwesomer purmet to afucfi thee occument 4 itewd. GWgi, except as otmermee esprocery prt>

vwled M euer Agreement, norther A&enced Nucear Fwe6e Corporsoon not any person t M on te bened meses any warrantv v representanon, encroceeg cr espeed, weet roepect to the escuracy, cornplete '.ac or useN. noes of tP6 ofore tw contenned a tN4 document. of that the ue e of any enformenon. 400eratus, memed or prese f1 M W1 %* 4 document mu rM anffino6 prTvetery uwneJ r*.)nte: or assumes arty heoshties witf1 respect to the use Cf arty M.stion, ae pareous, momed or process tsomoeed a true coeunent.

The arttermeson genessmed herem e for the sees use of Cuenorner in oreer to evend impearment W ngnie of Advanced Numeer Pvens Corporenon vi pesone er wyngrisong wheen eney De meuced in the untermeson oorftesned m thee Gesument, the roepsont. By se 42eatence of thee occument, adrese not to putrem or rnese pubese use On the pelect use of the term) 08 Ouca miermanon urmi so authorised in writing ty Advanced Nucieer Puese Carperacon or urmi after su (g) r%'Mme toetoweg termma:en or expretion of tPe aloroeesc Agreemem and any ementon thereof, uruene otnerwee omoserf ortmded e ?e Agreereont. No nonte or nooness a , a ey perents are wreem ey tne Nm eneg of we ooev.

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1- ANF 88 107 Ithl3 of content,1  ;

1.0 INTRODUCTION

............................ 1 2.0

SUMMARY

0F RESULTS ......................... 3 l

3.0 ANALYS!$ .............................. 6 3.1 Description of LBLOCA Transient . . . . . . . . . . . . . . . . . . . 6 3.2 Description of Analytical Models .................. 4 3.3 Plant Description and Summary of Analysis Parameters ........ 8 j 3.4 Break Spectrum Results ....................... 9 l

3.5 Axial shape Stud) 'ssults . . . . . . . . . . . . . . . . . . . . . . 10  !

3.6 Exposure Limits . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 l

4.0 CONCLUSION

S . . . . . . . . . . . . . . . . .'. . . . . . . . . . . 43  ;

5.0 REFERENCES

............................. 44 -

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l'ist of Tables  ;

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2.1 Summary of Results for 0.6 DECLG Limiting Break Size ....... 4 -

3.3.1 Palisades System Analysis Parameters . . . . . . . . . . . . . . 12 ,

i 3.4.1 Palisades Break Spectrum Analysis Results . . . . . . . . . . . . 14 3.4.2 Calculated Event Times for 0.4 DECLG Break . . . . . . . . . . . 15 l i

I 3.4.3 Calculated Event Times for 0.6 DECLG Break . . . . . . . . . . . 16 I

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" List of Fiaures

E1SE11 f4St 2.1 Allowable LHR as a Function of Peak Power Location . . . . . . . . 5 t

1 3.5.1 Normalized Power (E0C), 0.6 DECLG Break ............. 17 3.5.2 Double Intact Loop Accumulator Flow Rate, 0.6 DECLG Break .... 18 ,

3.5.3 Singic Intact loop Accumulator Flow Rate. 0.6 DECLG Break .... 19 3.5.4 Broken Loop Accumulator Flow Rate, 0.4 DECLG Break . . . . . . . . 20 j t

3.5.5 Total Intact Loop HPS! Flow Rate. 0.6 DECLG Break ........ 21 3.5.6 Total Intact Loop LPS! Flow Rate. 0.6 DECLG Break ........ 22 3.5.7 Broken Loop $!$ Flow Rate. 0.6 DECLG Break . . . . . . . . . . . . 23 3.5.8 Upper Plenum Pressure during Blowdown, 0.6 DECLG Break . . . . . . 24 3.5.9 Total Break Flow Rate during Blowdown. 0.6 Break . . . . . . . . . 25 3.5.10 Pressurizer Surge Line Flow Rate during Blowdown. 0.6 DECLG Break ..................,........... 26 j I

3.5.11 Downcomer Flow Rate during Blowdown, 0.6 DECLG Break . . . . . . . 27

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3.5.12 Average Core Inlet Flow Rate during Blowdown 0.6 DECLG Break, 28 I X/L = 0.8 ............................

