ML20087J831

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LOCA-ECCS Analysis for 2125 Mwt Operation & 50% Steam Generator Tube Plugging
ML20087J831
Person / Time
Site: Palisades Entergy icon.png
Issue date: 03/09/1984
From: Tahvili T
SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML18051A825 List:
References
XN-NF-84-24, NUDOCS 8403230111
Download: ML20087J831 (54)
v • d • e


Text

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L XN NF 84 24

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[ PAllSADES LOCA-ECCS ANALYSIS

[ FOR 2125 MWt OPERATION AND 50% STEAM GENERATOR

TUBE PLUGGING I

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l ERON NUCLEAR COMPANY, INC.

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! l XN-NF-84-24

( issue Date: 3/9/84 E

PALISADES LOCA-ECCS ANALYSIS FOR 2125 MWt OPERATION AND 50% STEAM GENERATOR TUBE PLUGGING i

Prepared by: .2 A' ~

3/7/#f T. Tahvi'i PWR Safety Analysis

)

l Concur: W Ow YY9 Y _

W.V. Kayser, Manager PWR Safety A alysis Concur: .

J.C. Chandler ud 3ht+ '

el5adfuelLicensing Concur: Vh.^- % ~j JLN. Morgan, ManWger

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Proposals ustomer Service Engineering >

Approve: / 7 m A/W l R.B. Stout, Manager Licensing Safe y & Engineering Approve: $ $ ~.wm 13///W/

"G'.' A. Sofer, Manager /

Fuel Engineering & Technical Services l

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[ E@ NUCLEAR COMPANY,Inc.

XN-NF-84-24 k TABLE OF CONTENTS Section Page 1.0 I NTRODUCT ION AND %MMARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 L IMITING BREAK L0i A ANALYSIS. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 2.1 LOCA ANALYS'.S M0 DEL................................... 4 2.2 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION..... 6 2.3 RESULTS............................................... 7 3 . 0 CO N C L U S I O N S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47

4.0 REFERENCES

................................................. 48 l

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XN-NF-84-24 s

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> List of Tables Table Page 1.1 Palisades Nuclear Reactor LOCA-ECCS Analysis Results...................................................... 3 2.1 Input Data For Reduced Power 0peration....................... 8 2.2 Nominal Fuel Design Parameters (Batch J Design) . . . . . . . . . . . . . . 9 2.3 Palisades Large Break Event Times - 0.6 DEG/PD Break. . . . .. . . 10 4

6 4

i XN-NF-84-24 List of- Figures Figure Page 2.1 RELAP4/EM Blowdown System hodalization for Palisades Nuclear plant.............................................. 11 2.2 REFLEX Nodalization EXEM/PWR, Palisades Nuclear Plant...... 12 2.3 Axial Power Profile With Peak at X/L=0.6 and with Skewing Factor of 1.0...................................... 13 2.4 Downcomer Flow Rate During Blowdown Period, 0.6 DEG/PD Break............................................... 14 2.5 Upper Plenum Pressure During Blowdown Feriod, 0.6 DEG/PD Break............................................... 15 2.6 Average Core Inlet Flow During Blowdown Period, 0.6 DEG/PD Break............................................... 16 2.7 Average Core Inlet Enthalpy Durina Blowdown Period,

0. 6 D E G / P D B r e z.'< . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 2.8 Average Core Outlet Flow During Blowdown Period, 0.6 DEG/PD Break........ .................................. 18 2.9 Average Core Outlet Enthalpy During Blowdown Period, 0.6 DEG/PD Break........................................... 19 2.10 Total Break- Flow During Blowdown Period, 0.6 DEG/PD Break...................................................... 20 2.11 Vessel Side Break Enthalpy During Blowdown Periad, 0.6 DEG/PD Break............................................... 21 2.12 Pump Side Break Enthalpy During Blowdown Period, 0.6 DEG/PD Break........................................... 22 2.13 Flow From Two Intact Accumulators During Blowdown Period, 0.6 DEG/PD Break................................... 23 2.14 Flow From Broken Side Intact Accumulator During Blowdown Period, 0.6 DEG/PD Break.......................... 24 2.15 Flow From Broken Loop Accumulator During Blowdown Period, 0.6 DEG/PD Break................................... 25

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4

XN-NF-84-24 List of Figures (Cont.) -

Figure Page 2.16 Pressurizer Surgeline Flow During Blowdown Period, 0.6 DEG/PD Break........................................... 26 2.17 Hot Assembly Inlet Flow During Blowdown Period, -

