ML19088A009
| ML19088A009 | |
| Person / Time | |
|---|---|
| Site: | Hatch  |
| Issue date: | 03/29/2019 |
| From: | John Lamb Plant Licensing Branch II |
| To: | Burns P, Lowery K Southern Nuclear Operating Co |
| References | |
| L-2018-LLA-0107 | |
| Download: ML19088A009 (26) | |
Text
NRR-DMPSPEm Resource From: Lamb, John Sent: Friday, March 29, 2019 7:02 AM To: Lowery, Ken G.; Burns, Pamela Diane
Subject:
Hatch 1 and 2 - LAR to Adopt NFPA 805 Fire Protection Standard, EPID L-2018-LLA-0107, Request for Additional Information Attachments: Hatch 805 RAIs 03202019.docx Importance: High Mr. Lowry and Ms. Burns:
By application dated April 4, 2018 (Agencywide Documents Access and Management System (ADAMS)
Accession No. ML18096A936), Southern Nuclear Operating Company (SNC) submitted a license amendment request (LAR) for the Edwin I. Hatch Nuclear Plant (HNP), Unit Nos. 1 and 2, and proposed to adopt National Fire Protection Association Standard 805 (NFPA 805), Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition (ADAMS Accession No. ML010800360), as incorporated into Title 10 of the Code of Federal Regulations, Part 50, Section 50.48(c).
To support the review of the proposed license amendment, an audit team consisting of U.S. Nuclear Regulatory Commission (NRC) staff from the Office of Nuclear Reactor Regulation conducted a regulatory audit at the HNP site from March 18 to March 21, 2019. Prior to the audit, on January 23, 2019 the NRC sent the audit plan and draft requests for additional information (RAIs) to the licensee (ADAMS Accession Nos.
ML19029B339 and ML19029B340).
As a result of the audit, the NRC staff has finalized the RAIs (see attached). The RAIs were discussed at length with the licensee during the audit and only one new RAI was developed that was not included with the draft RAIs.
SNC agreed to respond to the RAIs within 60 days from the date of this email.
Thanks.
John 1
Hearing Identifier: NRR_DMPS Email Number: 893 Mail Envelope Properties (BL0PR0901MB3731202DD3C73EA78E1B6751FA5A0)
Subject:
Hatch 1 and 2 - LAR to Adopt NFPA 805 Fire Protection Standard, EPID L-2018-LLA-0107, Request for Additional Information Sent Date: 3/29/2019 7:02:06 AM Received Date: 3/29/2019 7:02:00 AM From: Lamb, John Created By: John.Lamb@nrc.gov Recipients:
"Lowery, Ken G." <KGLOWERY@southernco.com>
Tracking Status: None "Burns, Pamela Diane" <PDBURNS@SOUTHERNCO.COM>
Tracking Status: None Post Office: BL0PR0901MB3731.namprd09.prod.outlook.com Files Size Date & Time MESSAGE 1404 3/29/2019 7:02:00 AM Hatch 805 RAIs 03202019.docx 83146 Options Priority: High Return Notification: No Reply Requested: No Sensitivity: Normal Expiration Date:
Recipients Received:
SOUTHERN NUCLEAR OPERATING COMPANY, EDWIN I. HATCH NUCLEAR PLANT, UNIT NOS. 1 AND 2 REQUEST FOR ADDITIONAL INFORMATION FOR THE LICENSE AMENDMENT REQUEST TO IMPLEMENT A RISK-INFORMED, PERFORMANCE-BASED, FIRE PROTECTION PROGRAM AS ALLOWED BY TITLE 10 OF THE CODE OF FEDERAL REGULATIONS, PARAGRAPH 50.48(c) (EPID L-2018-LLA-0107)
NFPA 805 Fire Modeling (FM) Request for Additional Information (RAI) 01 Section 2.4.3.3, Fire Risk Evaluations, of National Fire Protection Association (NFPA)
Standard 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants, 2001 Edition (NFPA 805) states: [t]he PSA [probabilistic safety assessment] approach, methods, and data shall be acceptable to the AHJ [authority having jurisdiction]...
Based on information provided by the licensee, the Nuclear Regulatory Commission (NRC) staff determined that FM is comprised of the following:
- A plant-specific Fire Modeling Workbook (FMWB) that was developed in lieu of using NUREG-1805, Fire Dynamics Tools (FDTs) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program (ADAMS Accession No. ML043290075), or Electric Power Research Institute (EPRI)
Fire Induced Vulnerability Evaluation Methodology, Revision 1 (FIVE Rev. 1), to determine the Zone of Influence (ZOI) for ignition sources and the time to Hot Gas Layer (HGL) conditions in all fire areas throughout the plant.
- Heskestad's plume temperature correlation which was used to determine the vertical separation distance based on temperature to a target in order to determine the vertical extent of the ZOI.
- The Consolidated Fire Growth and Smoke Transport (CFAST) model which was used to assess main control room (MCR) abandonment time and multi-compartment analysis (MCA).
- FLASH-CAT which was used for calculating fire propagation in stacks of horizontal cable trays.
- Heat soak method which was used for evaluating time to cable damage.
License Amendment Request (LAR) Section 4.5.1.2, Fire PRA, states that FM was performed as part of the Fire PRA (FPRA) development (NFPA 805-Section 4.2.4.2) and reference is made to LAR Attachment J, Fire Modeling Verification and Validation, for a discussion of the acceptability of fire models that were used to develop the FPRA. Based on the information in
the LAR, the NRC staff was unable to fully evaluate the FM performed as part of the FPRA and requests that the licensee:
(a) Regarding fires in the proximity of a corner or walls, explain how the FM approach was applied. Explain how wall and corner affects the ZOI and HGL timing calculations were accounted for, or provide technical justification if these effects were not considered.
Explain how transient fires against a wall or in a corner were considered in the MCR abandonment calculations.
(b) The NRC staff finds that typically, during maintenance or measurement activities in the plant, electrical cabinet doors remain open for a certain period of time. Describe whether there are any administrative controls in place to minimize the likelihood of fires involving such a cabinet, and describe how cabinets with temporarily open doors are treated.
(c) Describe and provide technical justification for the approach that was used in the FLASH-CAT model to determine the time to ignition, the heat release rate per unit area (HRRPUA), and the flame spread rate for cable trays that contain a mixture of thermoplastic and thermoset cables.
(d) Describe the Heat Soak Method that was used to convert the damage times in Appendix H of NUREG/CR-6850 to a percent of damage function for targets exposed to a time-varying heat flux.
(e) Describe how non-cable secondary combustibles were identified and accounted for.
(f) Explain how the model assumptions in terms of location and HRR of transient combustibles in a fire area or zone will not be violated during and post-transition.
(g) Describe how high energy arcing fault (HEAF) initiated fires are treated in the HGL development timing. Regarding HEAF generated fires, describe the criteria used to decide whether a cable tray in the vicinity of an electrical cabinet will ignite following a HEAF event in the cabinet. Explain how the ignited area was determined and subsequent fire propagation calculated. Describe the effect of cable tray covers and fire-resistant wraps on HEAF induced cable tray ignition and subsequent fire propagation.
