ML17318A556
| ML17318A556 | |
| Person / Time | |
|---|---|
| Site: | Cook  |
| Issue date: | 01/11/1980 |
| From: | Hunter R INDIANA MICHIGAN POWER CO. (FORMERLY INDIANA & MICHIG |
| To: | Harold Denton Office of Nuclear Reactor Regulation |
| References | |
| AEP:NRC:00322C, AEP:NRC:322C, NUDOCS 8001150426 | |
| Download: ML17318A556 (90) | |
Text
REGULATORY INFORMATION DISTRIBUTION SYSTEM (RIDS)
ACCESSION NBR:8001150426 DOC ~ DATt: 80/01/11 NOTARIZED:
NO DOCKET FACiL:50".516 Oonaid
(:,
Cook Nuclear Power Plantr Onit 2i.Indiana 8
05000316 AOYH ~ NAME AU fHOB AFFILIATION HUNTERrR ~ ST Indiana 8 Hichigan Electric Co,
- RECAP ~ NAME HEC lP IDENT AFF ILLATION DENN~rH ~ RE Office oi'uclear Reactor Regulation
SUBJECT:
Forwards ECCS analysis using approved Westinghouse Feb 1978 evaluation model,ECCS wil) meet acceptance criteria presented in code for bieaks up to 8 including double ended severance of reactor cooling pipe, DISTRIBUTION CODE:
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BOX 18 BOWLING GR KEN STATION NEW YORK, N. Y. 10004 January 11, 1980 AEP:NRC:00322C Donald C.
Cook Nuclear Plant Unit No.
2 Docket No. 50-316 License No.
Subject:
Peaking Factor Limits at Unit 2 In Light of Revised Analysis Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, D.C.
20555
Dear Mr. Denton:
The purpose of this letter is to formally transmit to'ou information which was presented at a meeting held on January 10, 1980 and which was attended by members of your staff, by AEP and by Westinghouse Electric Corporation.
Attachment A contains an ECCS analysis for Unit 2 of the Cook Plant using the approved Westinghouse February 1978 evaluation model.
This analysis resulted in a total F
of 2.02 while our previously submitted and approved analysis using the October 1975 model with the.Correct metal-water reaction de-scription yielded an Fq of 2.11.
The data explaining this reduction in Fq when using the newer model is contained in Attachment B.
The February 1978 -analysis and the change in peaking factors associated with its. use-was discussed in detail at the referenced meeting wherein all attendees agreed that the February 1978 analysis is valid.
In light of the data presented in draft NUREG-0630, the total Fq for Unit 2 was further reduced by 0.03 as explained in our letters of January 8,
1980 (AEP:NRC:00322A) and January 9,
1980 (AEP:NRC:00322B).
We have accepted this further reduction in Fq to 1.99'voluntarily.
80 Oy1 50 4
. go~"
~ C'i k
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Nr. Harold R. Denton, Director AEP:NRC:00322C Finally, we have reviewed the revised Technical Specification Figure 3.2-2 submitted in AEP:NRC:00322B.
This review confirms that the submitted figure is properly based on the February
- 1978, Unit 2 specific large LOCA analysis and the Unit 2 small break analysis previously submitted and approved (FSAR Section 14.3.'2),
and is therefore correct.
ery tr ly your RSH:em R. S. Hunter Vice President cc:
R.
C. Callen G. Charnoff J.
E. Dolan R.
W. Jurgensen D. V. Shaller-Bridgman
ATTACHMENT A TO AEP:NRC'00322C
The Loss of Coolant Accident (LOCA) has been reanalyzedfor sj
~
-2 The foll.owing information amends the Safety Analysis Report section on Major Reactor Coolant System Pipe
'Ruptures.
The, results are consistent with acceptance criteria provided in reference 1.
The description of the various aspects of the LOCA analysis is given in'CAP-8339 The individual.computer codes which comprise the
[2]
Mestinghoush Emergency Core Cooling System (ECCS) evaluation model are described in detail in separate reports a'long with code modifi-
[3-ag.
cations specified in references 7,
9 and 10.
The analysis presented here was performed with the February 1978 version of the evaluation model which includes modifications delineated in references 11, 12, 13 and 14.
