ML15280A114
| ML15280A114 | |
| Person / Time | |
|---|---|
| Site: | Arkansas Nuclear  |
| Issue date: | 10/06/2015 |
| From: | Andrea George Plant Licensing Branch IV |
| To: | David Bice Entergy Nuclear Operations |
| References | |
| TAC MF3419 | |
| Download: ML15280A114 (6) | |
Text
1NRR-PMDAPEm ResourceFrom:George, AndreaSent:Tuesday, October 06, 2015 3:45 PMTo:BICE, DAVID B (ANO)Cc:CLARK, ROBERT W; Miller, Barry
Subject:
Second Round Follow-up Requests for Additional Information - ANO-1 NFPA 805 LAR - TAC No. MF3419 (formal transmittal)Attachments:MF3419 - Second Round RAIs - 2.docxMr. Bice, By letter dated January 29, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14029A438), as supplemented by letters dated May 19, 2015 (ADAMS Accession No. ML15139A196), June 16, 2015 (ADAMS Accession No. ML15167A503), July 21, 2015 (ADAMS Accession No. ML15203A205), August 12, 2015 (ADAMS Accession No. ML15224A729), and September 22, 2015 (ADAMS Accession No. ML15265A113), Entergy Operations, Inc. (the licensee), submitted a license amendment request to transition the Arkansas Nuclear One, Unit 1 (ANO-1), fire protection program to one based on the National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, as incorporated into Title 10 of the Code of Federal Regulations (10 CFR) Section 50.48(c).
In the course of its review, the U.S. Nuclear Regulatory Commission (NRC) staff has determined that additional information is required in order to complete its evaluation. Please treat this email as formal transmittal of the second round follow-up RAIs, which are attached. These RAIs were discussed with you and a number of your staff in a clarification call on October 6, 2015. During that call it was decided that a response would be provided within 30 days, which is November 5, 2015.
If you anticipate any issues with this response date, please communicate your concerns to me at 301-415-1081 or Andrea.George@nrc.gov. Docket No. 50-313
Enclosure:
Second Round Follow-Up RAIs
Sincerely,
Andrea George Project Manager Division of Operating Reactor Licensing U.S. Nuclear Regulatory Commission 301-415-1081
Hearing Identifier: NRR_PMDA Email Number: 2434 Mail Envelope Properties (Andrea.George@nrc.gov20151006154500)
Subject:
Second Round Follow-up Requests for Additional Information - ANO-1 NFPA 805 LAR - TAC No. MF3419 (formal transmittal) Sent Date: 10/6/2015 3:45:23 PM Received Date: 10/6/2015 3:45:00 PM From: George, Andrea Created By: Andrea.George@nrc.gov Recipients: "CLARK, ROBERT W" <RCLARK@entergy.com> Tracking Status: None "Miller, Barry" <Barry.Miller@nrc.gov> Tracking Status: None "BICE, DAVID B (ANO)" <DBICE@entergy.com> Tracking Status: None Post Office: Files Size Date & Time MESSAGE 1791 10/6/2015 3:45:00 PM MF3419 - Second Round RAIs - 2.docx 32240 Options Priority: Standard Return Notification: No Reply Requested: Yes Sensitivity: Normal Expiration Date: Recipients Received:
REQUEST FOR ADDITIONAL INFORMATION RELATED TO LICENSE AMENDMENT REQUEST TO TRANSITION FIRE PROTECTION PROGRAM TO NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805 RENEWED FACILITY OPERATING LICENSE NO. DPR-51 ENTERGY OPERATIONS, INC. ARKANSAS NUCLEAR ONE, UNIT NO. 1 DOCKET NO. 50-313 By letter dated January 29, 2014 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML14029A438), as supplemented by letters dated May 19, 2015 (ADAMS Accession No. ML15139A196), June 16, 2015 (ADAMS Accession No. ML15167A503), July 21, 2015 (ADAMS Accession No. ML15203A205), August 12, 2015 (ADAMS Accession No. ML15224A729), and September 22, 2015 (ADAMS Accession No. ML15265A113), Entergy Operations, Inc. (the licensee), submitted a license amendment request (LAR) to transition the Arkansas Nuclear One, Unit 1 (ANO-1), fire protection program to one based on the National Fire Protection Association Standard 805 (NFPA 805), "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, as incorporated into Title 10 of the Code of Federal Regulations (10 CFR) Section 50.48(c). In the course of its review, the U.S. Nuclear Regulatory Commission (NRC) staff has determined that additional information is required in order to complete its evaluation, as stated below.
Probabilistic Risk Assessment (PRA) Request for Additional Information (RAI) 03.b.01 - Integrated Analysis
The response to PRA RAI 03.b should provide a summary of how each issue identified in PRA RAI 03 was resolved for the integrated analysis. The Table in the response to RAI 03.b submitted on August, 12, 2015, includes a number of entries with statements such as "is being revised" and "will be consistent." Additionally, based on PRA RAI responses received after PRA RAI 03 was issued on May 5, 2015, the NRC staff has identified two additional issues whose resolution should be confirmed in the Table. Please up-date and resubmit the Table with the final action taken on each issue, and add the following two issues to the Table. a) PRA RAI 01.e.d regarding spurious operation in other cable configurations.