(h) Facts and Observations (F&O) 20-19, Structural Steel Scenario Selection, was generated and dispositioned with the following:
The F&O identifies a statement in report H-RIEFIREPRA- U00-008D, Specific ignition sources proximate to or with direct impingement on exposed structural steel were not evaluated. The purpose of this statement is to distinguish that detailed analytical heat transfer and structural analysis of steel members was not performed.
As stated earlier in the report, the following criteria are used to develop structural steel scenarios, which makes the statement identified in the F&O unnecessary:
a) Exposed structural steel is present and, b) A high-hazard fire source is present.
Following these criteria, the scenarios developed are inherently more conservative than an analysis that relies on detailed fire modeling and analytical heat transfer modeling of individual structural steel members.
The F&O has been resolved by deleting the statement in question.
Describe and provide technical justification for the approach that make structural steel scenarios more conservative.
(i) Regarding the Main Control Abandonment (MCA), describe how the size of the opening between the exposing and exposed compartments assumed in the CFAST HGL calculations was determined, and explain to what extent these vent sizes are representative of conditions in the plant.
(j) Regarding the acceptability of CFAST for the control room abandonment time study, describe whether the volumes of the main control boards (MCBs), electrical panels, raised platforms, ductwork in the interstitial space above the egg-crate ceiling, and other obstructions are excluded from the effective control room volume used in the CFAST calculations.
(k) Because the Main Control Room (MCR) abandonment calculations are based on the assumption that all doors would normally remain closed, describe if any natural leakage vents were assumed in the analysis.
NFPA 805 FM RAI 02 Section 2.5, Evaluating the Damage Threshold, of NFPA 805, requires damage thresholds be established to support the performance-based approach. Thermal impact(s) must be considered in determining the potential for thermal damage of structures, systems, and components (SSCs). Appropriate temperature and critical heat flux criteria must be used in the analysis.
American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) Standard
- RA-Sa-2009, "Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessments for Nuclear Power Plant Applications," Part 4, indicates that damage thresholds be established to support the FPRA. The standard further indicates that thermal impact(s) must be considered in determining the potential for thermal damage of SSCs and appropriate temperature and critical heat flux criteria must be used in the analysis.
Unit 1 uses thermoplastic and thermoset cables, but due to insufficient cable material data, the FPRA assumes all cables are thermoplastic material, unless a definitive determination could be made that a cables insulation is thermoset or equivalent to thermoset based on cable material codes contained in the Plant Data Management System (PDMS) or in vendor supplied data.
Unit 2 uses all thermoset cables.
Provide the following information:
(a) For Unit 1, explain how raceways with a mixture of thermoplastic and thermoset cables are treated in terms of damage thresholds.
(b) For Unit 2 assumed to have thermoset damage criteria, confirm that the cables are actually thermoset and not just qualified by the Institute of Electrical and Electronics Engineers (IEEE) Standard 383, IEEE Standard for Qualifying Electric Cables and Splices for Nuclear Facilities.
(c) Explain how the damage thresholds for non-cable components (i.e., pumps, valves, electrical cabinets, etc.) were determined. Identify any non-cable components that were assigned damage thresholds different from those for thermoset and thermoplastic cables.
NFPA 805 FM RAI 03 Section 2.7.3.2, Verification and Validation, of NFPA 805, states that each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models.
LAR Section 4.5.1.2, Fire PRA states that FM was performed as part of the FPRA development (NFPA 805 Section 4.2.4.2). The LAR further states that the acceptability of the use of these fire models is included in Attachment J.
LAR Section 4.7.3, Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 states that calculational models and numerical methods used in support of compliance with 10 CFR 50.48(c) were verified and validated as required by Section 2.7.3.2 of NFPA 805.
Based on the information provided in the LAR, the NRC staff was unable to confirm whether each calculational model or numerical method was properly verified and validated, therefore, the NRC staff requests that the licensee:
(a) Describe how the Fire Modeling Workbook was verified, and describe how it was ensured that the empirical equations/correlations were coded correctly and that the solutions are identical to those that would be obtained with the corresponding chapters in the NUREG-1805 (FDTs) or FIVE-Rev.1.
(b) For any FM tool or method that was used in the development of the LAR, provide the Verification and Validation (V&V) basis if it is not already explicitly provided in the LAR Attachment J. Further, identify any applications of FM tools or methods used in the development of the LAR that are not discussed in LAR Attachment J.
(c) LAR Attachment J states that the smoke detection actuation correlation (Method of Heskestad and Delichatsios) has been applied within the validated range reported in NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications" (ADAMS Accession No. ML071650546). However, the latter reports a validation range only for Alpert's ceiling jet temperatures correlation. Provide technical details to demonstrate that the temperature to smoke density correlation has been applied within the validated range, or justify the application of the correlation outside the validated range reported in the V&V basis documents.
NFPA 805 FM RAI 04 Section 2.7.3.3, Limitations of Use, of NFPA 805, states, acceptable engineering methods and numerical models shall only be used for applications to the extent these methods have
been subject to verification and validation. These engineering methods shall only be applied within the scope, limitations, and assumptions prescribed for that method.
LAR Section 4.7.3, Compliance with Quality Requirements in Section 2.7.3 of NFPA 805, states that engineering methods and numerical models used in support of compliance with 10 CFR 50.48(c) were applied appropriately as required by Section 2.7.3.3 of NFPA 805.
Based on the information provided in the LAR, the NRC staff was unable to determine whether the engineering methods used were applied within the scope, limitations, and assumptions prescribed for those methods, therefore, the NRC staff requests that the licensee identify uses, if any, of the FMWBs outside the limits of applicability of the method and for those cases outside of the limits, explain the analysis that was used or why the use of the FMWBs was justified.
NFPA 805 FM RAI 05 Section 2.7.3.4, Qualification of Users, of NFPA 805, states, cognizant personnel who use and apply engineering analysis and numerical models (e.g., FM techniques) shall be competent in that field and experienced in the application of these methods as they relate to nuclear power plants, nuclear power plant fire protection, and power plant operations.
LAR Section 4.5.1.2, Fire PRA states that FM was performed as part of the FPRA development (Section 4.2.4.2 of NFPA 805). This requires that qualified FM and PRA personnel work together. Furthermore, LAR Section 4.7.3, Compliance with Quality Requirements in Section 2.7.3 of NFPA 805, states:
Cognizant personnel who use and apply engineering analysis and numerical methods in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by Section 2.7.3.4 of NFPA 805.
During the transition to 10 CFR 50.48(c), work was performed in accordance with the quality requirements of Section 2.7.3 of NFPA 805. Personnel who used and applied engineering analysis and numerical methods (e.g. FM) in support of compliance with 10 CFR 50.48(c) are competent and experienced as required by NFPA 805 Section 2.7.3.4.