I 3.5.13 Hot Channel Inlet Flow Rate during Blowdown. 0.6 DECLG Break, 29 X/L = 0.8 ............................ f 3.5.14 Hot Volume Inlet Flow Rate during Blowdown. 0.6 DECLG Break, l 30 X/L = 0.8 ............................ l

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-iv- ANF 88 107 M af Fiaures Fiaures 21g3 3.5.15 Hot Node Fluid Quality during 81owdown. 0.6 DECLG Break, 31 X/L = 0.8 ............................

3.5.16 PCT Node Fluid Temperature during Blowdown. 0.6 DECLG Break, 32 I X/L = 0.8 ............................

3.5.17 PCT Node Fuel Average Temperature during Blowdown 0.6 DECLG Break, X/L = 0.8 . . . . . . . . . . . . . . . . . . . . . . . . . 33 3.5.18 PCT Node Cladding Temperature during Blowdown. 0.6 DECLG Break, 34 X/L = 0.8 ............................

l 3.5.19 PCT Node Heat Transfer Coefficient during Slowdown, 0.6 DECLG Break, X/L = 0.8 . . . . . . . . . . . . . . . . . . . . . . . . . 35 3.5.20 PCT Node Heat Flux during Blowdown 0.6 DECLG Break, X/L = 0.8 . . 36 3.5.21 Containment Pressure, 0.6 DECLG Break, X/L 0.8 . . . . . . . . . 37 3.5.22 Upper Plenum Pressure after E0BY, 0.6 DECLG Break, X/L = 0.8 . . . 38

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i 3.5.23 Downcomer Mixture Level after E0BY, 0.6 DECLG Break, X/L = 0.8 . . 39 3.5.24 Core Flooding Rate after E0BY, 0.6 DECLG Break, X/L = 0.8 .... 40 3.5.25 Core Mixture Level after E08Y, 0.6 DECLG Break, X/L = 0.8 .... 41 3.5.26 PCT Node Cladding Temperature af ter E0BY, 0.6 DECLG Break, 42 X/L = 0.8 ............................

1

ANF 88 107 PALISADES LARGE BREAX LOCA/ECCS ANALYSIS WITH INCREASED RADIAL PEAXING

1.0 INTRODUCTION

This document presents the results of a large break loss of coolant accident (LOCA) analysis for the Palisades plant operating with Advanced Nuclear Fuels Corporation (ANF) fuel. The primary purpose of the analysis was to support an increase in the total radial peaning factor from 1.77 to 1.83.

The analysis was performed at a total radial peaking factor of 1.92 to bound.

l potential future increases in the total radial peaking factor. The analysis supports a maximum LHR of 15.28 kW/ft and a modification in the axial LHR limit curve shown in Figure 3.23-1 of the technical specifications. The analysis also provides justification fur removal of Figure 3.23 2 (allowable LHR as a function of burnup) and Figu ,3.23 3 (allowable LHR as a function of peak power location for interior and narrow water gap fuel rods) from the technical specifications. The analysis was performed for the Palisades plant operating at 2581 MWt (2530 MWt plus 2% uncertainty) and a maximum average steam generator tube plugging level of 29.3% with up to 4.5% asyneetry.

Numerous changes have occurred in the ANF LOCA methodology since the previous licensing calculations were performed for the Palisades plant.

Therefore, the scope of this analysis includes a mini break spectrum analysis.

Calculations were perfomed for a 0.4, 0.6, and 0.8 double ended cold leg guillotine break (DECLG) at the pump discharge to verify the previously determined 0.6 DECLG limiting break size (II. The analysis also includes calculations at the limiting break size for both a BOC axial power shape peaked at a relative core height of 0.6 and an EOC axial power shape peaked at a relative core height of 0.8. The calculations conservatively used the

2- ANF 88 107 maximum fuel stored energy near 80C where maximum densification occurs.

Justification is provided to support operation with ANF fuel up to a bundle average exposure of 52,500 Wd/MTU with regard to the large break LOCA.

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ANF-88 107 2.0 SIM ERY OF RESULTS The results of the analysis verified the 0.6 DECLG break as the limiting break size. The analysis demonstrates that the 10 CFR 50.45(b) l criteria are satisfied for the Palisades plant with the axially dependent power peaking limit curve shown in Figure 2.1. The analysis supports a maximum LHR of 15.28 k'4/ft up to a relative core height of 0.6 and a LHR of 14.75 kW/ft at a relative core height of 0.8. The analysis supports a total radial peaking factor of 1.92 and a maximum average steam generator tube plugging level of 29.3% with up to 4.5% asymetry. Results of the analysis l for both the BOC and EOC axial profiles at the limiting 0.6 DECLG break size are shown in Table 2.1. The peak cladding temperature was calculated to be 1914'F for the BOC profile and 2114'F for the E0C profile. The analysis

supports Cycle 8 operation and is intended to support operation for future cycles.

