ML19256F058
| ML19256F058 | |
| Person / Time | |
|---|---|
| Site: | Point Beach |
| Issue date: | 11/19/1979 |
| From: | Burstein S WISCONSIN ELECTRIC POWER CO. |
| To: | Harold Denton, Schwencer A Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 7911210050 | |
| Download: ML19256F058 (34) | |
Text
{{#Wiki_filter:Wisconsin Electnc era courar 231 W. MICHIGAN, P.O. BOX 2046, MILWAUKEE, WI 53201 November 19, 1979 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. NUCLEAR REGULATORY COMMISSION Washington, D. C. 20555 Attention: Mr. A. Schwencer, Chief Operating Reactors Branch 1 Gentlemen: DOCKET NO. 50-266 ECCS REANALYSIS FOR 18% STEAM GENERATOR TUBE PLUGGING LIMIT POINT BEACH NUCLEAR PLANT, UNIT 1 As a result of inservice inspection of the steam generator U-tubes during the current Point Beach Nuclear Plant Unit 1 refueling outage, additional tubes in both the A and B steam generators have been plugged and taken out of service. In the A steam generator a total of 77 tubes have been plugged during this outage and in the B steam generator 68 were plugged. These recent tube pluggings have resulted in a net total of 329 tubes plugged in the A steam generator, or 10.1% of the 3,260 U-tubes, and 319 tubes or 9.8% in the B steam generator. Ac you know, the present ECCS LOCA analysis for the Point Beach 11uclear Plant assumes up to 10% of the steam generator tubes are pit.gged. Although recent plugging is essentially at this nominal 10% level, we requested that the Westinghouse Electric Corporation complete an ECCS reanalysis using an assumption of 18% of the steam generator tubes plugged. This reanalysis was done for the specific limiting break for the Point Beach Nuclear Plant. A copy of the results of this reanalysis was mailed to Mr. Trammell of your office on November 14 and another copy is enclosed. The Point Beach Nuclear Plant utilizes upper plenum injection, and a 60'F increase in temperature should be added to the calculated peak clad temperature to account for explicit modelling of upper plenum injection. Westinghouse has determined for the Point Beach Nuclear Plant that fuel rod bursting and blockage is appropriately accounted for in the ECCS analysis results. The analysi,s was performed at 2250 psia primary system 13/S ?97 7911210 O ~1-
Mr. Harold R. Denton November 19, 1979 pressure and normal operating conditions. Initial operating conditions on return of the unit to service may be at 2000 psia and at a reduced core inlet temperature. Westinghouse ECCS sensitivity studies have shown that the reduced pressure will have an insignificant effect on analysis results and that the reduced core inlet temperature and resulting reduced power level result in lower peak clad temperature than would be expected under normal operating conditions indicated in the attached analysis. These results demonstrate that the Point Beach Nuclear Plant Emergency Core Cooling System continues to meet the Acceptance Criteria, as presented in 10 CFR Part 50.46 of the Commission's Regulations. Reactor coolant system flow will be maintained above the flow rate used in the thermal-hydraulic evaluation of the plant, even if 18% of the steam generator tubes were to be plugged. We have discussed this reanalysis with members of your Staff and have been advised that the Commission believes a licensing action on this evaluation is necessary. We do not agree. Under the provisions of 10 CFR Section 50.59, licensees are authorized to make changes in the facility without prior Commission approval, provided the changes do not involve a change in the Technical Specifications or an unreviewed safety question. The Manager's Supervisory Staff has conducted a safety evaluation of this reanalysis and has determined that this change does not involve an unreviewed safety question. However, in order to facilitate the prompt return to power of Unit 1 following this refueling outage, we are making this transmittal so that the NRC may conduct an independent review of this