Post-transition, for personnel performing FM or FPRA development and evaluation, SNC will develop and maintain qualification requirements for individuals assigned various tasks. Position Specific Guides will be developed to identify and document required training and mentoring to ensure individuals are appropriately qualified per the requirements of NFPA 805 Section 2.7.3.4 to perform assigned work. See Attachment S, Table S-3, Implementation Item IMP-8.
Based on the information provided in the LAR, the NRC staff was unable to determine how the licensee complies with NFPA 805, Section 2.7.3.4, therefore, the NRC staff requests that the licensee:
- a. Describe what constitutes the appropriate qualifications for the staff and consulting engineers to use and apply the methods and FM tools included in the engineering analyses and numerical models.
- b. Describe the processes/procedures for ensuring the adequacy of the appropriate qualifications of the engineers/personnel performing the fire analyses and modeling activities.
- c. Provide the position(s) and qualifications of the personnel who performed the walkdowns for the MCR (abandonment based on damage and inhabitability) and the remaining fire areas in the plant. Address whether the same people who performed walkdowns conduct the FM analysis.
- d. Explain the communication process between the FM analysts and PRA personnel to exchange the necessary information and any measures taken to assure the FM was performed adequately and will continue to be performed adequately during post-transition.
- e. Explain the communication process between the consulting engineers and plant and corporate personnel to exchange the necessary information. Describe measures taken to assure the FM was performed adequately and will continue to be performed adequately during post-transition.
NFPA 805 FM RAI 06 NFPA 805, Section 2.7.3.5, Uncertainty Analysis, states, An uncertainty analysis shall be performed to provide reasonable assurance that the performance criteria have been met.
LAR Section 4.7.3, Compliance with Quality Requirements in Section 2.7.3 of NFPA 805, states that Uncertainty analyses were performed as required by 2.7.3.5 of NFPA 805 and the results were considered in the context of the application. This is of particular interest in FM and FPRA development."
Based on the information provided in the LAR, the NRC staff was unable to determine how the licensee complies with NFPA 805, Section 2.7.3.5, therefore, the NRC staff requests for the uncertainty analysis for FM, that the licensee:
- a. Describe how the uncertainty associated with the fire model input parameters was accounted for in the FM analyses.
- b. Describe how the model uncertainty was accounted for in the FM analyses.
- c. Describe how the completeness uncertainty was accounted for in the FM analyses.
NFPA 805 Safe Shutdown (SSD) RAI 01 NFPA 805, Section 2.4.2, "Nuclear Safety Capability Assessment," requires licensees to perform a nuclear safety capability assessment (NSCA). RG 1.205, NEI 00-01, Chapter 3, as one acceptable approach to perform an NSCA. Nuclear Energy Institute, (NEI) 00-01, Guidance for Post-Fire Safe Shutdown Circuit Analysis, (ADAMS Accession No. ML091770265) Section 3.5.2, indicates that with respect to the electrical distribution system, the issue of breaker coordination must also be addressed.
The licensees cable selection and circuit failure analysis indicates that some devices may not be coordinated or coordination may be undetermined but will be addressed through procedures.
Since the probabilistic risk assessment (PRA) treats all credited power supplies as having proper electrical coordination:
- a. Discuss whether a comprehensive electrical coordination study for the credited power supplies has been completed and whether all issues have been identified and resolved.
If not, provide a proposed path forward to resolve the outstanding issues.
- b. Discuss any outstanding issues which should be considered for inclusion in LAR Attachment S, as modifications or implementation items as necessary.
NFPA 805 SSD RAI 02 NFPA 805 Section 3.11.3, Fire Barrier Penetrations, indicates that penetrations in fire barriers shall be provided with listed fire-rated door assemblies or listed rated fire dampers having a fire resistance rating consistent with the designated fire resistance rating of the barrier as determined by the performance requirements established by NFPA 805 Chapter 4, and that passive fire protection devices such as doors and dampers shall conform with NFPA 80, Standard for Fire Doors and Fire Windows.
In order to meet the requirements of NFPA 80, the licensee has proposed a plant modification to relocate or install fusible links on certain sliding fire doors, as described in LAR Attachment S, Table S-2, Modification Item 1. The licensee indicates that this modification also addresses the potential of water intrusion into Switchgear 1R23S004 located in Fire Area 1017 from fire suppression activities in Fire Zone 0014K. For a postulated fire in Fire Zone 0014, discuss:
- a. The actuation time of the fusible link and the suppression system, and whether the fusible link actuated sliding fire door will close in a timely manner as to preclude potential water damage to Switchgear 1R23S004 from fire suppression activities.
- b. Since sliding fire doors are not watertight doors, discuss the impact, if any, on Switchgear 1R23S004 due to water migration into Fire Area 1017.
Describe how this modification ensures Switchgear 1R23S004 will be protected.
NFPA 805 SSD RAI 03 NFPA 805 Section 4.2.1 requires one success path necessary to achieve and maintain the nuclear safety performance criteria (NSPC) shall be maintained free of fire damage by a single fire, and that the effects of fire suppression activities on the ability to achieve the NSPC shall be evaluated.
In LAR Attachment C, Table C-1, the discussion of fire suppression effects in many fire areas includes water from some deluge or sprinkler systems and from hose streams might temporarily exceed the capacity of the drain system in some areas. However, safety related equipment is elevated above the floor level by pads or pedestals, such that equipment is protected from flooding.
Based on the information provided in the LAR, the NRC staff was unable to determine whether the effects of fire suppression activities on the ability to achieve the NSPC have been properly evaluated for areas where flooding is a concern and pads and pedestals are credited to protect SSD equipment.
Provide a summary of the internal flooding analysis that demonstrates the raised pads and pedestals are adequate in height.
NFPA 805 Health Physics (HP) RAI 01 LAR Attachment E, Radioactive Release Transition states that the potential release of contaminated effluents resulting from a fire involving radioactive contents in the Bounded Areas compartment is bounded by Vendor Document S77684. In addition, Attachment E states, This calculation demonstrates that releases are below 10 CFR 20 limits and satisfies the acceptance criteria of FAQ 09-0056. For Vendor Document S77684, please provide a summary of the assumptions, methodology, input parameters, resulting doses and conclusions.
NFPA 805 HP RAI 02 To meet the radioactive release performance criteria for NFPA 805, licensees must demonstrate that radiation released to any unrestricted area due to the direct effects of fire suppression activities remains as low as is reasonably achievable (ALARA), not to exceed the limits in 10 CFR Part 20.