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4- ANF 88 107 TABLE 2.1

SUMMARY

OF REsblTS FOR 0.6 DECLG LIMITING BREAK $!ZE BOC Stored Energy BOC Stored Energy 80C Axial shape E0C Axial Shape 1/L = 0.6 1/L = 0.3 15.28 14.75 Peak LHR (kW/ft)

Hot Rod Burst 41.77 41.37 Time (Sec) 8.9 Elevation (ft) 7.4

- Channel Blockage Fraction 0.31 0.34 Peak Cladding Temperature

- Temperature ('F) 1913.7 2114.2

- 52.47 57.57 8.9 Time (Sec)

Elevation ( ft) 7.4 Metal Water Reaction

- 2.23 4.14

- Local Maximum Elevation of Loca (%)l Max. (ft) 7.4 8.9

- Hot Pin Total (%) 0.46, 0.47,,

<l.0 <1.0 Core Maximum (%)

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6- ANF-88 107 0.0 ANALYSIS Section 3.1 of this report provides a description of the postulatcd large break loss of coolant transient. Section 3.2 desa.ribes the methodology and major assumptions used in the analysis. Section 3.3 provides a description of the Palisades plant and a summary of the system, parameters used in the analysis. Section 3.4 provides a sumary of the results of the mini break spectrum calculations. Section 3.5 summarizes the results of the limiting EOC l

axial power shape and section 3.6 provides justification for the burnup l

independence of the LHR limit for ANF fuel.

3.1 Descrintion of LRLOCA Transient A loss of coolant accident (LOCA) is defined as the rupture of the ,

Reactor Coolant System primary piping up to and including a double ended

guillotine break. The limiting break occurs on the pump discharge side of a l cold leg pipe. The LOCA is assumed to result from an earthquake and is co-i incident with loss of offsite power. Primary coolant pump coastdown occurs co incident with the loss of offsite power. Following the break, depressurization of the reactor coolant system, including the pressurizer, occurs. A reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached. Reactor trip and scram are conservatively neglected in f l

the LOCA analysis. Early in the blowdown, the reactor core experiences flew  ;

reversal and stagnation which causes the fuel rods to pass through critier.1 i l

heat flux (CHF). Following CHF. the fuel rods dissipate heat through the transition and film boiling modes of hest transfer. Rewet is precluded during f blowdown by Appendix K of 10 CFR 50.

A Safety injection System (SIS) signal is actuated when the appropriate setpoint (high containment pressure) is reached. Due to loss of offsite [

power, a time delay for startup of diesel generators and SIS pumps is assumed.

Once the time delay criteria is met and the system pressure falls below the  :

shutoff head of the High Pressure Injection System (HPSI) and low Pressure  !

Injection System (LPSI) pumps, SIS flo.. is injected into the cold legs. f l

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7- ANF 88 107 Single failure criteria is met by assuming that one HPSI pump and one LPSI pump are not available for operation. When the system pressure falls below the accumulator pressure, flow from the accumulators is injected into the cold legs. Flow from the Emergency Core Cooling System (ECCS)is assumed to bypass the ore and flew to the break until the end of bypass (E0BY) is predicted to Following E0BY, ECCS flow fills occur (sustained downflow in the downcomer).

the downcomer and 1cwor plenum until the liquid level reaches the bottom of the core (beginning of core recovery or BOCREC time). During the refill period, heat is transferred from the fuel rods by radiation heat transfer.

The reflood period begins at BOCREC time. ECCS fluid fills the downcomer and provides the driving head to move coolant through the core. As the mixture level moves up the core, steam is generated. Steam binding occurs at the steam flows through the intact and broken loop steam genarators and pumps.

The pumps are assumet to have a locked rotor (per Appendix K of 10 CFR 50) which tends to reduce the reflood rate. The fuel rods are eventually cooled and quenched by radiation and convective heat transfer as the quench front moves up the core. The reflood heat transfer -ate is predicted through experimentally determined heat transfer and carry over rate fraction correlations.

The purpose of the LBLOCA analysis is to demonstrate that the criteria stated in 10 CFR 50.46(b) are met. The criteria are:

1) The calculated peak fuel element cladding temperature does not exceed the 2200 'F limit.