0.6 DEG/PD Break........................................... 27 2.18 Hot Assembly Outlet Flow During Blowdown Period, 0.3 DEG/PD Break........................................... 28 2,19 Heat Transfer Coefficient During Blowdown Period at PCT Location, 0.6 DEG/PD Break.......................... 29 2.20 Clad Surface Tenicerature During Blowdown Period at PCT Location, 0.6 DEG/PD Break.......................... 30 2.21 Depth of Metal-Water Reaction During Blowdown Period at PCT Location, 0.6 DEG/PD Break.......................... 31 2.22 Average Fuel Temperature During Blowdown Period at PCT Location, 0.6 DEG/PD Break.......................... 32 2.23 Two Intact Loop Accumulator Flow During Refill and Reflood Periods, 0.6 JEG/PC Break...................... 33 2.24 Single Intact Loop Accumulator. Flow During Refill and Reflood Periods, 0.6 DEG/PD Break...................... 34 2.25 Broken Loop Accumulator Flow During Refill and Reflood Periods, 0.6 DEG/PD Break.................................. 35 2.26 Intact Loops HPSI Flow During Refill and Reflood Periods, 0.6 DEG/PD Break.................................. 36 2.27 Broken Loop HPSI Flow During Refill and Reflood Periods, 0.6 DEG/PD Break.................................. 37 2.?8 Intact Loops LPSI Flow During Refill and Reflood Periods, 0.6 DEG/PD Break.......................... 38 2.29 Broken Loop LPSI Flow During Refill and Reflood Periods, 0.6 DEG/PD Break.................................. 39 2.30 Containment Back Pressure, 0.6 DEG/PD Break. . . . . . . . . . . . . . . . 40

XN-NF-84-24 i List of Figures (Cont.)

Figures Page

2. 31 Normal ized Power, 0.6 DEG/PD Break . . . . . . . . . . . . . . . . . . . . . . . . . 41 2.32 Reflood Core Mixture Level, 0.6 DEG/PD Break. . . . . . . . . . . . . . . 42 2.33 Reflood Downcomer Mixture Level, 0.6 DEG/PD Break. . . . .. .. .. 43 2.34 Reflood Upper Plenum Pressure, 0.6 DEG/PD Break. . .... .... . . 44 2.35 Core Flooding Rate, 0.6 DEG/PD Break. . . . . . . . . . . . . . . . . . . . . . . 45 2.36 T00DEE2 Cladding Temperature vs. Time, 0,6 DEG/PD Break...................................................... 46

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( XN-NF-84-24 L

I 1.0 INTRODUCTIO*! A?ID

SUMMARY

This document presents analytical results for a postulated large break loss-of-coolant accident (LOCA), performed for the Palisades nuclear reac-tor. The analyses assume: (a) a reactor operating power of 2167.7 MWt (2125.2 MWt + 2% power uncertainty) which is 84% of the rated power, (b) an increase in steam generator tube plugging to an average of 50% (40% in Loop 1 and 60% in Loop 2), and (c) an all Exxon Nuclear Company core (Batch J fuel design). The calculations were made for the double-ended cold leg guillotine break located at the pump discharge with a discharge coefficient of 0.6 (0.6 DEG/PD) which was previously identified as the most limiting break (1,2,3) ,

The analyses were performed using the EXEM/PWR ECCS evaluation model(4) ,

with the RODEX2 computer model(5) for evaluating the rod stored energy and fission gas release. The EXEM/PWR ECCS evaluation model includes the NRC fuel swelling and flow blockage model, contained in NUREG-0630(6) . The analyses are applicable up to an average of 50 percent steam generator (SG) tube plugging and maximum peak pellet exposure limit of 47,000 MWD /MTM. The

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allowable linear heat generation rate was 12.84 kW/ft (12.59 Kw/f t times 1.02 multiplier for power uncertainty), corresponding to a total power peaking factor of 2.76 (FgT ), and radial peaking factor of 1.45 with a local paking factor of 1.224. The axial power distribution is based on a skewed profile with a peak magnitude of 1.51 located at 60 percent of active core height (2) ,

The calculational basis and results are summarized in Table 1.1. The maximum calculated peak cladding temperature (PCT) is 15670F, occurring 160

(

seconds into the accident at a location 8.20 feet from the bottom of the active core, with the maximum Iocal metal-water reaction less than one -

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2 XN-NF-84-24 L

percent. The results of the analyses show that 10 CFR 50.46 and Appendix K criteria are satisfied for operation of the Palisades Nuclear Power Station within the established peaking limits, at the stated power level, and with 50%

average steam generator tube plugging.

