ML19320A772
| ML19320A772 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Calvert Cliffs, Palisades, Fort Calhoun  |
| Issue date: | 06/30/1980 |
| From: | ABB COMBUSTION ENGINEERING NUCLEAR FUEL (FORMERLY, EMVC-EPS |
| To: | |
| Shared Package | |
| ML19320A739 | List: |
| References | |
| TASK-06-02.B, TASK-6-2.B, TASK-RR NUDOCS 8007020600 | |
| Download: ML19320A772 (9) | |
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APPENDIX E
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l ASYMMETRIC LOADS EVALUATION PROGRAM c
E FINAL REPORT
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CALVERT CLIFFS 1&2 FORT CALHOUN MILLSTONE 2 E
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COMBUSTION ENGINEERING, INC-800702.o @
APPEllDIX E I
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1 REACTOR C00LNiT SYSTEM i
i ASYte!ETRIC LOADS FIflAL REPORT I
ECCS NlALYSIS APPROACH WITH REDUCED AREA C00LAflT CHNiilELS Ifl PERIPHERAL ASSEf4BLIES Prepared by COMBUSTI0fl ENGirlEERIflG, IflC.
for CALVERT CLIFFS 1 & 2 FORT CALHOUN MILLSTONE 2 June 30, 1980
TABLE OF C0?lTErlTS SECTI0ft SUBJECT PAGE flUMBER E1.0 IllTRODUCTI0tl E-1 i
E2.0 METil0D OF AtlALYSIS E-1 E2.1 Blowdown ilydraulics E-2 E2.2 Refill /Reflood Hydraulics E-3 E2.3 Temperature Analysis E-3 E3.0 REFEREllCES E-4 I
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E.
1.0 INTRODUCTION
AND
SUMMARY
The ECCS perfonnance evaluations demonstrating confonnance with 10CFR50.46, which presents the NRC Acceptance Criteria for Emergency Core Cooling Systems for Light Water Cooled Reactors (I), are presented in References 2 through 5.
These references provide analyses for Calvert Cliffs Units 1&2, Millstone 2, and Ft. Calhoun.
The purpose of this supplementary analysis is to demonstrate acceptable ECCS perfonnance with reduced area coolant channels assumed in the peripheral fuel assemblies.
While demonstrating acceptable ECCS performance, the intent of this analysis is to also show that the current licensing analysis, pertaining to the hottest fuel rod in the core, is more limiting than that for the hottest rod in a peripheral assembly with reduced area coolant channels.
Since this evaluation is to apply to the above plants, a generic analysis was perfonned.
The method of analysis is discussed in the following sections.
E.2.0 METHOD 0F ANALYSIS In the C-E ECCS evaluation model(6,7), the CEFLASH-4A(8) computer program is used to detennine the primary system thennal hydraulic behavior during the blowdown period, and the COMPERC-II(9) program is used to describe the system behavior during the refill and reflood periods. The esulting transient parameters from these computer programs, describing the thermal and hydraulic behavioroftheprimarysystem,supplytheinputtotheSTRIrsIN-Idl0) program which is used to calculate the hot rod peak clad tt,nperature and peak local clad oxidation percentage.
The objective of the analysis is to demonstrate that the ECCS performance for a peripheral assembly with reduced area coolant channels is less limiting than a hot rod in a channel without any reduction in flow area. To accomplish this objective it is necessary to evaluate the performance of the limiting fuel rod E-1
in the peripheral assembly containing reduced area fuel channels.
In evaluating the performance of the liraiting fuel rod in the peripheral assembly, blowdown refill /reflood, and temperature calculations were performed using the computer programs described above based on a conservative set of input assumptions.
The conservative assumptions are employed in the analysis so that the results will conservatively bound the response for the Calvert Cliffs Units 1&2, Millstone 2, and Ft. Calhoun plants.
The details of these assumptions and the analytical methods employed in this analysis are discussed in the subsections below.
I E.2.1 Blowdown Hydraulics The blowdown portion of the transient was analyzed using the CEFLASH-4A computer program.
In the CEFLASH-4A calculation, the peripheral assembly was explicitly represented with a 10% reduction in total assembly cross sectional flow area. This reduction in peripheral assembly flow area conservatively exceeds the maximum expected defonnation since the testing program identified this maximum blockage to be 9%.
This deformation was also assumed to occur along the entire length of the assembly to minimize the flow in this region.
In addition, the power level of the peripheral assembly was conservatively assumed to be at the core average power level.
This assumption is conservative since the peripheral assemblies are approximately 5% to 10%
lower than that for the core average which results in maximizing the heat addi-tion to this region.