A ~
4 C
The analysis of the loss of coolant. accident is performed=at 102 percent of the licensed core power'ating.
The peak linear power and total core I
power used in the analysis are given in Table 2.-
Since there is margin between the value of peak linear power density used in this analysis and the value of the peak linear power-density expected during plant opera-tion, the peak clad temperature calculated in= this analysis is greater than the maximum clad temperature expected to exist.
Table 1 presents the occurence time for various events throughout the accident transient.
Table 2 p'resents selected input values and results from the hot fu'el rod thermal transient calculation.
For these results, the hot spot is defined as the location of maximum peak clad temperatures.-
That loca-tion is specified in-Table 2 for each break analyzed.
The location is indicated in feet which presents elevation above the bottom of the active fuel stack.
Table 3 presents a summary of the various containment systems parameters Lo< IC.
and structural parameters which were used as input to the CR computer code
- used in this analysis.
l Tables 4 and 5 present reflood mass and energy releases to the contain-
'ment, and the broken loop accumulator mass and energy release to the containment, respectively.
The results of several sitivity studies are reported These results are for conditions which are not limiting in nature and hence are reported on a generic basis.
Figures 1 through 17 present the transients. for the principle parameters for the break sizes analyzed.
The following items are.noted:
Figures 1A - 3C:
guality, mass'elocity and clad heat transfer coeffi-cient for the hotspot and burst locations Figures 4A - 6C:
Core pressure, break flow, and core pressure drop.
The break flow is the sum of -the flowrates from both ends of the guillotine break.
The core pressure drop is taken as the pressure just before the core inlet to the pressure just beyond the core outlet Figures 7A - 9C:
Clad temperature, fluid temperature and core flow.
The clad and fluid temperatures are for the hot spot and burst locations Figures 10A - 11C:
Oowncomer and core water level during reflood, and flooding rate Figures 12A - 13C:
Emergency core cooling system flowrates, for both accumulator and pumped safety injection
0 Figures 14A - 15C: Containment pressure and core power transients Figures 16, 17:
Break energy release during blowdown and the contain-ment wall condensing heat transfer coefficient for the worst break
Conclusions - Thermal Anal sis For breaks up to and including the double ended severance of a reactor coolant pipe, the Emergency Core Cooling System will meet the Acceptance Criteria as presented in 10CFRg).46.~ j That is:
1.
The calculated peak clad temperature does not exceed 2200 F based on'a total core peaking factor of
~ O 2.-
The amount of fu'el element cladding that reacts chemically with water or steam does not exceed 1 percent of the total amount of Zircalloy in the reactor.
3.
The clad temperature transient is terminated at a time when the core geometry is still amenable to cooling.
The cladding oxidation limits. of 17K 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.
I J
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References for Section
~.4.1 e
1..
"Acceptance Criteria for Emergency Core Cooling Systems for Light Mater Cooled Nuclear Power Reactors",
10CFR50.46 and Appendix K of 10CFR50.46.
Federal
- Register, Volume 39, Number 3, January 4, 1974.
2.
- Bordelon, F. M., Massie, H. W., And Zordan, T. A., "Westinghouse ECCS Evaluation Model-Summary,"
WCAP-8339, July 1974.
3.
- Bordelon, F. M., et al.,
"SATAN-VI Program:
Comprehensive Space-Time Dependent Analysis of Loss-of-Coolant,"
WCAP-8302 (Pro-pr ietary Version),
WCAP-8306 (Non-Proprietary Version),
June 1974.
, 4.
- Bordelon, F. M., Et al.,
"LOCTA-IV Program:
Loss-of-Coolant Tran-
'sient Analysis," WCAP-8301 (Proprietary Version),
WCAP-8305 (Non-Proprietary Version),
June 1974.
5.
Kelly, R. D., et al., "Calculational Model for Core Ref looding after a Loss-of-Coolant Accident (WREFLOOD Code)."
WCAP-8170 (Proprietary Version),
MCAP-8171 (Non-Proprietary Version),
June 1974.
I
'sieh, T.