reanalysis. If this review is deemed by you to constitute licensing action resulting in an approval required by the Commission's regulations, we would appreciate your advice as to whether an approval fee is required. If such a fee is required, we will, of course, make an immediate remittance as required by 10 CFR 170.22. We are presently anticipating completion of the present Point Beach Unit 1 refueling outage during the week of November 19, 1979. We, therefore, request that the Staff give this matter its highest priority. Should you have any questions regarding this matter, please contact us at once. We will make every effort to facilitate your prompt review of this material. Very truly yours, Sol Burstein Exedutive Vice President Enclosure I375 ?98
The Loss of Coolant Accident (LOCA) nas been reanalyzed for point Geach Units 1 and 2. The foilowing in forma tion' amends the Sa fety Analysis Report section on Major Reactor Coolant System Pipe Ruptures. The re-sults are consistent with acceptance criteria provided in re ference (1). The description of the various aspects of the LOCA analysis is given in WCAp-8339.[2] The individual computer codes which comprise che Westinghouse Emergency Cora Cooling System (ECCS) evaluation model are described in detail in separate reportsb 3-b along with code mooifi-cations speci fied in re ferences 7, 9 and 10. The analysis presented here was per formed with the February 1978 version of the evaluation model which includes modifications delineated in re ferences 11, 12, 13 and 14. Resul ts The analysis of the loss of coolant accident is per formed at 102 percent o f the licerised core power rating. The peak linear power and total core power used in the analysis are given in Table 2. Since tnere is margin between the. value of peak linear power density used in this analysis and the valse of 15e peak linear power density expected during plant opera-tion, the peax clad temperature calculated in this analysis is greater than the maximum clad temperature expected to exist. Table I presents the occurrence time for various events throughout the ac:Ident transien. 157S ?99 e-O s e,
.e..-- Table 2 presents selected input values and results from the hot fuel rod ~ thermal transient calculation. For these results, the hot spot is de-fined as the location of maximum peak clad temperatures. That location is speci fied in Table 2 for each break analyzed. The location is indi-cated in feet which presents elevation above the bottom of the active fuel stack. Table 3 presents a sumary of the various containment systems parameters and structural parameters which were used ' input to the C0C0 computer code [6] used in this analysis. 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 sensitivity studies are reported.E03 These results are for conditions which are not limiting in nature and hence are reported on a genAric basis. Figures 1 through 17 present the transients for the principal parameters for the break sizes analyzed. The following. items are noted: by Figures 1 3: Quality, mass velocity and clad heat transfer i, coefficient for the hotspot and burst locations Figures 4 6: Core pressure, break flod, and core pressure drop. The break flow is the sum o f the flowratet from both y 1375 300 t .........,...-.....__m.... ...,.. ~
ends of the guillotine break. The core pressure drop is taken as the pressure just be fore the core inlet to the pressure just beyond the core outlet 9: Clad temperature, fluid temperature and core flow. Figures 7 The clad and fluid temperatures are for the hot spot and burst locations Figures 10 - 11: Downcomer and core water level during reflood, and flooding rate Figures 12 - 13: Emergency core' cooling system flowrates, for both accumulator and pumped sa fety injection Figures 14 - 15: Containment pressure and core power transients Figures 16 - 17: Break energy release during blowdown and the con-tainment walI condensing heat trans fer coef ficient for the worst break. E; 13/5 301 t.