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- 3 XN-NF-84-24 1

Table 1.1 Palisades Nuclear Reactor LOCA-ECCS Analysis Results Analysis Results Peak Pellet Exposure Peak Clad Temperature (PCT) 1567 Time of PCT, sec. 160 Peak Clad Temperature Location, ft. 8.20 Local Zr/H 20 Reaction (max.), %* 0.59 Local Zr/H20 Location, ft. from bottom 7.97 Total H2 Generation, % of total <1.0 Zr Reacted Calculational Basis License Core Power, MWt 2125.2 Power Used in Analysis, MWt** 2167.7 Peak Linear Fower for Analysis, kW/ft** 12.84 Total Peaking Factor, Fg T 2.76 i

Radial Peaking Factor FR 1.45 Local Peaking Factor FL 1.224 Peak Axial power 1.51 Z/L 0.6 Skewing Factor 1.0 Steam Generator Tube Plugging 40% Loop 1, 60% Loop 2 I

  • Computed value at 330 seconds.
    • Including 1.02 factor for power uncertainties.

4 XN-NF-84-24 a

2.0 LIMITING BREAK LOCA ANALYSIS r This report provides LOCA-ECCS analyses performed for Palisades nuclear reactor with an average steam generator tube plugging of up to 50%. The analytical techniques used are in compliance with Appendix K of 10 CFR 50, and are described in the ENC WREM models(8) , and the Einergency Core Cooling System Evaluation Model Updates: WREM-II I9) , and EXEM/PWR(4) .

A LOCA break spectrum analysis was performed and reported in XN-NF 24(I) . The 1?miting LOCA break was determined to be a large double-ended guillotine break of the pump discharge, with a discharge coefficient of 0.6 (0.6 DEG/PD). The current analysis assumes 84% of the rated power, reduced primary pressure and flow and an average steam generator tube plugging level

" of 50%. The secondary system pressure and flow are correspondingly reduced.

Table 2.1 shows the new system parameters for reduced power operation. The analyses performed and reported herein for the 0.6 DEG/PD break used:

(1) The revised stored energy model RODEX2(5) in place of the pre-viously applied GAPEX(11) model.

(2) Applicable EXEM/PWR models(4) .

(3) The ENC /NUREG-0630 clad swelling and rupture model(6) ,

(4) The REFLEX code for reflood calculations (10) ,

(5) The FLECHT/ ENC 2 WREM(8) heat transfer coefficient multipliers.

2.1 LOCA ANALYSIS MODEL The Exxon Nuclear Cconpany EXEM/PWR ECCS evaluation model I4) es used to perform the analyses. This model consists of the following computer l

codes: RODEX2(b) code for initial rod stored energy and internal fuel rod gas inventory; RELAP4-EM(12) for the system blowdown and hot channel blowdown I

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5 XN-NF-84-24 I ) for computation of calculations; CONTEMPT-LT/22 as modified in CSB 6-1 containment back pressure; REFLEX (4,6) for comparison of system reflood; and 100DEE2(4,6,14) for the calculation of final fuel rod heatup.

The reactor coolant system is nodalized into control volumes representing reasonably homogeneous regions, interconnected by flow-paths or I "junctons". The system nodalization is depicted in Figure 2.1 while the REFLEX nodalization is shown in Figure 2.2. The pump performance charac- L teristic of Combustion Engineering pump were used with analysis.III Asymmetric steam generator tube plugging is assuaed such that Loop 1 is 60%

plugged and Loop 2 is 40% plugged. The break is assumed to have occurred in ,

the most highly plugged loop since this results in higher peak clac -

temperatures. The transient behavior was determined from the governing conservation equations for mass, energy, and momentum. Energy transport, flow rates, and heat transfer are determined from appropriate. correlations.

System input parameters are given in Table 2.1.

The reactor core is modeled with heat generation rates determined from reactor kinetics equations with reactivity feedback and with decay heating as required by Appendix K of 10 CFR 50. The axial power profile used for the analyses is shown in Figure 2.3 with a maximum axial peaking factor of 1.51 peaked at 60% of the core height, corresponding to a total peaking factor of 2.76, and a radial peaking of 1.45 with local peaking factor of f 1.224.

l The ECCS calculations were performed with input which bounas the fuel history up to 47000 MWD /MTU. The most limiting fuel conditions from beginning-of-life to end-of-life exposures were determined and used in the calculations. Internal rod pressure and decay power are highest at E0L for

i 6 XN-NF-84-24 a

the rod while stored energy is highest at about 2000 MWD /MTM. This combination of highest stored energy, highest rod pressure, and higher decay power were used in the analysis to bound operation of the Palisades reactor for the entire fuel exposure history.

