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MONTHYEARML0703600122007-01-31031 January 2007
E-Mail from W. Kennedy to Dr. Richards of NIST Transmitting the Draft Rais for Re-Licensing of the Nbsr, Docket No. 50-184 (TAC No. MD 3410)
Project stage: Draft RAI
ML0703200102007-02-0202 February 2007
National Institute of Standards and Technology Test Reactor (Nbsr) - Draft Request for Additional Information (Rai), Regarding Amendment Request for Re-Licensing
Project stage: Draft RAI
ML0704400432007-02-13013 February 2007
Request for Additional Information Regarding the National Institute of Standards and Technology Test Reactor Application for Re-Licensing
Project stage: RAI
ML0818302942008-07-0101 July 2008
Request for Additional Information Regarding the National Institute of Standards and Technology Test Reactor Application for License Renewal
Project stage: RAI
ML0822802192008-08-25025 August 2008
Request for Additional Information Regarding Financial Qualifications and Decommissioning Funding for the National Institute of Standards and Technology Test Reactor Application for License Renewal
Project stage: RAI
ML0835107542008-09-12012 September 2008
NIST Response to NRC Request for Additional Information (TAC No. MD3410) Dated 07/01/2008
Project stage: Response to RAI
ML0828903382008-09-16016 September 2008
NRC Request for Additional Information Re Revised Nbsr Technical Specifications and Nbsr Emergency Plan
Project stage: Request
ML0830302092008-10-21021 October 2008
Response to NRC Request for Additional Information Financial Qualifications and Decommissioning Funding for the Test Reactor Application for License Renewal
Project stage: Response to RAI
ML0832307102008-11-19019 November 2008
Request for Additional Information Regarding Emergency Planning for the National Institute of Standards and Technology Test Reactor License Renewal
Project stage: RAI
ML0903707372009-02-18018 February 2009
National Institute of Standards and Technology Test Reactors - Request for Additional Information Operator Requalification Program
Project stage: RAI
ML0907001322009-03-0303 March 2009
Us Dept. of Commerce, NIST, Response to Request for Additional Information on Operator Requalification Program
Project stage: Response to RAI
ML0906100612009-03-0303 March 2009
National Institute of Standards and Technology Test Reactor - Request for Additional Information Advisory Committee on Reactor Safeguards Subcommittee Meeting Follow-up Items (Tac MD3410)
Project stage: RAI
ML0908903272009-03-19019 March 2009
Submittal of Response to Request for Additional Information Regarding Advisory Committee on Reactor Safeguards Subcommittee Meeting Follow-Up Items and the Revised Nbsr Technical Specifications
Project stage: Response to RAI
ML0911904642009-04-22022 April 2009
Memo to Bill Kennedy in Response to ACRS Question (April 2, 2009 Meeting)
Project stage: Meeting
ML0911904512009-04-22022 April 2009
NIST Response to ACRS Question
Project stage: Other
ML0911700022009-07-0202 July 2009
National Institute of Standards and Technology, Notice of Issuance of Renewed Facility Operating License No. TR-5
Project stage: Approval
ML0909901352009-07-0202 July 2009
National Institute of Standards and Technology - Safety Evaluation Report for Renewed Facility Operating License TR-5 (Tac MD3410)
Project stage: Approval
ML0909901072009-07-0202 July 2009
FRN: National Institute of Standards and Technology - Notice of Issuance of Renewed Facility License No. TR-5
Project stage: Approval
ML0909901202009-07-0202 July 2009
National Institute of Standards and Technology, License Amendment, Issuance of Renewed Facility License No. TR-5 for the National Bureau of Standards Reactor
Project stage: Approval
ML0909901412009-07-0202 July 2009
National Institute of Standards and Technology - Technical Specifications for Renewed Facility License TR-5 (Tac No. MD3410)
Project stage: Other
2008-08-25
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Text
MEMORANDUM April 22, 2009 To: Mr. Bill Kennedy Program Manager NIST/NBSR From: Dr. Wade J. Richards Chief, Reactor Operations and Engineering
Subject:
Response to ACRS Question (April 2, 2009 Meeting)
During our analysis of an accident postulated by the ACRS (accidental closing of the valve in the reactor outlet line, DWV-19), we became aware of a possible problem in our original analysis of the transient related to a loss of off-site power. In that analysis, a coast-down curve was derived within RELAP, based upon pump characteristics and the flow geometry. This curve was compared to a coast-down curve measured before initial reactor criticality, and shown to be conservative. However, the original curve was not measured under conditions appropriate to the accident being analyzed.
In the figure below, the flow for the initial few seconds of the new curve is compared to that used in the prior analysis. It should be noted that the minimum Critical Heat Flux Ratio (CHFR) occurs during this time period.
In order to address this issue, the primary coast-down curves were recently measured under various conditions, including the limiting case of no shut-down pump starting. The measured curve for the limiting case was ten used to derive a model as input to RELAP for complete re-analysis of the accident scenarios explored in the original analysis.
From the figure, it is clear that the two curves are almost identical for the first 3 seconds, and this fact is reflected in the analysis using RELAP. The minimum CHFR occurs in the outer plenum approximately 1.5 s after the primary pumps trip, and is 2.17 as compared to 2.19 previously. This result is a direct consequence of the comparison shown in Fig. 1, which is in turn a measure of the conservatism in the coast-down curve used in the prior analysis. It should be noted that the present curve is a conservative representation of the measured curve, and that there is substantial conservatism in the new result (arising from non-local deposition of fission power, neglect of clad and meat thermal conductivity, changes in the distribution of heat deposition for fission product decay, and other terms).
The result of the Loss of Offsite Power accident analysis (and by extension, the Loss of Both Shutdown Pumps) is nearly identical to the initial analysis.
The reactor scram was initiated by low flow to the outer plenum 1.4 seconds after the loss of power. The maximum fuel centerline temperature is 410 K, which occurs at the hot spot at the top of the hottest element in the outer core at the time of the scram. Long term cooling in the case of failed shutdown pumps is bounded by the case of the closure of DWV-19, in which the flow drops to zero just 4 seconds after the scram.
In summary, the new analysis, which is conservative, confirms the earlier analysis and the results presented in the Safety Evaluation Report. It will be included in the revised FSAR, as will all the responses to the previous RAI.
Figure 1. Comparison of the coast-down curve used in the prior analysis to that determined from new measurements.