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Conclusions - Thermal Analysis For breaks up to and including the double ended severance of a reactcr coolant pipe, the Emergency Core Cooling System will meet the Acceptance Criterie as presented in 10CFR50.46.EI3 That is: 1. The calculated peak clad temperature does not exceed 2200*F based on a total core peaking factor of 2.32. 2. The amount of fuel element cladding that reacts chemically with water or steam does not exceed 1 percent o f the total amount o f 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 17 percent are not. exceeded during or after quenching. 4. The core temperature is reduced and decay heat is removed for an exte, ied period of time, as required by the long-lived radioactivity remaining in the core. A The 1 f tacts of upper plenum injection for Westinghouse-designed 2-loop plants has been_ discussed with the sta f f.[15,16,17,18,19] Based on intarim calculations, a 60*F increase in calculated peak clad temper-atures results from explicit modelling of upper plenum injection in the Point Beach power plant. In order to use the present Westinghouse ECCS evaluation model[13,14,15] to analyze a postulated LOCA in the G 13/$ 502
Point Beach plants and remain in compliance with 10CFR50.46, a limit o f 2140*F on calculated peak clad temperatures must be observed. It can be seen from the results contained herein that this ECCS analysis for the Point Beach power plants remains in compliance with 10CFR50.46. 1.5/5 503 e e e e O y a. ~ e d a, f * *. ** 't e 4,0 4f 3 * .; s p!.. e G 0 O e
8. Salvatori, R., " Westinghouse ECCS - Plant Sensitivity Studies," WCAP-8340 (Proprietary Vers' ion), WCAP-8356 (Non-Proprietary Ver-sion), July 1974. 9. " Westinghouse ECCS Evaluation Model, October,1975 versions," WCAP-8622 (Proprietary Version), WCAP-8623 (Non-Proprietary Ver-sion), November,1975. 10. Letter from C. Eicheidinger of Westinghouse Electric Corporation to D. B. 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 Laop Out of Service for. Plants Without Loop Isolation Valves," WCAP-9166, February,1978. 12. Eicheidinger, C., "Westin9 house ECCS Evaluation Model, February 1978 Version," WCAP-9220-P-A (Proprietary Version), WCAP-9221-A (Non-Proprietary Vers ton), February,1978. 13. Letter from T. M. Anderson ofsWestinghouse Electric borporation to John Stolz of the Nuclear Regulatory Commission, letter number NS-TMA-1981, Nov.1,1978, - .4 4, a 14. Letter from T. M. Anderson';of Westinghouse Electric Corporation to R. L. Tedesco o f the Ituc1 ear Regulatory Commission, letter number NS-THA-2014, December kl,1978. i 13/5 504 v- ,s .ges..fy,we;wymw _. e. i rTee. 7mvf urre,ppenyg, e nt-ma -s, e a
Re ferences for Section 15.4.1 1. "Acceptanc'e Criteria for Emergency Core Cooling Systems for Light Water Cooled Nuclear Power Reactors," 10CFR50.46 and Appendix K o f 10CFR50.46. Federal Register, Volume 39, Number 3, January 4, 1974 2. Bordelon, F. M., Massie, H. W., and Zordan, T. A., "Wes ti nghouse ECCS Evaluation Model-Summary," WCAP-8339, July 1974. 3. Bordelon, F. M., et al., " SATAN-VI Program: Comprehensive Space-Time Dependent Analysis o f Loss-o f-Coolant," WCAP-8302 (Prop-rietary Version), June 1974. 4. Bordelon, F. M., et al., "LOCTA-IV Program: Loss-o f-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 Re flooding Af ter a Loss-of-Coolant Accident (WREFLOOD Code)." WCAP-8170 (Proprietary Version), WCAP-8171 (Non-Proprietary Version), June 1974. 6. Bordelon, F. M., and Murphy, E. T., " Containment Pressure Analysis Code (C0CO)," WCAP-8327 (Proprietary Version), WCAP-8326 (Non-Proprietary Version), June 1974.. 7. Bordelon, F. M., et al., "The Westinghouse ECCS Evaluation Model: Supplementary 'In formation," WCAP-8471 (Proprietary Version), WCAP-8472 (Non-Proprietary Version), January 1975. 13/5 505
15. " Safety Evaluation Report on ECCS Evaluation Modef for Wes tinghouse Two-Leop Plants," Novemoer,1977. 16.* Le t ter from V. J. Esposito to H. W. Gutzman o f NSD Operations Sup-port, both from Westinghouse Electric Corporation, letter number SE-SAI-2267, January 30, 1978. 17. "NRC Questions Regarding TAC 1/16/78 Submittal by Westinghouse Designed Two-Loop Plant Operators," February 1,1978. 18.* Letter from V. J. Esposito to H. W. Gutzman of NSD Operations Sup-port, both from Westinghouse Electric Corporation, letter number >no SE-SAI-2290, February 17, 1972.. ('O 19. " Safety Evaluation Report on Interim ECCS Evaluation Model for Westinghouse Two-Loop Plants," March 1978.
