ML093200660

Request for Additional Information Related to One-Time Extension of Essential Service Water Train Completion Time

ML093200660
Person / Time
Site:	Byron
Issue date:	12/18/2009
From:	David M Plant Licensing Branch III
To:	Pardee C Exelon Nuclear
david marshall NRR/DORL 415-1547
References
TAC ME2293, TAC ME2294
Download: ML093200660 (3)

Text

UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20555-0001 December 18, 2009 Mr. Charles G. Pardee President and Chief Nuclear Officer Exelon Nuclear 4300 Winfield Road Warrenville, IL 60555 SUB~IECT: BYRON STATION, UNIT NOS. 1 AND 2 - REQUEST FOR ADDITIONAL INFORMATION RELATED TO ONE-TIME EXTENSION OF ESSENTIAL SERVICE WATER TRAIN COMPLETION TIME (TAC NOS. ME2293 AND ME2294)

Dear Mr. Pardee:

By letter to the Nuclear Regulatory Commission (NRC) dated September 24, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML092680090), as supplemented by letter dated November 13, 2009 (ADAMS Accession No. ML093200065),

Exelon Generation Company, LLC (the licensee), submitted a license amendment request proposing a one-time extension of the Completion Time to restore a unit-specific essential service water train to operable status for Technical Specification Limiting Condition for Operation 3.7.8, "Essential Service Water (SX) System," from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br />.

The NRC staff is reviewing your submittals, and has determined that additional information is required to complete its review. The specific information requested is addressed in the enclosed Request for Additional Information (RAI). The RAI was discussed with your staff on December 17, 2009, and they agreed to respond within 30 days after the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-1547.

Sincerely,

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Ma hall J. David, enio roj Pia t Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-454 and STN 50-455

Enclosure:

Request for Additional Information cc w/encls: Distribution via Listserv

REQUEST FOR ADDITIONAL INFORMATION BYRON STATION, UNIT NOS. 1 AND 2 DOCKET NOS. STN 50-454 AND STN 50-455 The Nuclear Regulatory Commission (NRC) staff is reviewing Exelon Generation Company, LLC's (the licensee's) license amendment request (LAR) dated September 24, 2009 (Agencywide Documents Access and Management System (ADAMS) Accession No.

ML092680090), as supplemented by letter dated November 13, 2009 (ADAMS Accession No.

ML093200065). The LAR requests a one-time extension of the Completion Time (CT) to restore a unit-specific essential service water train to operable status for Technical Specification (TS)

Limiting Condition for Operation 3.7.8, "Essential Service Water (SX) System," from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br />. The NRC staff has determined that the following additional information is required to complete its review.

1. The licensee identified that it used a separate fire probabilistic risk assessment (PRA) model to assess the risk impact from internal fires for this LAR. This model was identified as being "not fully realistic," and undergoing a significant revision to incorporate methods defined in NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities,"

September 2005. While the licensee provided some of the characteristics of the fire PRA model, the NRC staff requests the following additional information relevant to the technical adequacy of this model for this LAR.

a. Address how the current Byron fire PRA model satisfies the technical elements identified in Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk Informed Activities," Revision 1, January 2007, Section 1.2.4. Provide an appropriate justification of any outstanding plant changes not incorporated into the fire PRA model per RG 1.200, Revision 1, Section 4.2.

b. Describe where the existing fire PRA methods deviate from the methods described in NUREG/CR-6850, and justify the methods actually used.

c. Describe how spurious component operations are addressed in the fire PRA model.

d. Describe how plant components modeled in the PRA, but not in the scope of the plant's fire safe shutdown licensing basis, have been addressed in the fire PRA. Specifically identify how cable route information for such components is handled if these components are credited in the fire PRA.

e. Identify how the human reliability analysis was modified for the fire PRA to account for the impacts of fire on these actions; identify any fire-specific recovery actions credited in the fire PRA, their failure probabilities, and the basis for the probabilities.

f. No large early-release frequency analysis was completed for fires, justified based on no impact of service water on containment isolation capability. Provide a more specific justification based on the fire PRA results obtained evaluating the specific sequences ENCLOSURE

-2 which dominate the fire risk profile and their conditional probability of a large early release.

g. Identify the reviews (internal, external, and/or peer) completed for the fire PRA model to assure its quality.

2. The licensee has not specifically addressed how the cause/effect relationship was modeled in the PRA for this application.

a. Describe which basic event(s) in the PRA models (internal events and fire PRA) were modified to evaluate the SX outage, including the separate impacts on mitigation and initiating events.

b. Describe how any compensatory measures have been credited in the analysis, addressing both the baseline model and the configuration-specific assessment.

c. Identify whether recovery of inoperable SX pumps is credited, either for mitigation or for the calculation of initiating event frequency, and if so, justify the recovery probability applied.

d. Provide a sensitivity analysis for crediting any compensatory measures or recovery actions, which are unique for this evaluation.

3. The tier 2 evaluation identifies compensatory measures in Table 3, "Byron SX A Train Outage Summary of Compensatory Measures," in Attachment 1 of the LAR as regulatory commitments, but the description of these items is vague.

a. Describe specifically what restrictions on availability are being committed by use of the term "protected equipment."

b. Identify specifically each component which is "protected," without the use of undefined plant acronyms.

c. Describe the process applied and basis to determine the scope of the protected equipment.

d. Describe the fire zones qualitatively, and identify how these zones were selected for a compensatory measure.

