ML20002C532

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Small & Intermediate Break LOCA Analysis for Big Rock Point Reactor W/Automatic Depressurization Sys & Jersey Nuclear Co Reload G Fuel
ML20002C532
Person / Time
Site: Big Rock Point File:Consumers Energy icon.png
Issue date: 10/03/1972
From:
CONSUMERS ENERGY CO. (FORMERLY CONSUMERS POWER CO.)
To:
Shared Package
ML20002C527 List:
References
NUDOCS 8101100424
Download: ML20002C532 (12)
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Text

{{#Wiki_filter:... ~ h-: ^ p i 2-i 1 } SMALL AND INTERMEDIATE BREAK LOSS OF COOLANT ACCIDEfO ANALYSIS-l FOR THE BIG ROCK POINT REACTOR WITH ADS 1 AND i JNC RELOAD G FUEL. 3 i t i j October 3, 1972 i l ' I l i l l-l l t. 810 Il0D #1

J ' l I. INTRODUCTION This report presents the analysis of the consequences of small and intermediate break size loss of coolant accidents (LOCA) for Reload G fuel. It is intended as a supple:,cnt to the June 16, 1972 submittal which presented the analysis and results of the Design Basis Accident (DBA) (a 4.5 ft2 doubic cnded break in the recirculation piping below thereactorvessel). The models and assumptions used in the analysis are in full co.n-pliance with the AEC's Interim Acceptance Criteria (IAC). i i b-

(

1 2 l II.

SUMMARY

The consequences of a LOCA for Reload G fuel in the Big _ Rock Point Reactor have been calculated in accordance with the IAC. The analysis utilized-the heat transfer coefficients presented in our September 22, 1972 submittal, and clearly demonstrates that Reload G fuel complies fully with the limits set forth in the IAC. The results are presented below. LOCA ANALYSIS RESULTS_ F0R-Ji (ELOAD G FUEL Break-Size Peak Clad Core Average Metal (ft ) Temperature (*F) WaterReaction('Q 2 DBA 4.5 2286

0. 6' Intermediate 0.1

-2175 Less than 0.6 Small 0.008 1263 Less than 0.02 i 4 i e

i 3 III. METit0D OF ANALYSIS A. Course of Accident The loss of coolant accident is assumed to follow the same course as that described in our Septeaber 22, 1972 submittal. To summarize the assumptions briefly, they were: 1. Recirculatien flow and feed flow are lost at the time of the accident. The flow available through the feed regulating bypass valve was' assumed to come on 10 minutes after the onset of the accident. 2, Automatic Depressurization System (ADS) was operable (a33 lb/sec'c# saturated steam at 1350 psia). 3. Only one core spray system was operable. 4. Break flow from one of the vessels (pressure vessel or steam drum) is limited by friction. B. Decay lleat The decay heat curve used was that presented in the prop;;cd i.SS Standard, plus 20%, as specified by the IAC. Ninety-six percent (96%)ofthisvaluewasassumedtobegeneratedinthefuel. C. Heatup Calculations Rod heatup calculations were performed with the M0XY code, developed for the AEC'by Idaho Nuclear Corp. An emissivity of 0.67 was used for both the dry rods and channel, whereas the wetted -is and channel were assumed to have an emissivity of 0.96. D. Wetting Times Wetting times were calculated using the Yamanouchi analysis. Calculations for Reload G fuel indicated a channel wettina time of 293 seconds for the 0.1 ft2 break, based on dryout at 70 seconds and rated spray at 190 seconds. For added conservatism, however, a time of 314 seconds was used--this is the same as was 1 r

4 calculated for NFS. fuel-(September _ 22, 1972 submittal). The center rod was calculated to wet at 401 seconds. For the 0.008:ft2 break, where dryout occurred at 567 seconds, core recovery was calculated to. take place at-740 seconds. E. Heat Trarisfer Coefficients The heat-transfer coefficients used are the same as those' presented .in our September 22, 1972 submittal. The curves of-heat transfer coefficients vs. time for the intermediate and small breaks are displayed in Figures 1 and 3. F. Metal Water Reaction The metal water reaction was calculated using the Baker-Just equation with.a coefficient of 1.0. G. Peaking Factors The radial-and axial peaking factors used were 1.45 and 1.51, respectively. The local peaking' factors used represent those at an assembly exposure of 11,000 MWD /MT, and are shown in the June 16, 1972 submittal. t I 1 h I l i I I Y

5 2 IV. IrlTERMEDI ATE '(0.1 f t ) BREAK Figures 1 and 2 show the heat transfer coefficients and cladding temperature variations with time. The peak clad temperature was conservatively calculated to be 2175"F at 540 seconds. The maximum local zirconium oxide thickness is.0035 inch, corresponding to a maximur. of about 10% of the 0.034 inch thick cladding. ' The core average metal-water reaction is slightly less than 0.6%. The significant events and 2 results for the 0.1 ft break are preso..H below.

SUMMARY

OF EVENTS - 0.1 FT2 BREAK Heat Transfer Cogfficients Time (sec.) Event (BTU /hr - ft' "F) 0 Break Occurs Channel 10,000 10,000 All Rods = 0 70 Dryout Channel = All' Rods 0 = 190 Rated Spray Dry Channel = 20 1.5, 1.7. 2.0, 3.2 Dry Rods = 340 Channel Wets Wetted Channt.1 = 1,000

1. 5, 1.7, 2.0, 3.2 Dry Rods

= 4 01 Center Rod Wets Wetted Channel = 1,000 Wetted Rod 1,000 Dry Rods = 1. 5, 1.7, 2.0, 3.2 540 Peak Clad Temp. i r

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8 2 V. - SMALL (0.008 f t ) BREAK Figures 3 and 4 show the heat transfer coefficients and cladding temperature variations with time for the small break. The peak clad temperature was conservatively calculated to be 1263 F at 740 seconds. The maximum local ~ zirconium oxide thickness is.00001 irch, corresponding to a maximum of about 0.03% of the 0.034 inch cladding. The core ave.cge metal-water reaction is less than 0.02%.

SUMMARY

OF EVENTS .008 FT2 BREAK 1 Heat Transfer Coefficients

  • Time (sec.)

Event (BTU /hr - ft2. 'F) 0 Break Occurs 10,000 567 Dryout 0 740 Reflood 25 740 Peak Clad Temperature s

  • All rods and channel.

P00R ORIGINAL

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~ ^ ( i 11 VI. CONCLUSIONS The consequences of the full spectrum of break sizes ~ of LOCA have been analyzed for JilC Reload G fuel in the Big Rock Point reactor with'an ADS. For the worst case (DBA, 4.5 ft2 break),boththepeak clad temperature of 2286 F and the core average metal-water reaction of 0.6% fall below the AEC limits of 2300 F and 1.0%, respectively. Thus, it is concluded that JNC Reload-G fuel in the Big Rock Point reactor complies fully with the AEC's Interim Acceptance Criteria. 9 + 1 I r ~ , - -,}}

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