The NRC staff noted that the licensee has performed a bounding analyses to demonstrate that the doses from the airborne and liquid pathways resulting from fire suppression activities will not exceed the limits of 10 CFR Part 20. In the licensees analysis, the calculated bounding doses are provided in terms of total effective dose equivalent (TEDE), which is consistent with the limits specified in 10 CFR Part 20. The limits in 10 CFR Part 20 are specified in terms of TEDE because the regulations in 10 CFR Part 20 are based on the International Commission on Radiation Protections (ICRP) recommendations in ICRP Reports 26 and 30. However, when using Radiological Assessment System for Consequence Anlysis (RASCAL) to perform the bounding analysis, the licensee chose to use ICRP 60/72 inhalation dose coefficients. In addition, when HotSpot Version 3.0.2 was used, the licensee selected the Federal Guidance Report (FGR) No. 13 dose conversion factors (DCFs). As a result, both the RASCAL and HotSpot calculations provided doses in terms of total effective dose (TED). Likewise, the use of the ICRP 60/119 DCFs for the liquid pathway calculations also resulted in doses in terms of TED. Nevertheless, the results provided in the licensees conclusions were provided in terms of TEDE. While both TEDE and TED calculate dose for external and internal exposure, the underlying dosimetry models used to develop the DCFs are not the same. The DCFs selected for the gaseous and liquid bounding analyses results in the use of dosimetry models and DCFs that differ from those used in ICRP Reports 26 and 30. Dose conversion factors acceptable to the NRC staff are derived from data and methodologies provide in ICRP Publication 30, Limits for Intakes of Radionuclides by Workers and can be found in FGR No. 11, Limiting Values of Radionuclide Intake and Air Concentration and Dose Conversion Factors for Inhalation, Submersion, and Ingestion, and FGR No. 12, External Exposure to Radionuclides in Air, Water, and Soil, for exposure to radionuclides in air, water, and soil.
Please provide a summary explaining why the use of the TED DCFs is acceptable, even though the dose limits in 10 CFR Part 20 are specified in terms of TEDE.
NFPA 805 Probabilistic Risk Assessment (PRA) RAI 01 - Fire PRA F&O Closure Review Process
NFPA 805 Section 2.4.3.3 states that the PRA approach, methods, and data shall be acceptable to the NRC. Regulatory Guide (RG) 1.200, An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities, Revision 2, (ADAMS Accession No. ML090410014), describes a peer review process utilizing an associated American Society of Mechanical Engineers/American Nuclear Society (ASME/ANS) PRA standard (currently ASME/ANS-RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Level 1/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications," Februrary 2, 2009) as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established. The primary results of a peer review are the Facts and Observations (F&Os) recorded by the peer review and the subsequent resolution of these F&Os. In a letter dated May 3, 2017 (ADAMS Accession No. ML17079A427) the NRC staff has accepted, with conditions, a final version of Appendix X to Nuclear Energy Institute (NEI) 05-04, Process for Performing Internal Events PRA Peer Reviews Using the ASME/ANS PRA Standard, NEI 07-12, Fire Probabilistic Risk Assessment (FPRA) Peer Review Process Guidelines, and NEI 12-13, External Hazards PRA Peer Review Process Guidelines, (ADAMS Accession No. ML17086A431), which defines a review process for closing finding-level F&Os.
LAR Attachment U and LAR Attachment V state that an F&O independent assessment (IA) was performed on the FPRA peer review results to close finding-level F&Os using the process documented in Appendix X to NEI 07-12.
Based on the information provided in the LAR, the NRC staff was unable to determine if the F&O closure reviews were performed fully consistent with the NRC accepted process described above, therefore, the NRC staff requests that the licensee:
a) Describe the process used to determine whether a change to the PRA is maintenance or an upgrade. Describe the actions taken or internal processes applied to ensure the robustness of your determination.
b) Confirm, for each FPRA F&O resolved, whether the resolution was determined to be a PRA upgrade or maintenance update. Include discussion of how the guidance in Appendix 1-A of ASME/ANS RA-Sa-2009 was used in the basis of each determination.
Discuss any changes made to your initial assignment of maintenance or upgrade through deliberations with the IA team.
c) If the request in part (b) above cannot be confirmed based on the current F&O closure review documentation, then provide for each finding-level F&O an indication of whether the resolution was determined to be a PRA upgrade or maintenance update along with the specific bases for those determinations as reviewed by the independent assessment (IA) team.
NFPA 805 PRA RAI 02 - Incorporation of Internal Events PRA Updates into the FPRA Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. It appears to the NRC staff that a number of internal event (IEPRA) model additions and revisions were made to resolve F&Os that were then closed in the April 2017 IE F&O closure review. The NRC staff notes that the F&O closure review for the IEPRA was followed closely by a FPRA F&O closure review in October/November 2017. It appears to
the NRC staff that a number of updates were made to the IE PRA to resolve F&Os that are relevant to the FPRAs underlying plant response model.
Therefore, the NRC staff requests that the licensee:
a) Confirm that applicable IE PRA model updates that were performed to resolve finding-level F&Os ahead of the IE F&O closure review were also performed for the FPRA model used to determine the fire risk estimates for the NFPA 805 LAR.
b) If the IEPRA model updates that were performed to resolve finding-level F&Os were not also performed for the FPRA model used to determine the fire risk estimates for the NFPA 805 LAR, then justify that these model updates have no impact on NFPA 805 LAR application. Alternatively, perform these updates for the integrated analysis provided in response PRA RAI 03.
NFPA 805 PRA RAI 03 - Integrated Analysis Section 2.4.4.1 of NFPA 805 states that the change in public health risk arising from transition from the current fire protection program (FPP) to an NFPA 805 based program, and all future plant changes to the program, shall be acceptable to the NRC. RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," Revision 2, (ADAMS Accession No. ML100910006), provides quantitative guidelines on core damage frequency (CDF), and large early release frequency (LERF), and identifies acceptable changes to these frequencies that result from proposed changes to the plants licensing basis and describes a general framework to determine the acceptability of risk-informed changes.
Based on other NRC staff RAIs, the PRA methods discussed in the following RAIs may need to be revised to be acceptable to the NRC:
- PRA RAI 16.d regarding the impact of other modeling conservatisms on change-in-risk This list may be revised following the NRC review of the licensees response to all the RAIs (not just those listed here).
a) Provide the results of an aggregate analysis that provides the integrated impact on the fire risk (i.e., the total transition CDF and LERF, and the change () in CDF (CDF),and LERF, of replacing specific methods identified above with alternative methods which
are acceptable to the NRC. In this aggregate analysis, for those cases where the individual issues have a synergistic impact on the results, a simultaneous analysis must be performed. For those cases where no synergy exists, a one-at-a-time analysis may be done. For those cases that have a negligible impact, a qualitative evaluation may be done.
b) For each method above, explain how the issue will be addressed in (1) the final aggregate analysis results provided in support of the LAR, and (2) the PRA that will be used at the beginning of the self-approval of post-transition changes. In addition, provide a method to ensure that all changes will be made, that a focused-scope peer review will be performed on changes that are PRA upgrades as defined in the PRA standard, and that any findings will be resolved before self-approval of post-transition changes.
c) Explain how the RG 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," Revision 1, (ADAMS Accession No. ML092730314) risk acceptance guidelines are satisfied for the aggregate analysis. If applicable, include a description of any new modifications or operator actions being credited to reduce delta risk as well as a discussion of the associated impacts to the FPP.
d) If any unacceptable methods or weaknesses will be retained in the PRA that will be used to estimate the change-in-risk of post-transition changes to support self-approval, explain how the quantification results for each future change will account for the use of these unacceptable methods or weaknesses.
e) Identify and summarize the changes to the FPRA model beyond those associated with the RAIs cited above that may need to be revised and confirm that the changes do not introduce approaches unacceptable to NRC.
