ML091190209

Transmit Revised Draft RAI from the Probabilistic Risk Assessment Licensing Branch on the Proposed Extended Power Uprate Amendment

ML091190209
Person / Time
Site:	Monticello
Issue date:	04/29/2009
From:	Tam P Plant Licensing Branch III
To:	Pointer K, Salamon G, Tam P Northern States Power Co, Plant Licensing Branch III
Tam P
References
TAC MD9990
Download: ML091190209 (4)

Text

Accession No. ML091190209 From:

Peter Tam Sent:

Wednesday, April 29, 2009 8:56 AM To:

Peter Tam; 'Pointer, Kenneth'; 'Salamon, Gabor' Cc:

Jigar Patel; Steven Laur; Andrew Howe; Donnie Harrison

Subject:

Monticello - Revised RAI from the PRA Branch re. the Proposed EPU Amendment (TAC MD9990)

Ken:

On 4/17/09 the NRC staff discussed with you and others a set of draft RAI from the Probabilistic Risk Assessment Licensing Branch and transmitted to you by an e-mail dated 3/18/09 (Accession No. ML090780004). Based on that teleconference we revised a number of questions and eliminated 2 questions. Please see below the revised set of PRA-related questions, which replaces the set sent by the 3/18/09 e-mail. We request that you respond by 5/31/09. If more time is needed to prepare your response, feel free to contact me to discuss.

(1) The NRC staff's evaluation of the individual plant examination of external events (IPEEE) report specifically specified that additional analysis is necessary to identify if single pump success is adequate for the Service Water System. The licensee's response dated 2/4/09, NRC Review Item (7), indicates that the current internal events Monticello Nuclear Generating Plant (MNGP) PRA Model of Record assumes that a single Service Water pump is adequate to successfully accommodate post transient cooling requirements. Describe how MNGP confirms that the single Service Water pump assumption modeled in the PRA is adequate for post-extended power uprate (EPU) requirements.

(2) EPU Safety Analysis Report (SAR) Enclosure 15, Section C.3 discusses, assessments for the 2003 Monticello PRA model against the American Society of Mechanical Engineers (ASME) standard and NRC draft Regulatory Guide DG-1122 performed by Applied Reliability Engineering, Inc. (ARE), in early 2004. This section does not provide information on findings and comments related to this assessment. Please identify and discuss any open items resulting from the 2004 ARE assessment, identify those that may impact the EPU, and provide justification for those open-items that affect the EPU.

(3) EPU SAR Enclosure 15, Section 4.1.1 and Section 4.5 states that no significant impact on internal flooding initiator frequencies are postulated due to the EPU. Since higher flow rates can contribute to changes in initiator events for floods, please provide a more thorough justification for your conclusion on how EPU flow rates will not affect internal flooding initiator frequencies.

(4) The NRC staff notes that success criteria changes crediting Control Rod Drive Hydraulic (CRDH) by depressurizing the reactor is unique compared to boiling water reactor (BWR)

EPUs previously approved. If depressurization is successful, then the low pressure injection sources would be available, obviating any need for successful CRDH injection. Further, if changes to emergency operating procedures (EOP) are required to implement this change, then this change could potentially complicate operator response to other events. The NRC staff requests responses to the following issues to better

understand the impacts of this change. If new operator actions are required, then questions d, e, and f apply.

a. Was depressurization required in order to credit two CRDH pumps pre-EPU?

Describe the analyses conducted which determined that depressurization was necessary for EPU conditions.

b.

Describe how the action for depressurization has been modeled in the PRA.

Explain how the human reliability analysis (HRA) for this action compares to any other actions to depressurize the reactor. Describe how the HRA dependency of this action on other operator actions was assessed.

c. Does the new requirement for reactor depressurization to allow CRDH injection create new sequences and end states? Provide the basis for this conclusion and a summary of any resulting changes to the sequences and end states.

d.

Describe the risk significance of the CRDH success criteria change (i.e., Fussell Vesely Importance and Risk Achievement Worth) and contribution of the sequences associated with this change to core damage frequency (CDF) and large early release frequency (LERF).

e.