I In performing the blowdown calculation, the Calvert Cliffs plant, a re-presentative 2700 Mwt class NSSS, is used.
This plant was chosen since its' core power level is highest of all the plants considered in this evaluation.
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.2. 2 Rcfill/Reflood Hydraulics Since the containment pressure and core average reflood rates are unaffected by the flow area reduction in a single peripheral assembly, no new COMPERC-II calculations were r ssa ry. As a consequence, the COMPERC-II refill /reflood hydraulics calculations from a representative 2700 Mwt class NSSS, presented in Reference 10, was chosen for use in this portion of the evaluation.
This particular analysis was chosen since the evaluation resulted in the lowest containment pressure, the lowest reflood rate, and hence the lowest reflood heat transfer coefficients, for the plants considered in this report.
E.2. 3 Temperature Analysis The STRIKIN-II and PARCH computer programs were used to evaluate the temperature transient and peak local clad oxidation percentage for the hottest rod in the peripheral assembly.
For conservatism, in modeling rod-to-rod thermal radiation, the power distri-bution surrounding the hot rod in the peripheral assembly was assumed to be a relatively flat distribution. As a consequence, the rods surrounding the hot rod in the peripheral assembly will be very nearly the same temperature as the hot rod during the entire transient thereby minimizing the benefits frcm rod-to-rod thennal radiation.
I In evaluating the response of the hottest rod in the peripheral assembly, the channel surrounding this rod was assumed to be reduced in flow area with percentage reductions in the range from 0 to 35%.
The maximum reduction in single channel flow area of 35% is conservative since it exceeds the maximum expected E-3
flow area reduction of 34% obtained from the testing program.
To evaluate the perfonnance of the hottest peripheral assembly fuel rod, temperature calculations using the STRIKlil-Il code was perfonned at various channel flow areas with percentage reductions up to 35%.
The results will be presented as a curve of allowable linear heat rate, for a peripheral assembly, as a function of percent reduction in single channel flow anca for the hottest pin in this assembly.
Figure E-1 illustrates the expected relationship between linear heat rate or kw/ft limit and percent channel flow area reduction for the peripheral assembly.
E.3.0.
REFERE?'CES 1.
Acceptance Criteria for Emergency Core Cooling Systems for Light-Water Cold fluclear Power Reactors, Federal Register, Vol. 39, flo. 3 -
Friday, January 4,1974.
2.
CG&E Letter, J. W. Gore to L. M. Muntzing, dated 9/12/74.
3.
BG&E Letter, A. E. Lundwall Jr. to A. Giambusso, dated 7/17/71 4.
ECCS letter, to Mr. R. Reid of flRC from W. G. Counsil, dated March 30, 1979,
" Millstone fluclear Power Station Unit 2 Large Break LOCA ECCS Performance Results".
5.
OPPD letter, W. Jones to R. Reid, dated February 12, 1980.
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6.
CENPD-132, " Calculative Methods for the CE Large Break LOCA Evaluation Model", August 1974 (Proprietary).
CEf4PD-132, Supplement 1, " Updated Calculative Methods for the CE Large Break LOCA Evaluation Model", December 1974 (Proprietary).
7.
CENPD-132, Supplement 2, " Calculational Methods for the CE Large Creak LOCA Evaluation Model", July 1975 (Proprietary).
8.
CENPD-133, "CEFLASH-4A, A FORTRAN IV Digital Computer Program for Reactor Blowdown Analysis", April 1974 (Proprietary).
CENPD-133, Supplement 2, "CEFLASH-4A, A FORTRAN IV Digital Computer Program for Reactor Blowdown Analysis (Modification)", December 1974 (Proprie tary).
9.
CENPD-134, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core", April 1974 (Proprietary).
CENPD-134, Supplement 1, "COMPERC-II, A Program for Emergency Refill-Reflood of the Core (Modification)", December 1974 (Proprietary).
- 10. CENPD-135, "STRIKIN, A Cylindrical Geometry Fuel Rod Heat Transfer Program, April 1974 (Proprietary).
CENPD-135, Supplement 2, "STRIKIN-II, A Cylindrical Geometry Fuel Rod Heat Transfer Program (Modification)", February 1975.
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l up I FIGURE E-1 EXPECTED RESULTS
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C00LAlilLITY STUDY
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<g TECH SPEC /LC0 l-g.
D PERIPHERAL ASSEMBLY 52
-PCT A 2200 F
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CLAD OXID 6 17%
g m
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I MAXIMU.'4 FOR HOTTEST l
FUEL.RODINPERIPHERBL ASSEMBLY-l l
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l UAXIMUn' CHANNEL
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AREA REDUCTION l
g 34 il0T CHANrlEL AREA REDUCTIO;l, %
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