8 Raymond M.," Long Term Ice Condenser Containment-LOTIC Code 5m)
SuPPlement 1,"
WCAP 8355 Supplement 1,
Way IP75.
WCAP-8354 (Proprietary), July 1974.
Letter from K. Kniel to J. Tillinghast, December 27, 1974-
"Order for Modification of License for the Donald C.
Cook Nuclear Plant Unit l."
~ 7.
Bordelon F. M., et al., "The Westinghouse ECCS Evaluation Model:
Supplementary Information," WCAP-8471 (Proprietary Version),
WCAP-8472 (Non-Proprietary Version), January 1975.
8.
0 Salvatori, R., "Westinghouse ECCS - Plant Sensitivity Studies,"
WCAP-8340 (Proprietary Version),
WCAP-8356 (Non-Proprietary Ver-sion), July 1974..
I
- 9. "Westinghouse ECCS Evaluation Model, October, 1975 Versions,"
WCAP-8622 (Proprietary Version),
WCAP-8623 (Non-Proprietary Version),
- November, 1975.
- 10. Letter from C. Eicheldinger of Westinghouse Electric Corporation to D. S. Vassalo of the Nuclear Regulatory Commission, letter number NS-CE-924, January 23, 1976.
- 11. Kelly, R. D., Thompson, C. M., et, al., "Westinghouse Emergency Core Cooling System Evaluation Model for Analyzing Large LOCA's During Operation With One Loop Out of Service for Plants Without Loop Iso-lation Valves," WCAP-9166, February, 1978.
- 12. Eicheldinger C., "Westinghouse ECCS Evaluation Model, February 1978 Version," WCAP-9220-P-A (Proprietary Version),
WCAP-9221-A (Non-Pro-prietary Version), February, 1978.
- 13. Letter.from T. M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-.1981, Nov. 1, 1978.
- 14. Letter from T. M. Anderson of Westinghouse Electric Corporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-2014, Dec.
11, 1978.
r l,
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CI F?GURE 4 g.
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WATER LEVELS
~
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~
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TINE {SEC) 8 8
FIGURE ll z.
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DECLG(CD = /.o )
CORE INLET VELOCITY
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PC TUBE PLUGGlNG CORRECT Sl FLO'VS NEV LOTlC V/PA1NT ACCUH.
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FIGURE 12 a.
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CORE FLOH (TOP AND BOTTOM)
DECLG(CD = oi6
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4.
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k 20.000 17.500
" AHERICAN ELECTRIC POMER COHPANY AHP UNIT 2'EB 1978 HOOEL 0.6 OECLG I
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DOMNCOMER AND'CORE MATER LEVELS 1
2.0000 1.7500 AMERICAN ELECTRIC POVER COMPANY AMP UNIT 2 FEB 1978 MOOEL 0.6 DECLG I
PC TUBE PLUGGING CORRECT SI FLO'VS NEV LOTIC V/PAINT FLOOO RATE(IN/SEC) 1.5000 1.2500 1.0000 Lal
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I.0000 AHERICAN ELECTRIC POllER COHPANY AHP 'UNIT 2 FEB 1978 HOOEL 0.6 OECLG 1 PC UBE PLUGGING CORRECT SI FLOVS NEV LOTIC V/PAINT POVER ~ 0.8000 0.6000 0.1000 0.2000 0.0 C3 Clo TIHE (SEC) CS Cl C3 FIGURE 15 c. CORE POWER TRANSIENT - DECLG(CD = y, g )
ATTACHMfNT B TO AEP:NRC:00322C
TABLE 1 TIME SEQUENCE OF EVENTS OCT'75 W/ZIRC FIX FEB'78 ANALYSIS REACTOR TRIP SIGNAL 0.72 0.71 SI SIGNAL 4.64 4.83, ACCUt4JLATOR INJECT SI INJECTS 14.5 29.64 14.6 29.83 END OF BYPASS 27.72 33.18 END OF BLOWDONN 32;04 39.07 BOTTOM OF CORE RECOVERY 44.48 48.97 ACCUMULATOR EMPTY 56.83 56.42
MASS DEFICIT IMPACT LIQUID REMAINING 8 EOB BYPASS DEFICIT LIQUID REMAINING FOR REFILL OCT'75 W/ZIRC FIX 25163.4 lbs. 28324.3 lbs. 0.0 FEB'78 ANALYSIS 35299.9 lbs. 30487.5 lbs. 4812.4 lbs. EOB ACCUMULATOR DELIVERY RATE AVAILABLEREFILL TIME DIFFERENCE EOB DIFFERENCE BOC DIFFERENCE 4600.0 lbs./sec. 1.05 sec. 5.46 sec. '.49 sec.
