ML18030B223
| ML18030B223 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 09/30/1980 |
| From: | GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML18030B222 | List: |
| References | |
| 80NED272, NEDO-24266, NUDOCS 8604080233 | |
| Download: ML18030B223 (19) | |
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ENCLOSURE /'EDO-24266 80NED272 Class I September 1980 DC POWER SOURCE FAILURE V
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,yi FOR I IL BROWNS FERRY NUCLEAR POWER STATION UNITS 1, 2 AND 3 8b04080233 8b032b PDR ADDCK 05000259 P
PDR NUCLEAR POWER SYSTEMS DIVISION ~ GENERAL ELECTRIC COMPANY SAN JOSE, CALIFORNIA95125 GENERAL 'LECTRlC 4
DISCLAIMEROF RESPONSIBILITY This document was prepared by or for the General Electric Company. Neither the General Electric Company nor any of the contnbutors to this document:
'I A.
Makes any warranty or representation, express or implied, with respect to the accuracy, completeness, or usefulness ofthe information contained in this docu-ment. or that the use of any information disclosed in this document may not infringe privately owned rights; or B.
Assumes any responsibility for liabilityor damage of any kind which may result from the use of any information disclosed in this document.
NEDO-24266 CONTENTS 1.
INTRODUCTION 2.
CONCLUSIONS
~Pa e l-l 2-1 3.
BREAK ANALYSIS 3.1 Discharge Breaks 3.2 Feedwater Line Breaks 3-1 3-1 3-2 4.
REFERENCES 4-1
NEDO-24266 Tab le TABLES Title
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3-1 3-2 Calculated PCT Results for a Recirculation Discharge Line Break With a DC Power Source Failure Calculated PCT Results for a Feedwater Line Break With a DC Power Source Failure 3-3 3-3
~PX ure 3-1 3-2a 3-2b 3-3 3-4 ILLUSTRATIONS Title Browns Ferry 1, 2 and 3 0.3 ft Recirculation Discharge Break DC Power Source Failure (No ADS)
Browns Ferry 1, 2 and 3 0.3 ft2 Recirculation Discharge Break DC Power Source Failure (No ADS) (8x8 Fuel Type)
Browns Ferry 1, 2 and 3 0.3 ft Recirculation Discharge 2
Break DC Power Source Failure (No ADS) (7x7 Fuel Type)
Browns Ferry 1, 2 and 3 0:25 ft2 Feedwater Line Break DC Power Source Failure Manual ADS (4 Valves) at 10 Minutes Browns Ferry 1, 2 and 3 0.25 ft Feedwater Line Break 2
DC Power Source Failure Manual ADS (4 Valves) at 10 Minutes r
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NEDO-24266 1.
INTRODUCTION The purpose of this study is to investigate, in a generic
- manner, the implications of a direct current (DC) power source failure for Browns Ferry Nuclear Plant Units 1, 2, and 3.
In addition, this study provides bounding peak cladding temperatures (PCT) as a function of break area for small breaks.
ECCS analyses were performed for the most severe break locations (the recirculation discharge line where LPCI injects and the feedwater line where HPCI injects),
assuming that the DC power failure also disables the automatic depressurization system (ADS).
The analyses were performed with the 1977 approved GE LOCA analysis models
- SAFE03, REFLOOD05 and
- CHASTE05, and considering the LPCI modification as described in Reference l.
This study is based on system availability following a DC power source failure defined by TVA for the three units.
NEDO-24266 2.
CONCLUSIONS The general conclusions reached from this study are:
(1)
The conservative bounding PCT calculated for small break LOCAs with the DC power source failure that also disables Depressurization System for Browns Ferry Units than 1900'F.
the Automatic 1, 2, and 3 is less (2)
For large break LOCAs, the maximum PCT for any unit is not affected by a DC power source failure (Reference 2).
(3)
The MAPLHGR for any unit is not affected by a DC power source
~l failure (Reference 2).
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NEDO-24266 3.
BREAK ANALYSIS For the Browns Ferry Units 1, 2 and 3 plants with the LPCI modification, the most severe DC power source failure that will also disable the ADS will result in one Core Spray system, three LPCI pumps injecting in both recirculation
- loops, and the HPCI (1CS + 3LPI2 + HPCI) all remaining operable (3LPI2 = 3 LPCI pumps in two recirculation loops).
.The worst postulated line break locations are the recirculation discharge line where two of the three LPCI pumps infect coolant to the reactor (1CS + 1LPCI + HPCI) and the feedwater line where the HPCI injects coolant (1CS + 3LPI2), because these break locations will result in the maximum degradation of the remaining emergency cooling capability.
'I Analysis was performed conservatively assuming total loss of the ECCS flow in the broken lines regardless of break size.
3.1 RECIRCULATION LINE DISCHARGE BREAKS 2
Ten break sizes were analyzed between 0.05 ft and
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0.50 ft with one gqre Spray recirculation loop)', and the I
As seen from the results in Table 3-1, no core uncovery is predicted for break sizes
<0.15 ft.
The CHASTE05 code has been used only when the ultraconserv-ative REFLOOD results exceeded the PCT limits, since the CHASTE05 code more accurately predicts PCT.
The highest PCT calculated with 2
break size of 0.3 ft.
The PCT calculated to be 1693'F.
Water break size are shown in Figures the CHASTE05 code was 1864'F for 7x7 fuel at for 8x8 fuel at the 0.3 ft break size was level, pr'essure and PCT plots for this limiting 3-1 and 3-2.
