ML21139A077
| ML21139A077 | |
| Person / Time | |
|---|---|
| Site: | 05200006 |
| Issue date: | 05/18/2021 |
| From: | Jennifer Dixon-Herrity Division of Operating Reactor Licensing |
| To: | Zeechung Wang Govt of China |
| Dixon-Herrity J | |
| References | |
| Download: ML21139A077 (4) | |
Text
From:
Dixon-Herrity, Jennifer To:
wang.zhaoran@mee.gov.cn Cc:
Belkys.SOSA@oecd-nea.org; zhang.lin; pei.wei; zhang.jing; ; zhaodanni Bcc:
Wittick, Brian; Hernandez, Raul; Jones, Steve
Subject:
FW: QUERY: SRP Section 15.7.5: Primary Branch Owner Date:
Tuesday, May 18, 2021 9:54:00 AM
Dear Zhaoran,
Im sorry about the delay in getting an answer back to you on your second set of questions. Technical staff at the NRC provided the following response.
From the heavy loads handling point of view, if a licensee is crediting the use of an approved single failure proof crane, regardless of the elevation the load is lifted to, no drop is postulated and no radiological analysis is required. As such the 3 follow-on questions are not applicable. To better explain this, the staff provides the following background.
NRC staff guidance for establishing postulated initiating events to be considered as design basis events is in SRP Chapter 15, Introduction - Transient and Accident Analyses. The selection of design basis accidents is primarily based on estimated frequency of occurrence, with consideration of certain events that could have particularly severe consequences (e.g., double-ended break initiators of loss of coolant accidents) without regard to frequency. The guidance includes frequency screening of initiating events that omits those events with frequencies below once in a million years if they are not significant from a consequence perspective. This approach is consistent with Requirement 16, Postulated Initiating Events, of IAEA SSR 2/1, Safety of Nuclear Power Plants - Design, Rev. 1.
The staff performs this initiating event frequency screening for heavy load handling in SRP Section 9.1.5, Overhead Heavy Load Handling Systems. The staff estimates of the frequency of heavy load handling system failures leading to load drops is on the order of once per one hundred thousand years when a standard handling system and good operating practices are employed. When a single-failure-handling system is used, as described in Section III.4.C of SRP Section 9.1.5, the staff estimate of the frequency of handling system failures drops by one to two orders of magnitude, such that the frequency is below once in a million years. The frequency estimates are based on analysis described in NUREG-0612, "Control of Heavy Loads at Power Plants, and have been updated through Generic Issue 186, Potential Risk and Consequences of Heavy Load Drops in Nuclear Power Plants. Therefore, a failure of a handling system meeting single-failure-proof criteria is not sufficiently likely to be included in the set of initiating events for design basis accidents.
The radiological consequence evaluation applies to design basis accidents. SRP Section 15.7.5, Spent Fuel Cask Drop Accidents, is marked with an asterisk (in the index in the NRC website at https://www.nrc.gov/reading-rm/doc-collections/nuregs/staff/sr0800/ch15/index.html) indicating that SRP Section 15.0.3, Design Basis Accident Radiological Consequences of Analyses for Advanced Light Water Reactors, should be used for the review of new applications. This is discussed in the Introduction to the SRP (NUREG-0800), which is available in ADAMS at ML070630046. As a consequence, SRP Section 15.7.5 was not updated for review of new light water reactors as other SRP sections were in 2007 and remains as Revision 1, issued in 1981. In 1981, no spent fuel storage cask had been licensed for use and only transportation casks licensed to 10 CFR Part 71 were licensed to transfer fuel from spent fuel pools.
Transportation casks are certified to withstand a 30 foot drop when impact limiting devices are present, and this is the basis for analyzing the consequences of casks lifted above the 30 foot level in SRP Section 15.7.5.
SRP Section 15.0.3 establishes dose consequence acceptance criteria for design basis accidents at the following locations: the exclusion area boundary, the low population zone boundary, the control room, and the technical support center. There are no acceptance criteria for accidents not considered in the design basis. Similarly, SRP Section 15.7.5 included the following statement in Section I, Areas of Review, Paragraph 1:
ASB [Auxiliary Systems Branch] is consulted to verify the potential drop height during handling of a loaded cask and the procedures for handling the cask with respect to the impact limiter. If the handling procedures meet all applicable criteria, then the radiological consequences of a spent fuel cask drop accident need not be estimated.
The second sentence provides for exclusion of spent fuel cask drop accidents when the handling procedures meet all applicable criteria, which refers to handling procedures that require handling of loaded casks using installed single-failure proof handling systems or that limit the handling conditions to those encompassed by the cask certification to 10 CFR Part 71 criteria. In that case, a cask drop accident is not considered part of the design basis.
Please let me know of questions or concerns.
Respectfully,
Jen
Jennifer L. Dixon-Herrity Chief, Licensing Branch IV Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation 301-415-2967
From: <wang.zhaoran@mee.gov.cn>
Sent: Thursday, May 06, 2021 11:40 PM To: Dixon-Herrity, Jennifer <Jennifer.Dixon-Herrity@nrc.gov>
Cc: Belkys.SOSA@oecd-nea.org; zhang.lin <zhang.lin@mee.gov.cn>; pei.wei <pei.wei@mee.gov.cn>;
zhang.jing <zhang.jing@mee.gov.cn>; <liuyu@chinansc.cn>; zhaodanni
<zhaodanni@chinansc.cn>
Subject:
[External_Sender] Re:RE: discussion refers to single failure proof polar crane
Dear Jen,
Thank you for your quick response.
Our techincal staff have some discussion and follow-up quesions, as below:
According to theanswer,myunderstanding is that:SRP 9.1.5 focus on the impact of that heavy loadsdrop on SSCs performing safety functions. If SRP section 9.1.5 (2007 Revision) section I.4 is adopted, the acceptance criteria in section 9.1.5 will be met. SRP 15.7.5 concerns the radioactive consequences of the spent fuel cask drops, regardless of whether the measures in I.4 of section 9.1.5 (such as single-failure-proof cranes) are taken, as long as the lifting height exceeds 9m or the spent fuel cask condition exceeds the configuration of the drop test, the analysis in section 15.7.5 is still required. I would like to know if my understanding is correct?
If the understanding above is correct, I have some follow-up questions:
a) Both the impact of heavy loads dropingon SSCs or spent fuel in SRP section 9.1.5 and the radioactive consequences of droping the spent fuel cask in SRP section 15.7.5, the radioactive consequences of nuclear power plant on the public and the environment are considered. SRP 9.1.5 considered that the use of single-failure-proof cranes can greatly reduce the probability of heavy load droping. Why should SRP 15.7.5 still assume the drop of spent fuel cask. It is hoped that you tell us the backgroud of this issue, what assessments have NRC made, or is it just a conservative assumption?
b) Can the seal integrity of the spent fuel cask be regarded as an acceptance criterion when analyzing the radioactive consequences of the drop of the spent fuel cask? That is to say, if the mechanical analysis can prove that the seal integrity of spent fuel cask will not be lost after dropingfrom a height of more than 9m, the analysis of radioactive consequences according to the method given in SRP 15.7.5 is not necessary? Has NRC used seal integrity as a justification for acceptance in the review?
c)For the second question, if the seal integrity can be taken as the acceptance criterion, what are the requirements for the calculation programe and models used in the analysis of the seal integrity of spent fuel cask falling from a height of more than 9m (such as 20 m), such as whether additional tests are required
to verify the analysis programe and models?
Best regards, Zhaoran