TABLE 2 RESULTS OCT'75 W/ZIRC FIX FEB'78 ANALYSIS FQT PCT 2.10 2188 F 2.02 2,171 F ZPCT 8.0 FT 7.5 FT ZR/H20 RXM Z ZRH20 RXH HOT ROD BURST 7.71% 8.0 FT 36.8 SEC 6.36K 7.5 FT 40.6 SEC Z HRBURST 6.0 FT 6.0 FT
0 TABLE 3 OCT'75 W/ZIRC FEB '78 ACCUMULATOR EMPTY TIME ZD 9 ACCUMULATOR EMPTY TIME ZC 9 ACCUMULATOR EMPTY TINE a STEAN COOLING. STARTS 56.83 SEC 15.99 FT 2.01 FT 86.63 SEC 56.42 SEC 10.'35 FT 1.55 FT 59.57 SEC Zp O'TEAM COOLING ZC 9 STEAM COOLING FULL DOWNCOMER TINE ACCUMULATOR WATER YOLUME PER TANK 9 BOC 16.0 FT 2.72 FT 63.6 SEC 377.67 FT 10.7 FT 1.715 FT 128 SEC '223.37 FT
4 ('fg b ~ ~ b J bb>> rb>> 2>>0000 I>>T500 l>>5000 AHERICAH ELECTRIC POVER COHPARY AHP UNIT 2 FEB ISTB HOOEL 0.8 OECLG I PC TUBE PLUGGIRG CORRECT Sl FLOVS REV LOTIC V/PAIHT FLOOO RATEIIM/SEC) b ~ j t>>2500 ocr rs u I 0000 R g O.T500 Ch Cl 0.5000 0>>2500 b ~ 0.0 CI O 8 8 ~bb TIHE {SEC) 8 8 bb fg>> I 5
(5
lS I 47, S ~ ~ S44, 80.000 IT>>500 AHERICAII ELECTRIC POVER COHPAllf AHP URIT 2 FEB IST~ HOOEL 048 OECLO I PC TUSE PLUCCIRO CORRECT Sl FLOVS REV LOTIC V/PAINT VATER LEVEL{FI) I5.000 I 44 I2.500 I0.000 oc7 78 145000 44S I 50000 845000 l I I 0.0 Cl CI 8 8 8 TIHE ISEC) 8 8 8 8n 4
- 444 4 ly 4 I
, MAJOR MODEL CHANGE IMPACTS A) 15 X 15 FLECHT NO IMPACT B) WATER PACKING AS NOTED EARLIER C) DYNAMIC STEAM COOLING 60'F [WCAP 9220 -P-A] D) 8= 1.25 DURING REFLOOD REFILL 5 REFLOOD BENEFIT E) CODE MAINTENANCE 5 ANALYTICAL IMPROVEMENT - LESS THAN -12'F
INPUT CHANGES IMPACT A) SATAN 1') CORE POWER IN SATAN* 2) STEAM GENERATOR INITIALIZATION/T-INLET" 3) INCREASED SAFETY INJECTION 4) MICELLANEOUS* B) REFLOOD 1) BROKEN LOOP ACCUMULATOR SPILL TO SPRAY** 2) INCREASED SAFETY INJECTION 3) LOWER PLENUM VOLUME REVISION C) LOTIC 1) PAINT ON CONTAINMENT SURFACES** O), LOCTA
- 1) AXIAL NODING SCHEME -
NO IMPACT
- 2) INITIAL ROD GAP GAS'BACKFILL PRESSURE
- LESS THAN 25'F IMPACT
- LESS THAN 3'F IMPACT}}