NUREG-1050, Forwards Regulatory Analysis of Generic Safety Issue 61, 'Safety/Relief Valve Discharge Line Break within Mark I or II Wetwell Airspace'. Based on Results Discussed in NUREG/CR-4594,risk Lower than Previously Estimated
| ML20205D931 | |
| Person / Time | |
|---|---|
| Issue date: | 08/08/1986 |
| From: | Harold Denton Office of Nuclear Reactor Regulation |
| To: | Speis T Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20205D934 | List: |
| References | |
| RTR-NUREG-0933, RTR-NUREG-1050, RTR-NUREG-933, RTR-NUREG-CR-2800, RTR-NUREG-CR-4594 NUDOCS 8608180209 | |
| Download: ML20205D931 (2) | |
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August 8, 1986 MEMORANDUM FOR: Themis P. Speis, Director Division of Safety Review & Oversight FROM: Harold R. Denton, Director Office of Nuclear Reactor Regulation
SUBJECT:
RESOLUTION OF GENERIC SAFETY ISSUE 61, "SRV LINE BREAK INSIDE THE WETWELL AIRSPACE OF MARK I AND II CONTAINME S" v
Generic Safety Issue 61 is resolved and requires no further study. This issue deals with a postulated stuck open SRV valve and a break occurring in that SRV discharge line inside the wetwell airspace. Such a break has the potential to overpressurize the wetwell containment structure leading to a loss of containment integrity and ultimately to possible core damage if alternate water sources are not available.
This issue was identified as a potential generic safety issue by the ACRS in April 1982, and was studied by Brookhaven National Laboratory (BNL) in 1982.
The results were reported in BNL-NUREG-31940 and were reviewed by NRC staff and the BWR Owners Group (BWROG). The assumptions and findings in BNL-NUREG-31940 were questioned by the staff and the BWROG from the view point of: energy and mass releases to containment based on transients analyzed, severity of assumptions embedded in scenarios selected, and assumptions related to containment failure leading to core melt. BNL reevaluated Generic Safety Issue 61 based on staff and public comments received and the results are reported in NUREG/CR-4594, June 1986.
The enclosed regulatory analysis which is based on the results discussed in NUREG/CR-4594, shows that the contribution of this postulated event to core melt frequency and public risk is much lower than previously estimated.
The conservative estimates are 2.5x10-7 to 2x10-10 core damage events per reactor year and a risk less than 5 person-rem per reactor year. It should also be noted that these calculated probabilities were developed for comparisons with other accident scenarios leading to core damage, thereby nificance. When calculated providing a basis probabilities becomeforless judging than relative safety10-approximately sig/Rx yr such values should not be viewed as an absolute numeric value, but rather interpreted as a probability of less than 10-7/Rx yr.
When these values are compared to other identified contributors to risk (e.g.,
NUREG-1050 and NUREG/CR-2800), the results do not support any new requirements or further study of this safety issue.
'T2D- 7 8608180209 860808 x RDo PDR TOPRP EMVGENE C PDR L ,
m August 8, 1986 NUREG/CR-4594 should be issued to provide a technical record of current findings related to the safety significance of Generic Safety Issue 61 and this issue should be declared technically resolved r.d removed from the list of active generic safety issues. The section def.ing with Generic Safety Issue 61 in NUREG-0933 should be revised accordingly.
Naind Satad bl N. R. Desten Harold R. Denton, Director Office of Nuclear Reactor Regulation
Enclosures:
- 2. Regulatory Analysis of GI-61 cc: V. Stello, ED0 J. Sniezek, CRGR R. Fraley NRR DD's Distribution C 1 Central:Filel a RSIB R/F DSR0 C/F ASerkiz GMazetis WMinners BSheron TSpeis JFunches RVollmer/HDenton ASerkiz c/f PDR
- see previous concurrence RSIB:DSR0* RSIB:DSR0* RSIB:DSR0* DD D:PPAS :NRR/ $$h ASerkiz:sm GMazetis WMinners BS IS is JFunches chut En696 06/02/86 06/02/86 06/02/86 (p/ /8 6 /86 f /g/86 / 6 /$/86 l
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