NFPA 805 PRA RAI 04 - Use of Unacceptable Methods LAR Attachment V states that the Hatch full-scope-scope FPRA peer review identified 0 unreviewed analysis methods (UAMs). Though UAMs, as evaluated by the EPRI/NRC panel were not used, this does not preclude the possibility that methods may have been used in the FPRA that deviate from guidance in NUREG/CR-6850, "EPRI/NRC-RES Fire PRA Methodology for Nuclear Power Facilities, (ADAMS Accession Nos. ML052580075, ML052580118, and ML103090242), or other acceptable guidance (e.g., frequently asked questions (FAQs),
NUREGs, or interim guidance documents). Based on the information provided in the LAR, the NRC staff could not determine whether any methods that deviate from NUREG/CR-6850 or other acceptable guidance were used, therefore the NRC staff requests that the licensee:
a) Identify methods used in the FPRA that deviate from guidance in NUREG/CR-6850 or other acceptable guidance.
b) If such deviations exist, then justify their use in the FPRA. Alternatively, replace those methods with a method acceptable to NRC in the integrated analysis performed in response to PRA RAI 03. Include a description of the replacement method along with justification that it is consistent with NRC accepted guidance.
NFPA 805 PRA RAI 05 - Implementation Item to Update Fire PRA When Modification are Complete LAR Attachment S, Table S-3, presents an implementation item (i.e., IMP-19) to update the FPRA after all plant modifications have been implemented to reflect the as-built, as-operated plant.
This implementation item does not indicate HNPs plan in the event that the updated FPRA results do not meet RG 1.174, Revision 2, risk acceptance guidelines. Also, implementation item IMP-19 does not indicate that updates to the FPRA should include adjustments needed to reflect completion of other implementation items such as update of fire response procedures.
Revise implementation item IMP-19 to include an action to update the FPRA following completion of modifications and implementation items and include a plan of action should the updated as-built as-operated FPRA results risk estimates exceed RG 1.174, Revision 2, risk acceptance guidelines (e.g., this plan could include refining the analytic risk estimates or performing additional modifications to the plant).
NFPA 805 PRA RAI 06 - Reduced Transient Heat Release Rates (HRRs)
Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA. The key factors used to justify using transient fire reduced heat release rates (HRRs) below those prescribed in NUREG/CR-6850 are discussed in the June 21, 2012, letter from Joseph Giitter, U.S. Nuclear Regulatory Commission, to Biff Bradley, NEI, "Recent Fire PRA Methods Review Panel Decisions and EPRI 1022993, Evaluation of Peak Heat Release Rates in Electrical Cabinet Fires," (ADAMS Package Accession No. ML12172A406).
The LAR and detailed FM analysis indicate that although a bounding 98% HRR of 317 kW from NUREG/CR-6850 was typically used, a reduced transient fire HRR seems to have been applied as part of detailed FM for certain fire areas (e.g., the CSR, Intake Structure, and East Cableway).
Discuss the key factors used to justify the reduced rate below 317 kW. Include in this discussion:
a) Identification of the fire areas where a reduced transient fire HRR is credited and what reduced HRR value was applied.
b) A description for each location where a reduced HRR is credited, and a description of the administrative controls that justify the reduced HRR including how location-specific attributes and considerations are addressed. Include a discussion of the required controls for ignition sources in these locations along and the types and quantities of combustible materials needed to perform maintenance. Also, include discussion of the personnel traffic that would be expected through each location.
c) The results of a review of records related to compliance with the transient combustible and hot work controls.
NFPA 805 PRA RAI 07 - Sensitive Electronics
Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 2, (ADAMS Accession No. ML081130188), as providing methods acceptable to the NRC staff for adopting a FPP consistent with NFPA 805. In a letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.
Though LAR Attachment H refers to FAQ 13-0004, Clarifications on Treatment of Sensitive Electronics (ADAMS Accession No. ML13322A085), the fire scenario development and detailed FM indicates that guidance from FAQ 13-0004 was not used. For example, it appears that for sensitive electronics enclosed in electrical cabinets that inspection and walkdowns of cabinet configurations as recommended by the guidance in FAQ 13-0004 were not performed.
However, it appears that a sensitivity study on sensitive electronics may have been performed.
Still, the LAR does not describe a sensitivity study performed for sensitive electronics or present the quantitative results of such a study, and the study does not appear to be included as part of the FPRA uncertainty analysis. In light of these observations:
a) Describe the treatment of sensitive electronics for the FPRA and explain whether it is consistent with the guidance in FAQ 13-0004, including the caveats about configurations that can invalidate the approach (i.e., sensitive electronics mounted on the surface of cabinets and the presence of louver or vents).
b) If the approach cannot be justified to be consistent with FAQ 13-0004, then justify that the treatment of sensitive electronics has no impact on the NFPA 805 application.
Alternatively, replace the current approach with an acceptable approach in the integrated analysis performed in response to PRA RAI 03.
NFPA 805 PRA RAI 08 - Consideration of Violations in Determining Hot Work/Transient Fire Frequency Influence Factors Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 2, (ADAMS Accession No. ML081130188), as providing methods acceptable to the NRC staff for adopting a FPP consistent with NFPA 805. In a letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. FAQ 12-0064 provides guidance on determining hot work/transient fire frequency influence factors (ADAMS Accession No. ML12346A488). Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.
During the audit, the licensee presented a Condition Report (CR) where a violation was self-identified in the CSR that could affect the influence factors assigned to the CSR. Violations of transient combustible controls play a role in the assignment of influence factors. As a result, the staff is requesting that the licensee perform a review of CRs for all violations in Units 1 and 2, and evaluate the impact of these violations on their assignment of influence factors for Units 1 and 2. Should changes be made in the influence factors, update the PRA as needed, and incorporate that update into the response for PRA RAI 3 if needed.
NFPA 805 PRA RAI 09 - Minimum Joint Human Error Probability NUREG-1921, EPRI/NRC-RES Fire Human Reliability Analysis Guidelines - Final Report, (ADAMS Accession No. ML12216A104), discusses the need to consider a minimum value for the joint probability of multiple human failure events (HFEs) in human reliability analyses (HRAs). NUREG-1921 refers to Table 2-1 of NUREG-1792, Good Practices for Implementing Human Reliability Analysis (HRA), (ADAMS Accession No. ML051160213), which recommends that joint human error probability (HEP) values should not be below 1E-5. Table 4-4 of Electrical Power Research Institute (EPRI) 1021081, Establishing Minimum Acceptable Values for Probabilities of Human Failure Events, provides a lower limiting value of 1E-6 for sequences with a very low level of dependence. Therefore, the guidance in NUREG-1921 allows for assigning joint HEPs that are less than 1E-5, but only through assigning proper levels of dependency.
