ML19260C454
| ML19260C454 | |
| Person / Time | |
|---|---|
| Issue date: | 07/03/1979 |
| From: | Parczewski K Office of Nuclear Reactor Regulation |
| To: | Almeter F, Strosnider J Office of Nuclear Reactor Regulation |
| References | |
| REF-GTECI-A-03, REF-GTECI-A-04, REF-GTECI-A-05, REF-GTECI-SG, TASK-A-03, TASK-A-04, TASK-A-05, TASK-A-3, TASK-A-4, TASK-A-5, TASK-OR NUDOCS 8001030488 | |
| Download: ML19260C454 (5) | |
Text
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j JUL 0 31979 MEMORANDUM FOR: Jack Strosnider, Manager, Tasks A-3, A-5 Frank Almeter, Manager, Task A-4 FROM:
Kris I. Parczewski, Reactor Safety Branch, DDR
SUBJECT:
PROPOSED STUDY OF THE EFFECT OF STEAM GENERATOR TUBE RUPTURE ON PEAK CLADDING TEMPERATURE AFTER A LARGE BREAK LOCA Introduction The rupture of steam generator tubes during a large break LOCA can have a significant effect on the peak cladding temperature reached during the accident.
The secondary water entering the primary coolant system interferes with the flow of safety injection water and affects the nature of f.uel quench process which may result, under certain circumstances, in considerably higher peak cladding temperatures reached after a LOCA. This phenomenon has been investigated experi-mentally and analytically. The experimental study was performed in the semi-scale test facility at INEL (Reference 1).
It was found that for the secondary to primary leakage rates corresponding to less than 10 single ended ruptures flow area) tne peak of steam genertor tubes (scaled to the semiscale cort cladding temperature remained practically unchanged. For the rates corre-sponding to a larger number of failed tubes, the steam generated by the incoming secondary fluid slowed down reflood rates and produced an increase in peak cladding temperature. At the leakage rate corresponding to 16 failed tubes the fluid in the core became completely stagnant and the peak cladding temperature increased by about 660aF. For the higher flow rates the secondary water produced Similar top fuel quench which resulted in lower peak cladding temperatures.
results were obtained from the analytical study which consisted of computer simulation (RELAP/4 MOD 5 and FLOOD 4 Codes) of a typical Westinghouse 4-loop plant (Zion 1).
In this study the maximum peak cladding temperature corresponding 20 failed tubes (2600 gpm), was well above 2200 F.
Sensitivity studies were performed in order to establish the worst location of tube rupture and the worst time in LOCA when this break would occur.
It was found that the largest change in peak cladding temperature was obtained when the ruptured tubes were located near the inlet plenum of the intact loop steam generator and the rupture occurred at the beginning of the reflood period.
Goals of the Study It is proposed to use the peak cladding temperture (PCT) reached during a large break LOCA as a criterion for evaluating the maximum pemissible number of steam genertor tubes failed during the accident. The limiting value of PCT, 1700 003 800/0304 F7
Jack Strosnider JUL 0 3199 Frank Almeter specified in 10 CFR 50.46, is 2200*F and according to proposed criterion, the secondary to primary leak rates due to failed tubes should not cause the PCT to exceed this value. The goal of the present study is to develop a pro-cedure by which the leak rates corresponding to the limiting value of PCT could be detemined. Since the PCT reached during a LOCA is plant specific, the exact values can be obtained only by analyzing individual plants. How-ever, as a first approximation, it can be assumed that the effect of steam tube leakage is similar in similar plants. The maximum PCT reached could be therefore obtained from the PCT corresponding to zero leakage and by apply-ing th leakage corrections derived from the generic study perfonned for sini11ar type of plants. The proposed program hinges upon this assumption.
It is proposed to generically determine the effect of steam genertor tube rupture on PCT in five different reactors representing five different types of operating plants and to use this information in developing the required leakage corrections.
The exact methodology for applying the results of generic study to specific plants has not been yet developed, but probably it would consist of generating normalized PCT vs leak rate curves which could then be applied to the plant specific data available to the NRC, Proposed Program The proposed program consists of computer simulation of the effect of steam generator tubes rupture on PCT in the following plants:
Westinghouse 2 loop plant Westinghouse 3 loop plant CE plant B&W plant The study of Westinghouse 4 loon plant was performed previously (Reference 1) and the results are available. However, some additional work may be needed in order to generate the normalized PCT vs leak rate curve.
The analysis will be performed using two computer codes:
RELAP/4 MOD 5 or MOD 6 for simulating blowdown and refill phases and FLOOD 4 for simulating core reflood phase. The study will include the following assumptions:
(a) The ruptured steam generator tubes will be located near the inlet plenum of the intact loop steam generator.
(b) The rupture will be assuinad to occur at the beginning of core reflood period.
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Jack Strosnider R 0 3 7379 Frank Almeter (c) Constant secondary to primary leak rate will be assumed which will be determined by a chocked flow regime.
(d) The assumed LOCA will correspond to cold leg duuble ended guillotine
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break with the discharge coefficient of 1.0.
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The range of leak rates investigated will extend from zero to about 2600 gpm,\\{# p'r 6F.( d, corresponding to about 20 failed tubes. However, this range would be modi-, /y, k f
d fied if it is found that the higher leak rates are required in order to
'ed."./jh bracket the 2200 F PCT limit.
4t C_onclusion
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t The proposed study would provide the means for detennining the secondary to Sc;Pg primary leakage rates at which PCT reaches 2200*F. However, it should be realized, that because of certain limitations inherent in the study, the results should be considered as best estimates rather than the rigorously calculatd values.
Kris 1. Parczewski
~ Reactor Safety Branch Division of Operating Reactors cc:
S. Hanauer M. Aycock.
References 1.
TREE-NUREG-1213. " Investigation of Influence of Simulated Steam Generator Tube Rugtures during Loss-of-Coolant Experiments in the Semiscale M00-1 System, May 1978.
2.
CAAP-TR-78-032 " Steam Genertor Tube Rupture Effects on a LOCA,"
November 1978.
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