NEDO-24266'.2 FEEDWATER LINE BREAKS Small feedwater line breaks with the total loss of both HPCI and ADS are similar in behavior to the steamline break outside of the containment case discussed in the Final Safety Analysis Report (FSAR), in that the reactor vessel will not depressurize to the operating point of the low pressure ECC systems in the short-term without manual action to depressurize the reactor.
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lt Manual actuation of four-out-of-six total ADS valves was assumed at 10 minutes after the instant of the break initiation, which is consistent with the above steamline break case assumption of 10-minute operator action time.
The assump-Ption of four-out-of-six valves is conservative because recent guidelines (THI) instruct the operator to open as many valves as there are in the ADS, which, in this case, is six valves.
The assumption of four valves resulted from the fact that, after the assumed subject DC power source failure, only four of the '~
ADS valves can be operated manually; but there are many additional SRVs that could be'manually actuated by the operator.
2 2
Thirteen break sizes, ranging from 0.01 ft to 0.50 ft, were analyzed with the SAFE03 and REFLOOD05 computer programs, and PCTs were conservatively calculated.
with the REFLOOD05 small break model.
As seen from the results in Table 3-2, 0
2 the maximum calculated PCT with REFLOOD05 was 1880'F at a break size of 0.25 ft 2
The water level, pressure and PCT plots for this 0.25 ft break size are shown in Figures 3-3 and 3-4.
3-2
NEDO-24266 Table 3-1 CALCULATED PCT RESULTS FOR A RECIRCULATION DISCHARGE WITH A DC POWER SOURCE FAILURE (BROWNS*FERRY UNITS 1, 2 AND 3)
I LINE BREAK, Break Area (ft2) 0.05 0.07, 0.09 0.11 0.13 0.15 0.20
- 0. 30
- 0. 40, 0.50 PCT (REFLOOD)
('F) core does not uncover core does not uncover core does not uncover core does not uncover core does not uncover core does not uncover 1840 2307 1753 1686 PCT (CHASTE)
( F) 1864 Table 3-2 CALCULATED PCT RESULTS FOR A FEEDWATER LINE BREAK WITH A DC POWER -SOURCE FAILURE (BROWNS FERRY UNITS 1, 2 AND 3)
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= Break Area (ft2) 0.01 0.03 0.05 0.07 0.10 0.15 0.20 0.25 0.30 0.35 0.40 0.45 0.50 PCT (REFLOOD)
('F) 1221 1233 1194 1151 1126 1269 1413 1880 1813 1586 1414 1311 1211
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NEOO-24266 1.0 0.9 1 VESSEL PRESSURE IIysie) 2 8AF 18.03 ft 3 TAF 30.03 ft 4 WATER LEVEL(R) 48 45 0$
40 0.7 35 o
3 x
06 3
0.5 g
0.4 30.03 llf 25 tt:
I 1ILoa 0.3 0.2 12 0.1 0
100 200 300 400 500 600 700 800 TIME {sec) 900 1000 Figure 3-1.
Browne Ferry 1, 2 and 3 0.3 ftl Recirculation Diecharge Break DC Power Source Failure (No ADS)
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3-4
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6 PEAK CLAOOING TEMPERATURE REFLOOO HYC St Ih.fi> F h ~ 10,000 h ~25 0
0 100 200 h ~ 0.0
, 300 TIME Isec) 400 Figure 3-2a.
Browns Ferry l, 2 and 3 0.3 ft Recirculation Discharge Break DC Power Source Failure (No ADS) (8x8 Fuel Type) 3-5
NEDO-24266 20 19
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PEAK CLADDING TEMPERATURE CHASTE OS HTC Bcu/hr ft> F h ~ 10,000 0
0 h~O 300 TIME bee) h ~25 400 Figure 3-2b.
Browns Ferry 1, 2 and 3 0.3 ft2 Recirculation Discharge Break DC Power Source Failure (No ADS) (7z7 Fue] Type) 3-6
NEDO-24266 1.0 4
0.9 1 VESSEL PRESSURE (psis) 2 8AF 18.03 ft 3 TAG 30.03 ft 4 WATER LEVELtff) 48 45 08 40 07 X
0.6 CCD g
0.5 0.4 1
4 35 W
30.03 25 18.03 0.3 0.2 0.1 0
0 100 200 300 400 500 600 700 800 900 1000 TIME bee)
Figure 3-3.
Browns Ferry 1, 2 and' 0.25 ft Feedwater Line Break DC Power Source Failure Manual ADS (4 Valves) at 10 Minutes 3-7
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- C PEAK CLADDING TEMPERATURE REFLODD 05 HTC Biu/hr.ft> F o
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13 us 12 KD 11 K
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~ 3 h ~ 10,000 h & 0.0 h 0 25 0
100 200 300 400 500 600 700 800 900 1000 1100 1200 TIME Isscl Figure 3-4.
Browns Ferry 1, 2 and 3 0.25 ft Feedwater Line Break DC Power Source Failure Manual ADS (4 Valves) at 10 Minutes
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NEDO-24266 4.
REFERENCES 1.
, "Safety Evaluation by the Office of Nuclear Reactor Regulation Supporting Amendment No.
51 to Facility Operating License No. DPR-33, Amendment No.
45 to Facility Operating License No. DPR-52, Amendment No.
23 to Facility Operating License No. DPR-68, Tennessee Valley Authority, Browns Ferry Nuclear Plant, Units Nos. 1,
~ 2, and 3".
2.
- Letter, R. E. Engel to P.
S. Check, "D.C. Power Source Failure for BWR/3 and 4", dated November 1, 1978.
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