ML19350C391

LOCA ECCS Reanalysis for DC Cook Unit 1 Using ENC Wrem 11A PWR ECCS Evaluation Model.

ML19350C391
Person / Time
Site:	Cook
Issue date:	02/12/1981
From:	Cherng J, Jensen S, Kayser W SIEMENS POWER CORP. (FORMERLY SIEMENS NUCLEAR POWER
To:
Shared Package
ML17319A796	List: ML17319A795 ML17319A798 ML19350C391
References
XN-NF-81-007, XN-NF-81-7, NUDOCS 8104010375
Download: ML19350C391 (38)

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LOCA ECCS REANALYSIS FOR D.C. COOK UNIT 1 USING THE ENC WREM.11A i PWR ECCS EVALUATION MODEL i

l FEBRUARY 1981

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XN-NF-81-07 ISSUE DATE: 02/12/81 LOCA ECCS REANALYSIS FOR D.C. COOK UNIT 1 USING THE ENC WREM IIA PWR ECCS EVALUATION MODEL Prepared by:

S.E. Jensen J.C. Cherng W.V. Kayser D.J. Braun .

Approved : /t / WA < # ~//' & /

J{p.' Morgan,Manhser l Licensing & Safety Engineering Approved [6 "'Ncp Nbi el eering

/mb ERON NUCLEAR COMPANY,Inc.

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1 i

i XN-NF-81-07 TABLE OF CONTENTS r

l

SECTION PAGE i

1.0 INTRODUCTION

AND SUPMARY . . . . . . . . . . . . . . . . . 1 2.0 ANALYTICAL AND SYSTEM MODELS . . . . . . . . . . . . . . . 5 i

7 3.0 SYSTEM ANALYSIS RESULTS . . . . . . . . . . . . . . . ..

28 4.0 FUEL EXPOSURE ANALYSIS RESULTS . . . . . . . . . . . . . .

................... 29

5.0 CONCLUSION

S . . . .

6.0 REFERENCES

. . . . . . . . . . . . . . . . . . . . . . . . 30 l

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11 XN-NF-81-07 LIST OF TABLES TABLE PAGE

!.1 0.C. COOK UNIT 1 EXPOSURE SENSITIVITY RESULTS . . . . . 3 3.1 D.C. COOK UNIT 1 REANALYSIS LIMITING BREAK EVENT TIMES (1.0 DECLS) . . . . . . . . . . . . . . . . . 8 LIST OF FIGURES FICURE PAGE 1.1 0.C. COOK UNIT 1, ALLOWABLE TOTAL PEAKING FACTOR AS A FUNCTION OF PEAK PELLET EXPOSURE . . . . . . . . . . 4 3.1 BLOWDOWN SYSTEM PRESSURE ................ 9 3.2 BLOWDOWN BREAK FLOW RATE ................ 10 3.3 PRESSURIZER SURGE LINE FLOW RATE ............ 11 3.4 ACCUMULATOR FLOW RATE TO IRTACT LOOPS . . . . . . . . . . 12 3.S BLOWDOWN CORE INLET FLOW RATE . . . . . . . . . . . . . . 13 3.6 BLOWOOWN CORE OUTLET FLOW RATE ............. 14 3.7 BLOWDOWN HOT ASSEMBLY INLET FLOW RATE . . . . . . . . . . 15 3.8 BLOWDOWN HOT ASSEMBLY OUTLET FLOW RATE ......... 16 17 3.9 BLOWDOWN PCT NODE CLADDING TEMPERATURE .........