2) The amount of fuel element cladding which reacts chemically with water or steam does not exceed 1% of the total amount of zi;caloy in the core.

3) The cladding temperature transient is terstnated at a time when the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limit of 17% is not exceeded during or after quenching.

4) The core temperature is reduced and decay heat is removed for an extended period of time, as required by the long lived radioactivity remaining in the core.

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8 ANF 88 107 3.2 Deterintion of Analytical Models The ANF EXEM/PWR evaluation model(2) was used to perform the analysis. This

. evaluation model consists of the following computer codes:

1) R00EX2(3) for computation of initial fuel stored energy, fission gas release, and gap conductance

2) RELAP4 EM for the system and hot channel blowdown calculationst

3) CONTEMPT /LT 22 as modified in accordance with NRC Branch Technical Position CSB 6-1 for computation of containment back pressure

4) REFLEX for computation of system refloods and

5) T000EE2 for the calculation of fuel rod heatup during the refill and reflood portions of the LOCA transient.

The quench time, quench velocity, and carryover rate fraction (CRF) correlations in REFLEX, and the heat transfer correlations in T000EE2 are based on ANF's Fuel Cooling Test Facility (FCTF) data.

The governing conservation equations for mass, energy, and momentum transfer are used along with appropriate correlations consistent with Appendix l X of 10 CFR 50. The reactor core in RELAP4 is modeled with heat generation .

l rates determined from reactor kinetics equations with reactivity feedback, and with actinide and decay heating as required by Appendix K. Appropriate conservatisms specified by Appendix K of 10 CFR 50 are incorporated in all l

the DEM/PWR models.

3.3 Plant Descrietion and Sumary of Analytit Parameters The Palisaties plant is a Combustion Engineering (CE) designed pressurized water reactor which has two hot leg pipes, two U tube steam generators, and four cold leg pipes with one recirculation pump in each cold leg. The plant uttitzes a large dry containment. The reactor coolant system was nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow paths or

• junctions". The two cold legs connected to

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the intact loop steam generator were t + .o be symetrical and were modeled as one intact cold leg with appropriately scaled input. The model consiJers four accumulators, a pressurizer, and two steam generators with both 1 primary and secondary sides of the steam generators modeled. The high pressure safety injection (HPSI) and residual heat removal (LPSI) pumps were modeled as fill junctions at the accumulator lines, with conservative flow 1 rates given as a function of system back pressure. The pump performance curves are characteristic of CE pumps. The reactor core was modeled radially with an average core and a hot assemoly as pt.'allel flow chaanels, each with three axial nodes. A steam generator tube plugging level of 2g.3% was assumed with an asymetric steam generator tube plugging of 4.55. The break was conservatively assumed to have occurred in the most highly plugged loop since this results in more steam binding during reflood and a higher peak cladding temperature, ti 5

Values for system parameters used in the analysis are given in Table l 3.3.1.

3.4 treak heetrum Results A mint break spectrum study was performed to confirm the previously l

l determined 0.6 DECLG break as the limiting break size since numerous changes have occurred in the ANF LOCA methodology since the previous licensing calculations were performed for the Palisades Plant. Calculations wre performed for 0.4, 0.6, and 0.8 DECLG break sizes with an axial power shape l

peaked at a relative core heignt of 0.6. Also, ANF methodology previously and currently shows that split breaks are less ilmiting the guillotine breaks.

Therefore, split break calculations were not included in this analysis, system blowdown calculations were first performed to the end of bypass (E04Y) f to confirm the 0.6 DECLG as the limiting break size. Fuel and cladding l

temperatures betwsen the 0.4 and 0.6 OECLG break sizes were fairly : lose at the end of bypass such that it was not conclusive that the 0.6 DECLG break was the limiting break. Therefore, calculations were perforino through the refill l

and reflood periods for these two break sizes. The results of the break

O 10- ANF 88 107 spectrum study are shown in Table 3.4.1. The 0.6 DECLG bregk size is confirmed as the limiting break. The peak cladding temperature (PCT) for the 0.6 DECLG break with an axial power shape peaked at a relative core height of 0.6 was calculated to be 1914 'F. Thus, a maximum LHR of 15.28 kW/f t is supported up to a relative core height of 0.4. Calculated event times for the 0.4 OECLG break are shown in Table 3.4.2. Calculated event times for the 0.6 DECLG break are shown in Table 3.4.3.