The FPRA uncertainty analysis appears to include a sensitivity study to evaluate the impact of the minimum joint HEP on the fire risk estimates. The study concludes that the FPRA CDF and LERF are not sensitive to assumptions made about the joint HEP value. However, the results of this sensitivity study and description about how the study was conducted is not on the docket.
The LAR does not provide this information and does not explain what minimum joint HEP value is currently assumed in the FPRA. Also, even if the assumed minimum joint HEP values are shown to have no impact on the current FPRA risk estimates, it is not clear to the NRC staff how it will be ensured that the impact remains minimal for future PRA model revisions supporting post-transition changes. In light of these observations:
a) Explain what minimum joint HEP value was assumed in the FPRA.
b) If a minimum joint HEP value less than 1E-05 was used in the FPRA, then provide a description of the sensitivity study that was performed and the quantitative results (i.e.,
CDF, LERF, CDF, and LERF) that justify that the minimum joint HEP value has no impact on the application.
c) If, in response part (b), if it cannot be justified that the minimum joint HEP value has no impact on the application, then provide the following:
- i. Confirm that each joint HEP value used in the FPRA below 1E-5 includes its own justification that demonstrates the inapplicability of the NUREG-1792 lower value guideline (i.e., using such criteria as the dependency factors identified in NUREG-1921 to assess level of dependence). Provide an estimate of the number of these joint HEP values below 1.0E-5, discuss the range of values, and provide at least two different examples where this justification is applied.
ii. If joint HEP values used in the FPRA below 1E-5 cannot be justified, set these joint HEPs to 1E-5 in the integrated analysis provided in response to PRA RAI 03.
d) If a minimum joint HEP value of less than 1E-05 was used but justified because it has no impact on the FPRA results, then add an implementation item that provides an action to confirm that the impact of joint HEP value continues to have a minimal impact on the FPRA estimates in future FPRA models used for post-transition changes.
NFPA 805 PRA RAI 10 - Obstructed Plume Model Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the staff for adopting a FPP consistent with NFPA-805.
NUREG-2178, Volume 1 "Refining And Characterizing Heat Release Rates From Electrical Enclosures During Fire (RACHELLE -FIRE), Volume 1: Peak Heat Release Rates and Effect of Obstructed Plume," (ADAMS Accession No. ML16110A140)16) contains refined peak HRRs, compared to those presented in NUREG/CR-6850, and guidance on modeling the effect of plume obstruction.
The FM performed in support of the FPRA appears to use the guidance from NUREG-2178, Volume 1, though it is not clear whether guidance on modelling the effect of an obstructed plume was used. NUREG-2178 provides guidance that indicates that the obstructed plume model is not applicable to cabinets in which the fire is assumed to be located at elevations of less than one-half of the cabinet.
If obstructed plume modeling was used, then indicate whether the base of the fire was assumed to be located at an elevation of less than one-half of the cabinet.
Justify any modelling in which the base of an obstructed plume is located at less than one half of the cabinets height, or remove credit for the obstructed plume model in the integrated analysis provided in response to PRA RAI 03.
NFPA 805 PRA RAI 11 - Treatment of Main Control Room Fires Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the staff for adopting a FPP consistent with NFPA-805. In a letter dated July 12, 2006, to NEI (ADAMS Accession No. ML061660105), the NRC established the FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02. Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method. FPRA FAQ 14-0008, Main Control Board Treatment (ADAMS Accession No. ML14190B307) provides guidance on modeling fires in the MCR.
The MCBs in the MCR appear to consist of a front and rear side connected by a single enclosure with a continuous ceiling. However, the FM performed for the MCBs appears to treat
the cabinets behind the MCB horseshoe as separate electrical cabinets instead of treating them as the rear side of the MCB. The guidance in FAQ 14-0008 indicates that if the front and rear side of such a configuration are connected together in an enclosure where the presence of a MCB cabinet ceiling would connote a single cabinet, then the rear cabinets should classified as an integral part of the MCB. For this MCB configuration, the guidance in FAQ 14-0008 provides three options for applying Appendix L of NUREG/CR-6850 to address fire progression associated with the MCB. HNPs treatment of the MCB appears to deviate from NRC accepted guidance.
In a separate MCB modelling concern, the NRC staff notes that a damage delay of 15 minutes was credited due to the presence of solid barriers between MCB cabinets. However, it seems that a number of, or all of, the MCR MCBs have open backs (or backs that are open within the large MCB enclosure). NUREG/CR-6850 Section 11.5.2.8 indicates that the approach described in NUREG/CR-6850 Appendix L may be used for cabinets separated by a single wall with back covers. It is not clear to NRC staff how the presence of solid barriers between MCB cabinet segments can be credited for a 15 minute damage delay for MCB cabinets with open backs.
In light of the observations above, address the following:
a) Describe the MCB configuration for the MCR and compare its configuration with those elements of FAQ 14-0008. Include discussion of the area between the cabinets that comprise the MCB horseshoe and the cabinets on the backside of the MCB horseshoe that appear to NRC staff to be part of single MCB enclosure.
b) Justify that the cabinets behind the MCB horseshoe are not part of single integral MCB enclosure using the definition in FAQ 14-0008.
c) Describe the mechanisms that were considered in the fire PRA which produced fire damage of targets across the walkway between the MCB and the cabinets just behind them. Include summary of the relevant fire modeling.
d) If it cannot be shown in response to part (b) above that the cabinets behind the MCB horseshoe are not part of single integral MCB enclosure using the definition in FAQ 14-0008, then justify treatment of the cabinets on the rear side of the MCB as separate electrical cabinets. Include clarification of how the back side cabinets are modelled and an explanation of how the treatment aligns with NRC accepted guidance.
e) If in response to parts (c & d) above, the current treatment of the MCB horseshoe and the cabinets behind the MCB horseshoe cannot be justified using NRC accepted guidance, then update the treatment of the MCB enclosure to be consistent with the guidance in FAQ 14-0008 in the integrated analysis provided in response to PRA RAI 03.
f) Clarify whether the MCB, or whether certain individual cabinets of the MCB, have an open back (or backs that are open within the large MCB enclosure).
g) If the MCB, or individual cabinets of the MCB, have an open back, then justify the credit taken in the FPRA for a damage delay of 15 minutes due to the presence of solid barriers between MCB cabinets. Include a description of the FM that supports the damage delay assumption of 15 minutes.
h) If in the response to part (g) above, the credit for a 15 minute delay in damage cannot be justified then update the fire propagation assumptions for MCB cabinets to be consistent with NRC guidance concerning cabinets with open backs in the integrated analysis provided in response to PRA RAI 03.
NFPA 805 PRA RAI 12 - MCR Abandonment on Loss of Habitability Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. RG 1.205 identifies NUREG/CR-6850 as documenting a methodology for conducting a FPRA and endorses, with exceptions and clarifications, NEI 04-02, Revision 2, as providing methods acceptable to the staff for adopting a FPP consistent with NFPA-805.