Did changes to CRDH success criteria require changes to EOPs? If yes, describe the changes, operator training and validation methods, and why the changes to plant-specific EOPs remain consistent with BWR EOP guidelines or were otherwise determined to be acceptable. Your response should also discuss any potential negative impacts from the operator inappropriately depressurizing the reactor due to the new procedures.

f.

Was a focused peer review performed for the PRA model changes necessary to incorporate the new CRDH success criteria including any new event tree structure and operator actions? If yes, provide the results of that peer review. If not, please justify why a peer review was not judged to be required.

(5) Describe the calculation for the change in risk as a result of needing one additional safety/relief valve (SRV) to open for anticipated transient without scram (ATWS). Does the change in risk calculated for EPU include a contribution due to the SRV success criteria? Discuss how the change in SRV success criteria has been incorporated into the PRA model and include a discussion of the changes to common cause failure events and their basis.

(6) EPU SAR Enclosure 15, Section 5.6 states that the EPU change in power represents a relatively small change to the overall challenge to the containment under severe accident conditions. Please provide additional details which justify this conclusion.

(7) EPU SAR Enclosure 15, Section 4.3.1 states fire PRA results are less impacted by changes in operator actions timings than the internal events PRA results. The re-rate safety evaluation dated September 16, 1998, Section 5.3, states: The CDF contribution from internal fires increased from 8.34E-6/Year to 8.8E-6/Year. This was attributed solely to the increase in human error rates because the time available to perform various accident mitigating tasks decreases with uprate. Please provide additional justification for the conclusion stating fire PRA results are less impacted by changes in operator

actions timings than the internal events PRA results. Please explain the inconsistency between the re-rate fire CDF increase and the EPU fire CDF increase.

(8) The NRC staff requests responses for the following information on Low Power and Shutdown PRA Explain how the EPU affects the scheduling of outage activities.

Provide additional information regarding the reliability and availability of equipment used for shutdown conditions.

Explain how the EPU affects the availability of equipment or instrumentation used for contingency plans.

Explain how the EPU affects the operation to close containment during loss of shutdown cooling.

(9) EPU SAR Enclosure 15, Section 3.3.2, indicates that the changes to EOPs and severe accident management guides as a result of the EPU were not available prior to completion of the PRA evaluation and it was assumed that the procedural changes would have a minor impact on the PRA results. The NRC staff needs to conclude that EOP impacts are minimal. Therefore, please provide a schedule for the development of the final draft of EOPs, and confirm that the PRA results would only be minimally impacted.

(10) EPU SAR Enclosure 15 provides a quantitative assessment of the risk impact of the COP credit for low pressure ECCS pump NSPH for ATWS, SBO, and internal fires events. The analysis indicates a CDF for each of these events, but does not provide an LERF metric. Please provide an LERF metric for each of the aforementioned events.

Peter S. Tam, Senior Project Manager Plant Licensing Branch III-1 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Tel. 301-415-1451 E-mail Properties Mail Envelope Properties ()

Subject:

Monticello - Revised RAI from the PRA Branch re. the Proposed EPU Amendment (TAC MD9990)

Sent Date: 04/29/2009 7:33:32 AM Received Date: 04/29/2009 8:56:00 AM From: Peter Tam

Created By: Peter.Tam@nrc.gov Recipients:

Peter.Tam@nrc.gov (Peter Tam)

Tracking Status: None Kenneth.Pointer@xenuclear.com ('Pointer, Kenneth')

Tracking Status: None Gabor.Salamon@xenuclear.com ('Salamon, Gabor')

Tracking Status: None Jigar.Patel@nrc.gov (Jigar Patel)

Tracking Status: None Steven.Laur@nrc.gov (Steven Laur)

Tracking Status: None Andrew.Howe@nrc.gov (Andrew Howe)

Tracking Status: None Donnie.Harrison@nrc.gov (Donnie Harrison)

Tracking Status: None Post Office:

Files Size Date & Time MESSAGE 35603 04/29/2009 Options Expiration Date:

Priority: olImportanceNormal ReplyRequested: False Return Notification: False Sensitivity: olNormal Recipients received: