ML19329B355
| ML19329B355 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse  |
| Issue date: | 08/14/1975 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Deyoung R Office of Nuclear Reactor Regulation |
| References | |
| NUDOCS 8002040777 | |
| Download: ML19329B355 (7) | |
Text
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ri AUG k41975 R. C. DeYoung, Assistant Director for Light Water Reactors, Group 1. RL AA3 INPUI TO DAVIS-BESSE SER PLANT HAME: Davis-Besse Unit 1 LICE:XSING STAGE: OL DOCKET NU!GER:
50-346 HILESTONE NDGER: 24.31 RESPONSIBLE BRANCH: LUR l-4; L. Engle, LPM REQUESTED COMPLEIION DATE: June 30, 1975 REVIEW STATUS: AAB input partially complete Attached is additional Accident Analysis Branch input for the Javis-3 esso SER. This input by H. Fontecilla includes additional design basis accident doses and assumptions.
This review was coordinated by C. Perrell.
Harold R. Denton, Assistant Director for Site Safety Division of Technical Reviev Office of Nuclutr Reactor Regulation
Enclosure:
As stated cc: w/o enclosure A. Giambuseo W. Mcdonald J. Pansarella w/ enclosure See page 2 l
8002040771
7-s w
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R. C. DeYoung,
1 cc: w/ enclosure S. Hanauer R. Heineman R. Boyd RL A/D's TR A/D's SS 3/C's TR T/C's D. Eisenhut S. Verga R. Klecker A. Schwencer L. Engle W. Pasadag
-E. Adansaa K. Campe II. Fontecilla K. Murphy C. Ferrell Distribution:
Central Files NRR/ Reading AAB/ Reading 1
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m 15.3 & 15.4 STEAM GENERATOR TUBE FAILURE & MAIN STEAM LINE FAILURE ACCIDENTS On the basis of our experience with the evaluation of the steam line break and the steam generator tube rupture accidents for PWR plants of similar design, we have concluded that the consequences of these accidents can be controlled by limiting the permissible reactor coolant and secondary coolant radioactivity concentrations so that potential offsite doses are small. We will include appropriate limits in the Technical Specifications on these coolant activity concentrations.
Similarly, we will include appropriate limits in the Technical Specifications on gas dacay tank activity so that a single failure (such as sticking and lifting of a relief valve) does not result in doses that are more than a small fraction of the 10 CFR 100 guidelines.
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TABLE V DOSES THYROID WHOLE BODY (REM)
(REM)
Tube Rupture Accident 1.5
.4 Tube Rupture Accident with Coincident Iodine Spike 30
.4 Steam Line Break 1.5
.35
<.1
. Loss of Offsite Power with Coincident Iodine Spike 0.6
<.1 Rod Ejection Accident Case I*
121
<1 Case II**
60 1.6 Red Ejection Accident (0-8 hr LPZ X/Q =
1,8 x 10-5 sec/m3) 4
<,1 Red Ejection Accident (8-24 hry"Z X/Q =
3 1,3 x 10 sec/m )
<.1
<.1
- Releases through the containment.
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- Releases through the secondary system.
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n STEAM LINE BREAK & STEAM GENERATOR TUBE RUPTURE ACCIDENT ASSUMPTIONS:
Power =
2772 Mwth 3
2-hour X/Q = 5.2 x 10-4sec/m at exclusion boundary Iodine decontamination factor of 10 between water and steam Primary and secondary coolant equilibrium concentratiens as limited by standard Technical Specificaticas (l.s uCi/gr I-131 Eq and 100/E uCi/gr Noble Gases for primary coolant and.1 uci/gr I-131 Eq for secondary coolant)
Primary to secondary leak rate as limited by standard Technical Specifications (1 gpm)
For accidents assumed to occur in coincidence with an iodine spike, the* primary coolant concentration is as li=ited by the standard Technical Specificaticns for 48-hcur pericds (60 uCi/gr I-131 Eq at 100% power)
Source spike factor of 500 after accidents 10% of iodine and roble gases fuel activity in gaps All releases through the secondary system (except Rod Ejection Accident, Case 1) 28% fuel with clad failures after rod ejection accident 0% fuel reaching initiation of melting W
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m Control Rod Ejection Accidbnt The assumptions used to calculate offsite doses from e control rod ejection accident are:
Case I 1.
Power level of 2772 Mwt.
2.
'28% fuel failed in transient.
3.
10% of iodine and noble gas inventory in gap of fr.iled fuel.
4.
Release of total gap activity in failed fuel to containment building.
5.
50% plate-out of radioactive iodines.
6.
Containment building sprays are not initiated.
7.
Containment building leak rate of 0.30%/ day for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> and one-half this value thereafter.
8.
Standard ground level release meteorology and dose conversion factors.
Case II 1.
Power level of 2772 Mwt.
2.
28% fuel failed in transient.
3.
10% of iodine and noble gas activity in gap of failed fuel.
4.
Release of total gap activity in failed fuel to reactor coolant.
5.
Reactor coolant to secondary coolant operational leakage is 1 gpm.
l
mm
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1 4
2-I Case II (Cont.'d.)
6.
Loss of off-site power so that steam is released from secondary side relief valve.
7.
Reactor coolant-secondary coolant equilibrium reached at 16 minutes t
i after the accident.
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