Methods that have not been determined to be acceptable by the NRC staff or acceptable methods that appear to have been applied differently than described require additional justification to allow the NRC staff to complete its review of the proposed method.
The LAR does not describe how MCR abandonment scenarios due to loss of habitability (LOH) were modelled. NCR staff notes that LAR Attachment W, Table W-2 includes among the top fire CDF contributors three MCR abandonment scenarios due to LOH with Conditional Core Damage Probabilities (CCDPs) ranging as large as 2.5E-01 to 8.18E-01. Nonetheless, it is still not completely clear to the NRC staff how the treatment of MCR abandonment due to LOH addresses the complexity associated with the full range of fire impacts that can occur from fires in the MCR. NRC staff notes that this complexity can present a significant modelling challenge.
In light of the observations above, address the following:
a) Explain how the CCDPs and conditional large early release probabilities (CLERPs) were estimated for MCR abandonment scenarios due to LOH. Include:
- i. Identification of the actions required to execute successful alternate shutdown and how they are modeled in the FPRA, including actions that must be performed before leaving the MCR.
ii. Explanation of how command and control is performed given that Unit 1s Remote Shutdown Panel is divided between four panel locations in four separate fire zones.
iii. Explanation of how the complexity associated with actions performed from multiple panel locations is considered in the HEPs that are used to estimate the CCDP and CLERP.
iv. Discussion of the challenge of maintaining communication between operators at different panels who are coordinating plant control and how this is factored into development of the MRC abandonment HEPs.
b) Explain how various possible fire-induced failures are addressed in the CCDP and CLERP estimates for fires that lead to abandonment due to loss of habitability.
Specifically include in this explanation, a discussion of how the following scenarios are addressed:
- i. Scenarios where fire fails only a few functions aside from forcing MCR abandonment and successful alternate shutdown is straightforward; ii. Scenarios where fire could cause some recoverable functional failures or spurious operations that complicate the shutdown, but successful alternate shutdown is likely; and, iii. Scenarios where the fire-induced failures cause great difficulty for shutdown by failing multiple functions and/or complex spurious operations that make successful shutdown unlikely.
c) Provide the range of CCDP and CLERP values for MCR abandonment scenarios due to loss of habitability for the appropriate fire areas for the post-transition plant model.
Include explanation for why the range of CCDPs and CLERPS for MCR abandonment scenarios were similar values for both units even though the complexity of the alternate shutdown actions is much greater for Unit1 than for Unit 2.
d) Provide the range of frequency of MCR abandonment scenarios due to loss of habitability for the post-transition plant cases.
e) If in the response to part (b) and (c) above, it cannot be justified that the current modelling of MCR abandonment due to LOH addresses the complexity associated with the full range of fire impacts that can occur from fires in the MCR, then replace the current approach with an approach that does address the full range of fire impacts that can occur from fires in the MCR due to LOH in the integrated analysis provided in response to PRA RAI 03.
NFPA 805 PRA RAI 13 - MCR Abandonment on Loss of Control The LAR does not describe how MCR abandonment scenarios due to loss of control (LOC) were modelled in the FPRA. LAR Attachment W, Tables W-2 and W-3 do not include among the top contributors to fire CDF MCR abandonment scenarios due to LOC. Based on the information provided, it is not clear to the NRC staff whether the treatment of MCR abandonment due to LOC addresses the complexity associated with the full range of fire impacts that can occur from fires in MCR abandonment areas (which appear to be the MCR, Cable Spreading Room (CSR), and Computer Room). The NRC staff notes that this complexity can present a significant modelling challenge. The LAR does not describe what cues and procedures would be used by operators in an actual fire scenario to trigger the decision to abandon the MCR due to LOC. Accordingly, it is not clear to the NRC staff that the failure of operators to make the decision to abandon the MCR and perform alternate shutdown is modeled in the FPRA.
In light of the observations above, address the following:
a) Identify those locations in the plant in which fire could lead to LOC for which MCR abandonment and alternate shutdown actions are credited in the FPRA.
b) Explain how various possible fire-induced failures are addressed in the CCDP and CLERP estimates for fires that lead to MCR abandonment due to LOC. Specifically include in this explanation, a discussion of how the following scenarios are addressed.
As a part of this response, indicate if the plant response is fully integrated into the PRA.
- i. Scenarios where fire fails only a few functions aside from forcing MCR abandonment and successful alternate shutdown is straightforward; ii. Scenarios where fire could cause some recoverable functional failures or spurious operations that complicate the shutdown, but successful alternate shutdown is likely; and, iii. Scenarios where the fire-induced failures cause great difficulty for shutdown by failing multiple functions and/or complex spurious operations that make successful shutdown unlikely.
c) Identify the range of CCDP and CLERP values for MCR abandonment scenarios for the appropriate fire areas due to LOC for the post-transition models. Identify those scenarios which have a CCDP of 1, or explain why there are no such scenarios.
d) Provide the range of frequency of MCR abandonment scenarios due to LOC for the appropriate fire areas for the post-transition plant case.
e) Explain how command and control is performed given that Unit 1s Remote Shutdown Panel is divided between four panels in four separate fire zone locations. Include discussion of the challenges of maintaining communication between operators who must perform actions at the four different panels and how this is factored into development of the HEPs that are used to estimate the CCDP and CLERP.
f) If in the response to part (b) and (c) above, if it cannot be justified that the current modelling of MCR abandonment due to LOC addresses the complexity associated with the full range of fire impacts that can occur from fires in MCR abandonment areas, then replace the current approach with an approach that does address the full range of fire impacts that can occur from fires in MCR abandonment areas in the integrated analysis provided in response to PRA RAI 03.
g) Indicate how the decision to abandon the MCR due to LOC is made procedurally by operators. Include discussion of the cues that would trigger the decision to abandon the MCR due to LOC.
h) Explain how the failure of operators to make the decision to abandon the MCR and perform alternate shutdown actions is modeled in the FPRA. Include in the explanation justification that the modeling is consistent with the guidance in NUREG-1921.
i) If failure of operators to make the decision to abandon the MCR and perform alternate shutdown is not modeled in the FPRA, then justify that this exclusion does not impact the application. Alternatively, incorporate failure of operators to make the decision to abandon the MCR and perform alternate shutdown in the integrated analysis provided in response to PRA RAI 03 consistent with the guidance in NUREG-1921.
NFPA 805 PRA RAI 14 - PRA Treatment of Dependencies between Units 1 and 2 Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in public health risk arising from transition from the current FPP to an NFPA 805 based program, and all
future plant changes to the program, shall be acceptable to the NRC. RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes to these frequencies that result from proposed changes to the plants licensing basis and describes a general framework to determine the acceptability of risk-informed changes. Section C of RG 1.174 states that PRAs supporting risk-informed applications should be based on the as-built, as-operated and maintained plant.