3.10 BLOWDOWN PCT N0DE VOLUME AVERAGE- FUEL TEMPERATURE . . . . 18 3.11 PCT N0DE BLOWDOWN HEAT TRANSFER COEFFICIENT . . . . . . . 19 N

iii XN-NF-81-07 LIST OF FIGURES (contd.)

i FIGURE PAGE 3.12 PCT NODE BLOWDOWN DEPTH OF ZIRCONIUM-WATER REACTION . . . . . . . . . . . . . . . . . . . . . . . . 20 4

3.13 NORMALIZED CORE POWER ................. 21 3.14 ICECON COATAINMENT BACKPRESSURE ............ 22 i 3.15 REFLOOD UPPER PLENUM PRESSURE ............. 23 3.16 CORE REFLOODING RATE , . . . . . . . . . . . . . . . . . 24 4

3.17 REFLOOD DOWNCOMER MIXTURE LEVEL ............ 25 3.18 REFLOOD CORE MIXTURE LEVEL . . . . . . . . . . . . . . . 26 3.19 CLADDING SURFACE TEMPERATURE DURING HEATUP ENC FUEL BOL . . . . . ... . . . . . . . . . . . . . . . 27 k

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1 XN-NF-81-07

1.0 INTRODUCTION

AND

SUMMARY

In 1976, Exxon Nuclear Company (ENC) performed a LOCA ECCS analysis for ENC

' fabricated fuel in the Donald C. Cook Unit 1 reactor and established limits (I) assuring conformance to NRC 10 CFR 50.46 criteria (2) . The ENC WREM II PWR ECCS evaluation model was used for the 1976 0.C. Cook analysis. The ENC PWR evaluation

.model has since been updated to the ENC WREM IIA model, and the ENC ice condenser containment code, ICECON, has been reviewed and approved by the NRC. Several features of the latest ENC PWR ECCS Evaluation Model and removal of known over-conservatisms in input data have been shown to produce benefits with respect to the pravious D.C. Cook LOCA ECCS results. This report details a reanalysis for ENC fuel in the D.C. Cook Unit I reactor using the latest NRC approved ENC WREM IIA PWR ECCS evaluation model. The analysis results show compliance with NRC 10 CFR 50.46 criteria, and establish increased allowable LOCA ECCS operating limits for the ENC fuel in the D.C. Cook Unit I reactor.

The analysis consists of a recalculation of the previously established limiting break LOCA, the equivalent double-ended split break of the cold leg or reactor vessel inlet line (1.0 DECL9). A complete system calculation was performed including the'RELAP4-EM system blowdown, accumulator discharge, ICECON containment pressure, ae.d REFLEX reflood calculations. Multiple fuel heatup calculations were performed fer the ENC fuel at various linear heat rates and exposure conditions to establish allowable ECCS limits. These analyses consist of GAPEX, RELAP4-EM hot channel, and T000EE2 heatup calculations.

The effects of the NRC model for enhanced fission gas release and fuel rod pressure uncertainties are also considered in the analysis.

2 XN-NF-81-07 The results of the calculations are shown in Figure 1.1 which provides the revised maximum LOCA ECCS allowed peaking with exposure for ENC fuel in'the D.C. Cook Unit I reactor. Corresponding linear heat generation rates and ECCS results are given in Table 1.1.

Details of the analytical models used and revised input are described in Section 2.0. Section 3.0 shows the complete calculated results for the system analysis and the beginning-of-life ENC fuel heatup analysis. Final fuel analysis results at exposed conditions are given in Section 4.0.

The conclusion of the reanalysis is that, based on the LOCA ECCS analysis results shown, the D.C. Cook Unit 1 reactor can be operated with ENC fuel at or below the limits defined by Figure 1.1 and Table 1.1 which will assure conformance with the NRC 10 CFR 50.46 criteria and 10 CFR 50 Appendix K requi rements.

Table 1.1 D.C. Cook Unit 1 Exposure Sensitivity Results Peak Pellet Burnup (GWD/MTM) BOL 12.0 23.5 34.5 42.2 T 2.07 2.10 2.04 1.98 1.89 Total Peaking, Fg 14.24 14.45 14.03 13.62 13.00 Peak Linear Heat Generation Rate (Kw/ft) 2177 2195 2185 2186 Peak Clad Temperature (PCT, *F) 2199 Max. Local Zr/H 20-Reaction, % 6.42 6.09 6.25 5.95 .5.62 Core Wide Zr/H 0-Reaction, % <1% <1% <l% <1% <1% w 2

