ML20035G013
| ML20035G013 | |
| Person / Time | |
|---|---|
| Site: | 05200001 |
| Issue date: | 04/21/1993 |
| From: | Kelly G Office of Nuclear Reactor Regulation |
| To: | Duncan J GENERAL ELECTRIC CO. |
| Shared Package | |
| ML20035G010 | List: |
| References | |
| NUDOCS 9304260026 | |
| Download: ML20035G013 (5) | |
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UNITED STATES NUCLEAR REGULATORY COMMISSION n
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April 21, 1993-NOTE !0:
J ck uncan, GE f FROM:
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SUBJECT:
T 00GHTS Of THE ABWR INTERNAL FLOODING ANALYSIS I have enclosed some thoughts I put together from some comments I received from my contractor on the ABWR internal flooding analysis.
I would like to discuss these with you later this week.
Please give me a call if you have any questions. Some of the points enclosed are similar to those I faxed to you on March 25, 1993, but for which I have received no written reply.
Enclosure:
as stated 9304260026 930421
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{DR ADOCK 05200003 PM 6
l ENCLOSURE Thoughts on the ABWR Internal Flooding Analysis April 21, 1993 (1)
The GE bounding analysis identified the potential flood sources and selected the ones expected to have the greatest impact on the operability of systems required to safely shutdown the plant. No sources were identified that could affect more than one division of equipment without the failure of at least some flood protection features, either isolation barriers or mitigating features.
In the review of the flood sources the staff found that at least one flood source, the reactor cooling water surge tanks, was prevented frcm affecting more than one division through the use of raised sills on the entrance to the room containing the tanks. This is the only feature that prevents this flood source from potentially affecting two divisions of ECCS equipment.
Because of the significance of a flood that would affect multiple system divisions, GE should assure that the proper 1
installation of the sills is incorporated into the ITAAC for flood protection. This assurance should be provided by an ITAAC.
(2)
The analysis did not address flooding in areas of the design that are the responsibility of the COL applicant, specifically the ultimate heat sink pump house. The COL Action Item 6 of Section 19.9.10 should be revised to include the requirement of a probabilistic assessment of the site specific design features.
(3)
The review of the event trees and the quantification of the event trees did result in some identification of issues, the resolution of which would have some affect on the results of the analysis.
In the turbine building flood analysis, one of the events taken credit for is the release of flood watert through the turbine building truck entrance door.
Because this door is not a watertight door, the analysis models the failure of the door to fail open with a probability of 0.05.
However, no justification for this value is provided and the assumption that the door would fail is contrary to normal PRA practices that do not use beneficial failures as possible mitigating measures. The staff requests that GE provide the staff with an assessment of the height of water that could be retained by the truck door before it would be expected to fail and how this height relates to the flood level necessary to allow for flooding in the control building. Alternatively, credit for the failure of the door should be removed from the event tree.
(4)
This event tree also takes credit for the protection provided by the watertight door between the turbine building and the control building, failure of the door to stop the flood is assumed to result in the failure of all equipment in the control building. While this is reasonable, it highlights the need to ensure that the watertight door is maintained and inspected properl".
n a response to an RAI, GE states that (r?dit was taken because in u. ant operating plants the status of the door is typically visually checked during each shift.
Since this
' Action Item is required to ensure that these inspectias incorporated into plant operating procedures oor, a COL ons are (5)
In the quantification of the control buildin event titled " automatic flooding isolation."g event tree, there is an as representin This event is quantified water system. g the failure to isolate the supply side water system valve, siphon breaker and pump trip failures conceptual design of the reactor service water system shows onl However, the motor opera discharge. ted valve between the control building an y one elevation than the basement of the control building break, backflow into the building would be possible
, the location of the flood becomes the most dominant flood in all of the thThis The NRC is unclear whether additional protection is ree buildings.
likelihood of this unisolated flood. Insufficient inforneeded available to determine if this issue is of concern for the c mation was water system this concern,and the turbine building service water system.
rculating the discharge lines should be addressed.the susceptibility of these sy As part of (6) event titled " sump level switches detect flood "In the rea ree there is an cannot isolate the flood. credit is taken for the operation of the sump p.
In the event tree no umps and the operator Additionally, in this event tree and in the reactIt is the event tree.
building flooding in corridor, the possibility of fl all three divisions of ECCS was not considered or ood waters entering between the conditional core damage probabilities asDue to the relationsh
%ss of one, two, and three divisions of ECCS sociated with the failure of all three watertight doors and the subseq, the sequence con frequency of core damage than the sequences conta uent failure to g er one or two ECCS divisions due to the flood.
g the failure of The reactor building flood event trees consi [this is an insight]
(7) watertight doors protecting the ECC5 rooms. der the failures of the considered, common cause failures are not.Only random failures are two event trees were requantified to address thesAs part of the review, these three watertight doors and the data of table 19RS 2 sequ The ure of all common cause failure of the watertight doors was used to address requantified This sequence,,had a core damage frequency gre.
Only one se flood in the ECCS corridor followed by the common caw ater th per year.
per year is a three watertight doors and the failure of the react use failure of all shutdown with all three divisions of the ECCS unavailabl or to be safely consistent with the GE analysis, this requantification us d th To be value for shutdown of the plant (which assumes that the p e
e 0.1 system is unavailable) rather than the loss of division 1 ower conversion power or service water systems.
significantly increases the fre Although the requantification, 2, and 3 floods in the reactor building,quency of core damage due to internal there are reasons that this particular I
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I internal flood becoming a potential vulnerability. This is because of the conservatism of the underlying analysis, the absolute value-of the estimated core damage frecuency, and the significant uncertainty in the common cause failure value used for the three ECCS doors.
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