The response to PRA RAI 01.e.d submitted on July 21, 2015, explains that "other cable configurations" (as defined by Section 7.4 of NUREG/CR-7150) will be evaluated using guidance in NUREG/CR-7150 in the integrated analysis provided in response to PRA RAI 03. The list of issues in PRA RAI 03 did not include this issue because it was unknown at the time. Please add "PRA RAI 01.e.d regarding spurious operation in other cable configurations" to the summary table and confirm that "other cable configurations" have been evaluated using guidance in NUREG/CR-7150 in the integrated analysis provided in response to PRA RAI 03. b) PRA RAI 02.b regarding completion of the large early release frequency (LERF) analysis. The response to PRA RAI 02.b identified two LERF analysis limitations associated with modeling of the atmospheric dump valves (ADVs) and the reactor coolant system electromagnetic relief valve (ERV) and explained these analysis limitations will be resolved in the integrated analysis provided in response to PRA RAI 03. The list of issues in PRA RAI 03 did not include this issue because it was unknown at the time.
Please add "PRA RAI 02.b regarding completion of LERF analysis" to the summary table and confirm that modelling of the ADVs and ERV has been included in the integrated analysis provided in response to PRA RAI 03. PRA RAI 11.01 - Main Control Room (MCR) Abandonment Due to Loss of Control (LOC)
In the LAR supplement dated August 12, 2015, the response to PRA RAI 11 states, in part, that "[in] all non-loss of habitability cases, command and control for post fire shutdown is expected to remain in the MCR." The response further clarifies that all fire scenarios which do not cause loss-of-habitability are modeled based on the equipment lost in each scenario. As clarified in Regulatory Guide 1.205, when command and control remains in the MCR, all operator actions to mitigate fire-induced failures outside of the MCR are recovery actions. Please confirm that actions to mitigate fire-induced failures outside of the MCR are recovery actions in the variant plant and are included in the additional risk of recovery action estimates. PRA RAI 18.01 - Minimum Joint HEPs In the LAR supplement dated August 12, 2015, the response explains that the Fire PRA (FPRA) will be re-quantified using joint Human Error Probabilities (HEPs) greater than or equal to 1.0E-5, but that depending on the results, values less than 1E-05 may be introduced back into the FPRA model. The response explains, that in this case, the basis for these minimum joint HEPs, which are primarily related to separation in time, shall be justified and the results documented in the Fire Human Reliability Analysis notebook. If minimum joint HEPs less than 1E-05 are introduced back into the Fire PRA then per the request in PRA RAI 18:
a) Confirm that each joint HEP value used in the Fire PRA below 1.0E-05 includes its own justification that demonstrates the inapplicability of the NUREG-1792 lower value guideline. b) Provide an estimate of the number of these joint HEPs below 1.0E-05 and at least two different examples of the justification. Safe Shutdown Analysis (SSA) RAI 11.01 In its response to SSA RAI 11.a in the LAR supplement dated August 12, 2015, the licensee stated that the modifications to address the circuit failure concerns in Information Notice (IN) 92-18, "Potential for Loss of Remote Shutdown Capability During a Control Room Fire," for certain valves, were not credited in areas that use the deterministic approach of NFPA 805 Section 4.2.3. However, in LAR Attachment S, the licensee stated that these modifications will reduce risk because it will preclude the spurious operation; therefore, these modifications are credited to deterministically prevent any spurious operation in a performance-based analysis. The licensee further stated in its response to SSA RAI 11.b, that the installation of an inhibit circuit uses conductors and a control switch to intentionally short across the target coil when the device is in its de-energized state and will prevent intra-cable (internal) and inter-cable (external) hot shorts from energizing the target coil and causing the valve to spuriously actuate, and that this inhibit circuit could be bypassed by manual operation of the "break-before-make" control switch. i. In SSA RAI 11.c, the NRC staff requested that the licensee discuss the effects of hot shorts (external and internal), open circuits, and shorts-to-ground on the circuits and include a discussion on the potential for "collateral damage" from open circuits in adjacent cables that have fusing greater than 10 amps based on the results of the DESIREE-Fire test documented in NUREG/CR-7100, "Direct Current Electrical Shorting in Response to Exposure Fire." However, the response to SSA RAI 11.c did not provide all of the requested information. Rather, the licensee provided a qualitative analysis stating that the failure of the inhibit circuit to result in a spurious operation is unlikely to occur from an energized conductor (either intra-cable or inter-cable), without discussing the effects of circuit damage within the zone of influence due to significant arcing of adjacent cables that have fusing greater than 10 amps. Without the requested information, the NRC staff is not able to conclude if spurious operation is unlikely.
Provide the information requested in SSA RAI 11.c (i.e., discuss the effects of hot shorts, open circuits and shorts-to-ground from adjacent circuits and the potential for "collateral damage" from arcing in adjacent cables with fusing greater than 10 amps), or provide further justification for concluding that failure of the inhibit switch circuit is unlikely, in light of the results of the DESIREE-Fire test. If a probabilistic approach (either qualitative or quantitative) is utilized, then provide the bases for acceptability of the approach when used in the performance-based analysis considering that the Fire PRA credits the modification to "preclude the spurious operation." ii. In SSA RAI 11.d, the NRC staff requested the licensee to describe how fire damage to the new shorting switch in the control room will not affect the desired nuclear safety function. In its response to SSA RAI 11.d, the licensee stated that the postulated damage to the shorting switch would not be sufficient to affect the nuclear safety function based on the fire testing performed and documented in Exelon Nuclear Evaluation EC-EV A1831999 01, "Evaluation of Shorting Switch Modification," Rev. 0, dated December 2011, and that a peak temperature of 575ûF (300ûC) was maintained for 15 minutes. It does not appear to the NRC staff that with the peak temperature reached, the assumptions used to assess the shorting switch are consistent with those used to evaluate MCR fires (for either control room abandonment through loss of habitability or component damage) in accordance with NUREG\CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities." Discuss how the results of this Exelon fire test bound the fire modeling parameters defined in NUREG/CR-6850 (i.e., heat release rates, rate of fire growth, non-suppression probability, etc.), and discuss how the integrity of the shorting switch is affected if the fire modeling parameters defined in NUREG/CR-6850 are considered.