2.2 IDENTIFICATION OF CAUSES AND ACCIDENT DESCRIPTION For the purpose of LOCA analyses, a loss of coolant accident is defined as a rupture of the Reactor Primary Coolant System piping including the double-ended rupture of the largest pipe in the Reactcr Coolant System or of any line connected to that system up to the first closed valve.

Should a major break occur, depressurization of the Reactor Coolant System results in a pressure decrease in the pressurizer. A reactor trip signal occurs when the pressurizer low pressure trip setpoint is reached.

Reactor trip and scram were conservatively neglected for the large break analyses. A Safety Injection System signal is actuated when the appropriate setpoint (high containment pressure) is reached. These countermeasures will limit the consequences of the accident in two ways:

1. Reactor trip and borated water injection complements void formation in causing rapid reduction of power to a residual level corresponding to fission product decay heat.
2. Injection of borated water enhances heat transfer from the reactor core and prevents excessive clad temperatures.

l At the beginning of the blowdown phase, the entire Reactor Coolant System contains subcooled liquid which transfers heat from the core by forced convection cooling. After the break develops, the time to departure from nucleate boiling is calculated, consistent with Appendix K of 10 CFR 50.(7)

Thereafter, the core heat transfer is unstable, with both transition and film .

I

7 XN-NF-84-24 boiling occurring. As the core becomes uncovered, both turbulent and laminar I forced convection to steam are corsidered as core heat transfer mechanisms.

When the Reactor Coolant System pressure falls below 215 psia the _

accumulators begin to inject borated water. The conservative assumption is made that accumulator ECC water bypasses the core and goes out through the break until the termination of bypass. This conservatism is again consistent with Appendix K of 10 CFR 50.

2.3 RESULTS Table 2.3 presents the timing and sequence of events as determined for the large guillotine break with a discharge coefficient of 0.6 located at pump discharge. Comparison of these results with the previous LOCA-ECCS analysis (1,2,3) for E, G, and H fuel loadings show an improvement in the LOCA-ECCS analysis due to the reduction in the LHGR. Figures 2.4 through 2.18 present plotted results for system blowdown analysis. Unless otherwise noted on the figures, time zero corresponds to the time of break initiations.

Figures 2.19 through 2.22 present results for - the hot channel blowdown calculations. Figures 2.23 through 2.29_present the accumulators, HPSI and LPSI flows during the refill and reflood periods of LOCA transient. Figure 2.30 presents calculated containment backpressure time history. Figure 2.31 shows the normalized power calculation results. The reflood calculation results are shown in Figures 2.32 through 2.35.

The maximum peak cladding temperature (PCT) calculated for the 0.6 DEG/PD break is 15670F, Figure 2.36, which is applicable for the entire fuel exposure. The maximum linear heat generation rate is 12.84 kW/f t (FgT =2.76).

I The maximum local metal-water reaction is less.than 1%. The PCT location is at an elevation of 8.20 feet from the bottom of active core.

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8 XN-NF-84-24 s

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Table 2.1 Input Data For Reduced Power Operation l Reactor Power (MWt) 2167.7 Primary System Pressure (psia) 1950 Average Coolant Temperature (OF) 564 I Reactor Coolant Flow (M1bm/hr) 99 Average S.G. Tube Plugging (%) 50 Loop 1 Loop 2 S. G. Tube Plugging (%) 60 40 S.G. Primary Fluid Volume (ft3) 403.7 605.6 S.G. Primary Flow Area (ft2 ) 7.9 11.9 S.G. Heat Transfer Area (inside) (ft 2) 32396 48596 S.G. Heat Transfer Area (outside) (f t )2 37152 55728 Secondary Pressure (psia) 600 600 I

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9 XN-NF-84-24 Table 2.2 Nominal Fuel Design Parameters (Batch J Design)

Parameter Dimension Cladding, 0.0., in. 0.417 Cladding, I.D., in. 0.358 1 Cladding Thickness, in. 0.0295 Pellet 0.D., in. 0.350  ;

Diametral Gap, in. 0.0080 Pellet Density, % TD 94.0 Active Fuel Length, in. 131.8 ,

Enriched U0 2, in. 131.8 Rod Pitch, in. 0.55 No. of Fuel Rods / Assembly 208 No. of Fuel Assemblies 204 y

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s 10 XN-NF-84-24

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Table 2.3 Palisades Large Break Event Times -