- Wisconsin Electric Power should supply the proper references by which these re ferences were formally transmitted to the NRC.
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TABLE 1 LARGE BREAK - TIME SE00EtiCE OF EVEtiTS Occurrence Time (Seconds ) Even t DECLG, CD = 0. 4 Acciden't Initiation 0.0 Reactor Trip Signal .66 Sa fe,ty Injection Signal .8 Start Accumulator Injection 10.1 End of ECC Bypass 22.0 End o f Blowdown 22.0 Bottom of Core Recovery 42.4 Accumulators Empty 57.6 Start Pumped ECC Injection 25.8 e e O t 4 '- 13/5 507
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TABLE 2 ( ) LARGE BREAr, - ANALYSIS INPUT AND RESULTS Quantities in the Calculations: Licensed core power rating 102 percent o f 1518.5 MWt Total core peaking factor 2.32 Peak linear power 102 percent o f 13.23 kw/ ft , Accumulator water volume 1100 cubic feet per tank Accumulator pressure 700 psia Number o f sa fety. injection pumps operating 2 Steam generator tube plugging level 18 percent (uni form) Fuel parameters - Cycle Generic Region Generic Results DECLG, CD = 0.4 Peak clad temperature (*F) 2053 Location (feet) 7.5 Maximum local clad / water reaction (percent) 5.11 Location (feet) 7.5 Total core clad / water reaction (percent) <0.3 ~ Hot rod burst time (seconds) 29.8 Location ( feet) 5.75 is ^_, 3 y i. IJ/f]- 3(l8 v. ? n' 0 e 4 we
TABLE 3 C0r TAlfdEf4T DATA (DRY C0tiTAlfNENT) 6 3 tiet Free Volume 1.065 x 10 ft Initial Conditions Pressure 14.7 psia Temperature 90 *F .RWST Temperature 34 *F Service Water Temperature 33 *F Outside Temperature -25 *F Spray System ibmber of Pumps Operating 2 Runout Flow Rate 1950 gpm each Actuation Time 10 secs Sa feguards Fan Coolers Number of Fan Coolers Operating 4 Fastest Post Accident Initiation of Fan 35 secs Coolers 8 ,t..
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TnBLE 3 (Cont) STRUCTURAL HEAT Sit;K OATA 2 Thickness (in) Material Area ft .322 Steel /concre te,12 56020 .188 Steel / concrete,12 6230 .25 Steel / concrete,12 2480' .188 Steel / con crete,12 690 .0.14 Steel 103724 Steel 11710 .304 .443 Steel 4730 .584 Steel-5441 .712 Steel 4490 1.0 Steel 957 2.634 Steel 3667 .125 Steel 10221 .209 Steel 16551 .5 Steel 2707 .322 Steel' 13835 .065 Steel 141106 .036 Steel 38250 Steel / concrete,3 19500 3.0 Concrete 19500 30.0 Concrete 61500 0 9'!.' 15/5 310 . Y- < e, g s' ?N
TABLE 3 (Cont) PAINTED STRUCTURAL HEAT Sif.K OATA Structural Heat Sink Structural Heat Sink Paint Thickness 2 Sur face Area (Ft ) Thickness (In) (Mils) 56020 .322 7.5 2480 .25 7.5 i 103724 .094 7.5 11710 .304 7.5 4730 .443 7.5 5441 .5 84 7.5 4490 .712 7.5 957 1.0 7.5 3667 2.634 6.0 10221 .125 6.0 16551 .209 6.0 2707 .5 6.0 13/5 $11 O e e O
TABLE 4 REFLOOD MASS AND ENERGY RELEASE Tjme (Sec) Mass Flow (Lb/Sec) Energy Flow (Lb/Sec) 43 .013 16.4 44.7 .38 496.5 49.2 28.44 36925.0 60.1 49.84 64718.1 73.5 52.27 66428.7 89.6 189.91 96555.0 107.5 210.67 97611.9 127.1 215.01 94879.6 171.8 222.55 88109.1 225.8 230.78 80803.1 299.0 243.36 73133.1 lj/5 $12 e 4
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