3.5 Axial Shane Study Results An EOC (top skewed) axial power shape was analyzed to define the axially dependent LHR limit curve shown in Figure 2.1. The axial power shape was peaked at a relative core height of 0.8 with an LHR of 14.75 kW/ft. The i axial shape was selected from those nial shapes allowed by Tjn),g LC0 barn.

A BOC fuel stored energy was conservatively used in conjunction with this axial shape. The results for the EOC shape are shown in Table 2.1. The PCT was calculated to be 2114'F. Plots of parameters depic. ting calculations for the limiting 0.6 DECLG break and tA EOC shape are shown in Figures 3.5.1 through 3.5.26.

3.6 EElg.g re Limits The results of previous exposure anilyses for the Palisades plant III required a reduction in the LHR limit at high exposures. This tas a result of the use of the previous ANF fuel rod code GAPEX. Exposure calculations have been performed with the current EXEM/PWR methodology using R00EX2 for two plants with a maximum bundle average exposure of 52,500 mwd /MTU. The current ANF methodology predicts maximum fuel storage energy to occur near BOC where maximum densification occurs. Closure of the fuel cladding gap at higher expostres significantly reduces the fuel stored energy. At high exposures, gap closure significantly outweighs the effect of higher concentrations of fission gases which tend to reduce the gap conductance and increase fuel stored energy. Also, the reduced stored energy at high expcsures outweighs any adverse effects of increased rod internal pressure at high exposures.

Thus, the peak cladding tereperature will be lower at high exposures than for

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l the limiting case reported in section 3.5 which assumes a 30C fuel stored l energy. 31nce this phenomena is fuel related rather than system related, the }

exposure study results for other plants are spplicable to the Palisades plant, i Thus, the LHR limit is independent of exposure up to a maximum bundle average  ;

t exposure of 52,500 M /MTU.

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o ANF-88-107 TABLE 3.3.1 PALISADES SYSTEM ANALYSIS PARAMETERS Primary Heat Output, MWt 2530*

Primary Coolant Flow Rate, lbm/hr 1.203 x 108 (318,770 gpm)

Primary Coolant System Volume, ft 3 8808**

Operating Pressure, psia 2060 Inlet Coolant Temperature (hottestloop),'F 544 Reactor Vessel Volume, ft3 4782 Pressurizer Total Volume, ft3 1504 Pressurizer Liquid Total, ft 3 803 Accumulator Total Volume, ft3 (one of four) 2011 Accumulator Liquid Volume, ft 3 1116 Accumulator Pressure, psia 215 Accumulator Fluid Temperature, 'F 90 Total Number of Tubes per Steam Gener'ator 8519 Steam Generator Tube Plugging 33.8 - 24.8 % split Number of Tubes Plugged (33.8 % SGTP) 2878 Number of Tubes Plugged (24.8 % SGTP) 2114 Steam Generator Secondary Side Heat Transfer Area, 33.8% SGTP, ft2 48,661 SteamGeneratorSecondarySidegeat 55,245 Transfer Area, 24.8% SGTP, ft Steam Generator Secondary Flow Rate, Ibm /hr (47 53% power split) 5.241x10l(33.8%SGTP) 5.949 x 10 (24.8% SGTP)

Steam Generator Secondary Pressure, psia 730 Steam Geiterator Feedwater Enthalpy, Btu /lbm 414 Primary Heat Output used in RELAP4 EM Model - 1.02 x 2530 2580.6 MWt.

Includes pressurizer total volume and 29.3% average SGTP.

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e ANF 88-107 TABLE 3.3.1 PALISADES SYSTEM ANALYSIS PARAMETERS (CONTINUED)

Reactor Coolant Pump Rated Head, ft 260 Reactor Coolant Pump Rated Torque, ft-lbf 32,520 Reactor Coolant Pump Rated Speed, rpm 880

. Reactor goolant Pump Moment of Inertia, 98,000 lbe-ft Containment Volume, ft3 1.64 x 106 Containment Temperature, 'F 90 SIS Fluid Temperature, 'F 70 HPSI Dolay Time, Sec. 27.0 l LPSI Delay Time, Sec. 28.0 1

4 ANF-88-107 TABLE 3.4.1 PALISADES BREAK SPECTRUM ANALYSIS RESULTS DECLG 0.4 DECLG 0.6 DECLG 0.8 X/L - 0.6 X/L - 0.6 X/L - 0.6 15.28 15.28 15.28 Peak LHR (kW/ft) '

24.77 19.17 16.74 E08Y Time (Sec)

Fuel Average Temperature 1386.8 1424.3 1429.1 at E08Y (*F)