4. In Attachment 5 of the LAR, the tier 3 evaluation identifies the plant Configuration Risk Management Program being applied to assess risk of emergent conditions. However, as identified in RG 1.177, "An Approach for Plant-Specific, Risk-Informed Decisionmaking:

Technical Specifications," August 1998, there is no discussion of how external event risk, specifically fire risk, which comprises 90 percent of the total risk reported for this TS change, is assessed within the tier 3 evaluation. Describe how this significant risk contributor is evaluated.

5. The LAR identifies the use of the fire protection system to maintain flow from the charging pumps to the reactor coolant pump seals, using dedicated operators and non-seismic connections. The NRC staff's understanding of the current guidance is that seal cooling cannot be restored after about 10 - 15 minutes of interruption, due to seal heatup and

-3 concerns for thermal shock of the seals. Provide a more detailed discussion of this action as to the duration of the action, and impact on interruption of reactor coolant pump seal cooling and any relevant assumptions made regarding seal performance upon restoration of cooling.

6. Throughout the submittal, reference is made to RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," July 1998. However, the current version of RG 1.174 is dated November 2002. Explain why the current version of RG 1.174 was not used as a basis for this submittal.

7. NUREG-0800, Standard Review Plan (SRP) 16.1, Revision 1, "Risk-Informed Decision Making: Technical Specifications," describes NRC staff acceptance criteria regarding "defense in depth," considerations for TS changes. Criterion 11.1A.(iii)(3) states that the licensee should consider, "Whether the TS change specifies that voluntary removal of equipment from service should not be scheduled when adverse weather conditions or other situations that likely may subject the plant to abnormal conditions are predicted."

The licensee has considered adverse weather conditions in its PRA analysis for external events, but has made no specified provision for adverse weather in its TS change in accordance with the guidelines of SRP 16.1, if adverse weather actually occurs.

Explain why the TS change (specifically, in Attachment 1, Table 3) does not specify that the removal of 1/2SX001A from service will not be scheduled when adverse weather conditions or other situations that likely may subject the plant to abnormal conditions are predicted. The NRC staff notes that implementation of Attachment 1, Table 3 is a regulatory commitment per Attachment 4 of the LAR.

8. The licensee has stated, in Attachment 5 of the LAR, that defense in depth is maintained, in part, because a best estimate flow analysis has shown that a single SX pump can provide cooling on both units with the exception of the reactor containment fan coolers (RCFC) and emergency diesel generators (EDGs) on the unit without an SX pump and one train of RCFCs on the unit with an available SX pump.

a. If, during the one-time extension of the CT, an SX pump is lost, describe the sequence of events and operator actions that will bring both units to safe shutdown.

b. If the SX pump of the operating unit were lost, all RCFCs in the operating unit would lose SX flow. With no RCFCs available in the operating unit, explain why the other means of containment heat removal, i.e., the containment spray system of the operating unit, are not on the protected system list during the extended CT.

c. If the SX pump of the operating unit were lost, both EDGs of the operating unit would be lost. Describe the feasibility of making provisions for temporary use of fire protection water, or another available water source, to supply at least one EDG of the unit without an SX pump during the extended CT and subsequent shutdown.

d. Provide piping and instrumentation drawings for the SX system from the units' current licensing basis.

ML093200065),

Exelon Generation Company, LLC (the licensee), submitted a license amendment request proposing a one-time extension of the Completion Time to restore a unit-specific essential service water train to operable status for Technical Specification Limiting Condition for Operation 3.7.8, "Essential Service Water (SX) System," from 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> to 144 hours0.00167 days <br />0.04 hours <br />2.380952e-4 weeks <br />5.4792e-5 months <br />.

The NRC staff is reviewing your submittals, and has determined that additional information is required to complete its review. The specific information requested is addressed in the enclosed Request for Additional Information (RAI). The RAI was discussed with your staff on December 17, 2009, and they agreed to respond within 30 days after the date of this letter.

The NRC staff considers that timely responses to requests for additional information help ensure sufficient time is available for staff review and contribute toward the NRC's goal of efficient and effective use of staff resources. If circumstances result in the need to revise the requested response date, please contact me at (301) 415-1547.

Sincerely, IRA!

Marshall J. David, Senior Project Manager Plant Licensing Branch 111-2 Division of Operating Reactor Licensing Office of Nuclear Reactor Regulation Docket Nos. STN 50-454 and STN 50-455

Enclosure:

Request for Additional Information cc w/encls: Distribution via Listserv DISTRIBUTION:

PUBLIC RidsRgn3MailCenter Resource RidsOgcRp Resource LPL3-2 R/F RidsNrrDssSbpb Resource GPurciarello, NRR RidsNrrDorlLpl3-2 Resource RidsNrrDraApla Resource Andrew Howe, NRR RidsNrrLATHarris Resource RidsNrrDirslhpb Resource KMiller, NRR RidsNrrPMByron Resource RidsNrrDirsltsb Resource GLapinsky, NRR RidsAcrsAcnw_MailCTR Resource RidsNrrDeEeeb Resource MHamm, NRR RidsNrrDorlDpr Resource ADAMS Accession No' .. ML093200660

• RAI Memo Date NRR-088 OFFICE LPL3-2/PM LPL3-2/LA SBPB/BC APLAlBC LPL3-2/BC NAME MDavid THarris GCasto* DHarrison* CGratton for SCampbell DATE 12/17/09 12/17/09 12/11/09 12/15/09 12/18/09