Hot Rod Burst Time, sec 47.5 70.9 73.5 85.7 101.1 4

Hot Rod Burst Location, ft. 6.0 6.25 6.5 6.5 6.75 Time of PCT, sec. 230 232 263 271 294 PCT Location, ft. 7.81 7.0 7.25 7.25 7.50 Max. Zr/H O Reaction Location, ft. 7.50 7.0 7.25 7.25 7.50 2

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Figure 1.1 D.C. Cook Unit 1, Allowable Total Peaking Factor as a Function of Peak Pellet Exposure

5 XN-NF-81-07 2.0 ANALYTICAL AND SYSTEM MODELS The D.C. Cook Unit i reanalysis used the ENC WREM IIA PWR ECCS Evaluation Model(3,4,5,6) as approved by the NRC and applied in several of the latest ENC PWR analyses. Only the CONTEMPT code which computes containment backpressure was replaced by the ICECON computer code as appropriate for the D.C. Cook Unit 1 ice condenser containment design. ICECON(7) has been reviewed and approved by the NRC(0) for LOCA ECCS application with ice condenser containments.

The 1976 LOCA ECCS analyses for D.C. Cook Unit 1 was performed with the ENC WREM II PWR ECCS evaluation model. During the development of the ENC WREM IIA model, several of the changes made were shown to give small improvements in ECCS margins for the D.C. Cook Unit I reactor. Model changes from the ENC WREM II model include: (1) RELAP4-EM updates from RELAP4-EM/ ENC 25 to RELAP4-EM/ ENC 28; (2) Replacing RELAP4-EM FLOOD with REFLEX; (3) Replacing the Westinghouse calculated reflood backpressure with the ENC ICECON calculated results; and (4) T00DEE2 updates from T00DEE2/76 to T00DEE2/DMAY79(9) .

In addition to the ENC WREM IIA updates, revised system input was obtained through ENC neutronics calculations for moderator density reactivity data.

A correction was also made to accumulator line dimensions input. The effect of this correction with regard to Peak Cladding Temperature (PCT) results is insignificant. The initial steady-state conditions were reestablished based on nominal conditions for core inlet temperature and secondary pressure as per a previous NRC request.

t 6 XN-NF-81-07 The system and fuel nodalization for the O.C. Cook Unit 1 reanalysis remains as in the previous analysis or as documented for REFLEX (5) and ICECON(7) Only a minor change in the T00DEE2 nodalization was made to center the maximum power node at the peak axial power location.

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l 3.0 SYSTEM ANALYSIS RESULTS The D.C. Cook Unit 1 ECCS reanalysis was performed for the previously identified limiting large break, the large split break of the reactor vessel recirculation inlet line or cold leg with the break area equal to twice the pipe cross sectional flow area. This break is referred to as the equivalent double-ended cold leg split break (1.0 DECLS). System behavior is essentially un-affected by exposure and the system calculation is performed for the highest stored energy beginning-of-life (80L) case.

Calculated event times for the ECCS reanalysis are given in Table 3.1.

RELAP4-EM system blowdown results are given in Figures 3.1 through 3.6. Figures 3.7 through 3.12 present results of the RELAP4-EM hot channel calculation.

Extended decay power is shown in Figure 3.13, and the ICECON computed containment pressure is given in Figure 3.14. REFLEX reflood results are shown in Figures T

3.15 through 3.18. T00DEE2 results for the BOL case with Fq of 2.07 are shown in Figure 3.19.

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8 XN-NF-81-07 Table 3.1 D.C. Cook Unit I Reanalysis Limiting Break Event Times (1.0 DECLS)

Event Calculated Event Time (sec)

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4.0 FUEL EXPOSURE ANALYSIS RESULTS Effects of exposed fuel conditions on ECCS analysis results were computed using RELAP4-EM for blowdown hot channel calculations and T000EE2 for fuel heatup and reflood analyses. BOL system boundary conditions were assumed.

The fuel analyses used the approved ENC Cladding swelling ind flow blockage model. Fission gas release was computed using GAPEX, and the NRC fission gas enhancement was employed at appropriate fuel exposures. The effects of pressure uncertainties were included in these calculations as in previous ENC analyses.