Cladding Temperature at 1176.8 1250.9 1245.1 E08Y (*F)

Hot Rod Burst

- 48.77 41.77 7.4 7.4 Time Elevation (Sec)(ft)

- Channel Blockage 0.32 0.32 Fraction Peak Cladding Temperature 1851.0 1913.7

- Temperature (*F) 52.47 57.57 7.4 7.4

- Time Elevation (Sec)( ft) ,

Metal Water Reaction '

l Local Maximuni, (%) 1.93 2.23 l

l - Elevation of Local 7.4 i Max. (ft) 7.4 r i 0.46 l - Hot Pin Total (%) 0.43* <l.0*

- Core Maxi' sum (%) <1.0 l l

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TABLE 3.4.2 CALCULATED EVENT TIMES FOR 0.4 DECLG BREAK gygni Time (Sec.)

Start 0.0 Break is Fully Open 0.05 Safety Injection Signal 0.81 Pressurizer Empties 12.6 Accumulator Injection Begins, Broken loop 18.5 Accumulator Injection Begini, Single Intact Loop 20.3 Accumulator Injection Begins, Double Intact Loop 20.3 End-of-Bypass (E0BY) 24.77 Start of Reflood 44.41 Peak Cladding Temperature is Reached (X/L - 0.6) 57.57 Accumulators Empty, Broken loop 74.74 Accumulators Empty, Single Intact Loop 77.41 Accumulators Empty, Double Intact Loop 78.35

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. l ANF-88-107 TABLE 3.4.3 CALCULATED EVENT TIMES FOR 0.6 DECLG BREAX Eygni Time (Sec.)

Start 0.0 Break is Fully Open 0.05 Safety Injection Signal 0.62 Accumulator Injection Begins, Broken Loop 11.70 Pressurizer Empties 12.26 Accumulator Injection Begins, Single Intact Loop 15.65 Accumulator Injection Begins, Double Intact Loop 15.65 End-of-Bypass (E0BY) 19.17 Start of Reflood 27.70 Peak Cladding Temperature is Reached (X/L = 0.6) 52.47 Peak Cladding Temperature is Reached (X/L = 0.8) 57.57 Accumulators Empty, Broken Loop 68.9 Accumulators Empty, Single Intact Loop 72.85 ,

Accumulators Empty, Double intact Loop 73.75 I .

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S The analysis with the current EXEM/PWR models for the Palisades plant confirms the 0.6 DECLG break size as the limiting break size. The analysis suppor*s operation of the Palisades plant at a power level of 2530 MWt, an increase in the total radial peaking factor from 1.77 to 1.83, and an average /

steam generator tube plugging level of 29.3% with a maximum asymmetry of 4.5%.

The analysis supports a peak LHR of 15.28 kW/ft with the axially dependent power peaking limit shown in Figure 2.1. The analysis supports Cycle 8 oper.. Lion and is intended to support operation for future cycles.

Operation of the Palisades plant with ANF 15x15 fuel at or below the LHR limits shown in Figure 2.1 assures that the NRC acceptance criteria (10 CFR 50.46(b)) for Loss-of Coolant Accident pipe breaks up to and including the double ended severance of a reactor coolant pi p will be met with the emergency core cooling system for the Palisades plant.

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5.0 REFERENCES

]

1 LOCA Analysis for Palisades at 2530 MWt usina the ENC WREM II PWR ECCS Evaluation Model, XN NF-77 24, Exxon Nuclear Company, Richland WA 99352, July 1977, 2

Dennis M. Crutchfield (USNRC Asst. Director division of PWR Licensing B)

"Safety Evaluation of Exxon Nuclear Company's large Break ECCS Evaluation Model EXEN/PWR and Acceptance for Referencing of Related Licensing Topical Reports', dated July 8,1986.

4

3. R00EX2r Fuel Rod Thermal Mechanical Resoonse Evaluation Model, XN NF 81-Exxon Nuclear Company, 58(P)(A), Revision 1, and Supplements 14, Richland, WA 99352, February 1983. .

}

}

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Issue Date:8/2/88 PALISADES LARGE BREAX LOCA/ECCS ANALYsl$ WITH INCREASED RADIAL PEAKING Distribution TH Chen i RA Copeland NF Fausz LJ Federico l RC Gottula JS Holm JW Hulsman JD Kahn j LA Neilsen LD O' Dell GL Ritter BE Schmitt HG Shaw (1)/ Customer (15)

EL Tolman HE Williamson Docuement Control (5) 1 I

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