NRC-2016-0233, Comment (4) of Anonymous Individual on Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents
| ML17109A359 | |
| Person / Time | |
|---|---|
| Site: | Summer  |
| Issue date: | 04/14/2017 |
| From: | - No Known Affiliation |
| To: | Rules, Announcements, and Directives Branch |
| References | |
| 81FR83288 00004, NRC-2016-0233 | |
| Download: ML17109A359 (1) | |
Text
'Z>l7 As of: 4/17/17 10:28 AM Received: April 14, 2017 Status: Pending_Post Page 1of1 PUBLIC SUBMISSIONt ti.r~ ! 7 i.tt 10: :,o Tracking No. lkl-8vtu-n8fj Comments Due: April 21, 2017 Submission Type: Web
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Docket: NRC-2016,-0233 Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents Comment On: NRC-2016-0233-0003 Pressurized Water Reactor Control Rod Ejection and Boiling Water Reactor Control Rod Drop Accidents; Extension of Comment Period Document: NRC-20 l 6""023 3-DRAFT-0005 Comment on FR Doc# 2017-02073 Submitter Information Name: Anonymous Anonymous General Comment
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/,f-. 0 Virgil C. Summer Nuclear Station Unit 1 is providing the following comments concerning its review of Draft Regulatory Guide (DG) 1327.
Section 2.1.3: Please clarify what kind of manufacturing tolerances are referred to here. Does this require a statistical analysis with 95/95 uncertainty?
Section 2.2.3: Given that a large majority of the time each reactor spends at power is near 100%, can low power conditions be excluded from the analysis? Many transient analyses are perfofll1ed at zero power.and full power based on probability. It would be very time-consuming to determine if intermediate power levels are more limiting at each bumup interval. It would seem that even for a load-following plant, examinations of 0, 80%, 90%, and 100% would be sufficient to cover 99% of the probability distribution.
Section 2.5.1: For control rod ejection (CRE), since.the reactivity-initiated accident (RIA) transient is caused by the pressure boundary breach, the analysis should be able to credit the pressure boundary breach in the peak RCS pressure analysis.
Section 4: This section should be removed from DG-1327. Information related to the performance of radiological consequence analyses should remain in RG 1.183.
Section 6: The reactor coolant peak pressure acceptance criterion is already defined in a plant's Final Safety Analysis Report and may differ from the limit defined in DG-1327. The Regulatory Guide should not override existing license~ limits._
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