T Total peaking, Fq , and the corresponding peak linear heat generation rate (LHGR) were adjusted until 10 CFR 50.46 criteria were achieved. The final heatup analyses results for the exposures calculated are presented in Table 1.1.

The ECCS allowed peaking is limited to 2.07 at BOL due to the high initial stored energy. The allowed peaking increases with exposure as fuel swelling and cracking reduce stored energy. At high burnup, increased fuel rod swelling and flow blockage are calculated to occur due to increased fission gas release.

This results in reduced heat transfer during reflood and a corresponding reduction of allowed peaking limit with exposure accumulation.

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i 29 XN-NF-81-07 l

5.0 CONCLUSION

S The reanalysis of the limiting break (1.0 DECLS) for the O.C. Cook Unit i reactor with the ENC WREM IIA PWR ECCS evaluation model shows that the reactor can operate at increased allowed peaking and continue to meet the NRC 10 CFR 50.46 criteria with analyses performed in conformance to 10 CFR 50 Appendix K requi rements. Operation within the ECCS allowed limits as defined in Figure 1.1 and Table 1.1 assures that the NRC acceptance criteria are met. That is:

(1) The calculated peak fuel cladding temperature does not exceed 2200'F.

(2) The calculated local cladding oxidation does not exceed 17% of the cladding thickness during or after quenching, and the temperature transient is terminated while the core geometry is amenable to cooling.

(3) The calculated core-wide reaction of cladding with water or steam does not exceed 1% of the total mass of zircaloy in the reactor.

(4) System long term cooling capabilities provided for previous cores will also cool ENC fueled cores.

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6.0 REFERENCES

(1) Exxon Nuclear Company, Donald C. Cook Unit 1 LOCA Analyses Using the ENC WREM-Based PWR ECCS Evaluation Model (ENC WREM-II) , XN-76-51, October 1976 and Supplements; Flow Blockage and Exposure Sensitivity Study for D.C. Cook Unit 1 Reload Fuel Using ENC WREM-II Model ,

XN-76-51 Supplement 1, January 1977; XN-NF-76-51(P) Supplement 2 January 1978; XN-NF-76-51(P) Supplement 3, March 1978.

(2) U.S.N.R.C. Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors ,10 CFR 50.46 and Appendix K to 10 CFR 50. Federal Register, Volume 39, Number 3, January 4,1974.

(3) Letter, G.F. Owsley (ENC) to D.F. Ross (NRC), Description of RELAP4-EM ENC 288, dated October 30, 1978.

(4) Letter, Thomas A. Ippolito (NRC) to Warren S. Nechodom (ENC),

ENC-EM Update Evaluation, March 1979.

(5) Exxon Nuclear Company, Exxon Nuclear Company WREM-Based Generic PWR ECCS Evaluation ?odel Update ENC WREM-IIA , XN-NF-78-30, August 19'8, and XN-NF-78-30 Amendment 1 February 1979.

(6) Letter, Thomas A. Ippolito (NRC) to Warren S. Nechodom (ENC, Topical Report Evaluation, dated March 30, 1979.

(7) Exxon Nuclear Company, ICECON: A Computer Program Used to Calculate Containment Break Pressure for LOCA Analysis (Including Ice Condenser Plants) , XN-CC-39 Rev. 1, November 1977.

(8) Letter,ThomasA.Ippolito(NRC)toWarrenS.Nechodom(ENC),

Topical Report Evaluation, dated June 30, 1978.

(9) Letter, G.F. Owsley (ENC) to D.F. Ross (NRC), Updates of T00DEE2 Program, dated April 1, 1980.

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l XN-NF-81-07 ISSUE DATE: 02/12/81 LOCA ECCS REANALYSIS FOR D, C. COOK UNIT 1 USING THE ENC WREM 11A PWR ECCS EVALUATION MODEL 4

Distribution D.J. Braun J.C. Cherng R.E. Collingham G.C. Cooke S.L. Garrett K.P. Galbraith S.E. Jensen J.D. Kahn W.V. Kayser J.E. Krajicek J.N. Morgan G.F. Owsley G.A. Sofer H.E. Williamson American Electric Power (5)/H.G. Shaw Document Control (10)

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