0.6 DEG/PD Break l Event Time (sec)

Start Transient 0.0 ,

Initiate Break 0.1 Safety Injection Signal C.83 Accumulator Injection, Broken Leg 10.80 Accumulator Injection, Intact Leg 17.20 Accumulator Injection, Intact Loop 17.20 End-of-Blowdown (Break Flow Reversal) 20.20 Safety Pump Injection HPSI 21.83 End-of-Bypass (Downcomer Flow Reversal) 25.97 Safety Pump Injection LPSI 28.83 Bottom of Core Recover 43.42 Start of Reflood 44.45 l Accumulator Empty, Broken Leg 68.61 Accumulator Empty, Intact Leg 75.41 Accumulator Empty, Intact Loop 75.66 Peak Clad Temperature Reached 160 l

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3.0 CONCLUSION

For pipe breaks up to and including the double-ended severance of a reactor coolant pipe, the Emergency Core Cooling System for the Palisades nuclear reactor will meet the Acceptance Criteria as presented in 10 CFR 50.46 when calculated according to the requirements of Appendix K, with a total

( peaking factor of 2.76 (Fg )T and the operating parameters shown on Table 2.1.

The criteria are as follows:

(1) The calculated peak fuel element clad temperature does not exceed the 22000F limit.

(2) The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of zircaloy in

(

the reactor.

(3) The cladding temperature transient is terminated at a time when the core geometry is still amenable to cooling. The hot fuel rod cladding oxidation limits of 17% are 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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48 XN-NF-84-24 r

L I

4.0 REFERENCES

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1. "LOCA Analysis for Palisades at 2530 MWt Using the ENC WREM-IT PWR ECCS Evaluation Model", XN-NF-77-24, Exxon Nuclear Company, July 1977.

I

2. " Analysis of Axial Power Distritsution Limits for the Palisades Nuclear Reactor at 2530 MWt", E!-NF-78-16, Exxon Nuclear Company, June 1978.
3. "ECCS and Thermal-Hydraulic Analysis for the Palisades Reload H Design",

XN-NF-80-18, Exxon Nuclear Ccmpany, April 1980.

4. " Exxon Nuclear Company Evaluation Model EXEM/ PWR ECCS Model Updates",

XN-NF-82-20, Revision 1, Exxon Nuclear Company, August 1982; Supplement 1, March 1982; and Supplement 2, March 1982.

5. "RODEX2: Fuel Rod Thermal-Mechanical Response Evaluation Model", XN-hF-81-58(A), Revision 2, Exxon Nuclear Company, February 1983.

( 6. " Exxon Nuclear Company ECCS Cladding Swelling and Rupture Model", XN-NF- a 82-07(A), Revision 1, Exxon Nuclear Company, August 1982.

7. " Acceptance Criteria for Emergency Core Cooling Systems for Light Vater Cooled Nuclear Power Reactors," 10 CFR 50.46 and Appendix K of 10 CFR 50.

f 8. " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model", -

XN-NF-75-41, Exxon Nuclear Company, July 1975, and Supplements and Revisions thereto.

9. " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model Update EC WREM-Il", XN-76-27(A) and Supplements 1 and 2, Exxon Nuclear Company, 1976.
10. " Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation Model -,

Update ENC WREM-IIA", XN-NF-78-30(A), Exxon Nuclear Company, May 1979.

( 11. "GAPEXX: A Computer Program for Predicting Pellet-to-Cladding Heat u Transfer Coefficients", XN-73-25, Exxon Nuclear Company, August 13, 1973.

12. U.S. Nuclear Regulatory Commission Letter, T.A. Ippolito (NRC) to W.S.

Nechodom (ENC), "SER for ENC RELAP4-EM Update," March 1979.

13. U.S. Nuclear Regulatory Commission, " Minimum Containment Pressure Model for PWR ECCS Performance Evaluation," Branch Technical Position CSB 6-1.
14. G. N. Lauben, "T00DEE2: A Two-Dimensional Time Dependent Fuel Element Thermal Analysis Program," NRC Report NUREG-75-057, May 1975.

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4 XN-NF-84-24 r ISSUE DATE 3/9/84

f. PALISADES LOCA-ECCS ANALYSIS FOR 2125 MWt OPERATION AND 50% STEAM GENERATOR TUBE PLUGGING i .

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f- DISTRIBUTION F.T. Adams -

J.C. Chandler R.A. Copeland J.S. Holm i

W.V. Kayser l

W.T. Nutt H.G. Shaw R.B. Stout

( T. Tahvili Consumers Power Company /H.G. Shaw (10)

Document Control (5)

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