The LAR indicates that Unit 1 and 2 are adjoined and it makes reference to common areas, a cross-tie (e.g., RHRSW to RHR cross-tie) and systems shared between units (e.g., the Diesel Generator 1B). LAR Attachment W shows contribution by fire area for CDF, LERF, CDF, and LERF, but does not explain how the risk contribution from fires originating in one unit is addressed for impacts to the other unit given the physical proximity of the other unit, common areas, and the existence of shared systems. Therefore, address the following:
a) Explain how the risk contribution of fires originating in one unit is addressed for the other unit given impacts due to the physical proximity of equipment and cables in one unit to equipment and cables in the other unit. Include identification of locations where fire in one unit can affect components in the other unit and explain how the risk contributions of such scenarios are allocated in LAR Attachment W, Tables W-4 and W-5.
b) Explain how the contributions of fires in common areas are addressed, including the risk contribution of fires that can impact components in both units.
c) Explain the extent to which systems are shared by both units and whether shared systems are credited in the PRA models for both units. If shared systems are credited in the PRA models for each unit, then explain how the PRAs address the possibility that a shared system is demanded in both units in response to a single IE or fire initiator.
NFPA 805 PRA RAI 15 - Calculation of the Change in Risk Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in public health risk arising from transition from the current FPP to an NFPA 805 based program, and all future plant changes to the program, shall be acceptable to the NRC. RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes to these frequencies that result from proposed changes to the plants licensing basis and describes a general framework to determine the acceptability of RI changes. LAR Attachment W, Section W.2.1 provides a general description of how the change-in-risk associated with variances from deterministic requirements (VFDRs) is determined, including discussion about setting fire-induced failures events to false in the FPRA model as a way to mimic the compliant plant condition. Based on the information provided in the LAR, the NRC staff was unable to fully understand how the change in risk is calculated, therefore, the NRC staff requests that the licensee:
a) Describe the kinds of model adjustments (if there is more than one type) made to remove different types of VFDRs from the compliant plant model, such as adding events or logic, or the use of surrogate events. Clarify whether the approach used is consistent with guidance in FAQ 08-0054, Demonstrating Compliance with Chapter 4 of NFPA 805 (ADAMS Accession No. ML15016A280 and associated references therein). In addition, identify any major changes made to the FPRA models or data for the purpose of evaluating VFDRs.
b) Because the determination of the change-in-risk for MCR abandonment scenarios can be more complex than for other scenarios in the FPRA:
- i. Describe the model adjustments that were made to remove the VFDRs to create the compliant plant model for MCR abandonment scenarios due to both LOH and LOC.
ii. Describe the criteria used to identify Primary Control Station (PCS) locations.
iii. Explain whether VFDRs were identified differently for fire areas in which MCR abandonment (alternate shutdown) may be required compared to fire areas where MCR abandonment would not occur. If VFDRs were identified differently for MCR abandonment scenarios compared to other areas of the plant, then describe that difference.
iv. If assumptions were made, specific to MCR abandonment scenarios, about modeling the compliant plant (e.g., assumptions about how the CCDP values were determined), then describe and justify those assumptions. As part of the justification, provide an indication of the impact that those assumptions make on the NFPA 805 transition change-in-risk.
c) Describe the types of VFDRs identified, and discuss whether and how the VFDRs identified but not modeled in the FPRA impact the risk estimates. Describe the qualitative rational for excluding VFDRs from the change-in-risk calculations.
d) Explain, for both the compliant and transition plant PRA models, whether plant modifications are credited in the model. Clarify whether plant modifications that do not resolve VFDRs are credited in the transition (variant) plant model, but not in the compliant plant model, as a way to reduce risk (i.e., indicative of a combined change as discussed in Section 1.1 of RG 1.174). If modifications are credited in the transition plant model to reduce risk but do not resolve a VFDR, then provide the total risk increase associated with unresolved VFDRs and the total risk decrease associated with non-VFDR modifications.
NFPA 805 PRA RAI 16 - Assumed Cable Routing and Other Conservative Modeling Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in public health risk arising from transition from the current FPP to an NFPA 805 based program, and all future plant changes to the program, shall be acceptable to the NRC. RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes to these frequencies that result from proposed changes to the plants licensing basis and describes a general framework to determine the acceptability of RI changes.
Table 6-3, Fire PRA Sources of Uncertainty, of your application to adopt 10 CFR 50.69, risk-informed (RI) categorization and treatment of structures, systems, and components for nuclear power reactors (ADAMS Accession No. ML18158A583), states that that a sensitivity study was performed to address the uncertainty associated with un-located/untraced secondary-side cables given the conservative assumption made in the FPRA that secondary-side systems are failed in all fires. The LAR does not discuss this sensitivity study nor does it provide the
quantitative results of the sensitivity study. The assumption that untraced cables are failed in all fire sequences is a conservative approach for modeling untraced cables in the post-transition plant model, but can lead to underestimation of the change-in-risk when used in the compliant plant model a) Describe the extent of untraced FPRA cables and how they were treated in the FPRA.
Include an explanation of how they were modelled in both the compliant and post-transition plant FPRA models.
b) Justify that assumptions made about untraced cables do not contribute to underestimation of the transition change-in-risk. Include a description of the sensitivity study that was performed to address un-located/untraced cables as well as the quantitative results of that sensitivity study.
c) If failing all untraced cables in the FPRA leads to underestimation of the transition change-in-risk, then demonstrate that the application is not impacted by the underestimation of the transition change-in-risk. Alternatively, replace this conservative approach with an acceptable approach that does not underestimate the change-in-risk in the integrated analysis requested in PRA RAI 03.
d) If other conservative treatments used in the compliant plant model can be identified as contributing to the underestimation of the total change-in-risk, then identify those conservatisms and demonstrate that the application is not impacted by the corresponding underestimation of the transition change-in-risk. Alternatively, replace such approaches with more realistic approaches in the integrated analysis requested in PRA RAI 03 that do not underestimate the change-in-risk.
NFPA 805 PRA RAI 17 - Defense-in-Depth (DID) and Safety Margin NFPA 805, Section 1.2 indicates that defense-in-depth (DID) shall be achieved when an adequate balance of each of the following DID elements is provided: (1) Preventing fires from starting, (2) Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage, and (3) Providing an adequate level of fire protection for structures, systems, and components important to safety, so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed. LAR Section 4.5.2.2 provides a high-level description of how impacts on DID and safety margin were reviewed for the transition to NFPA 805, but did not provide sufficient information for the NRC staff to determine whether each DID element was properly addressed. Also, LAR Section 4.5.2.2 states that [f]ire protection features and systems relied upon to ensure DID were identified as a result of the assessment of DID, but LAR Attachment C, Table C-2 does not identify any fire protection systems or features to be credited for DID. Based on the above identified issues, the NRC staff requests that the licensee:
a) Explain the criteria used to determine when a substantial imbalance between DID echelons exist in the fire risk evaluations (FREs), and identify the types of plant features and administrative controls credited for providing DID for each of the three DID echelons.
b) Clarify what fire protection features and systems were relied upon to ensure DID and explain why none are identified in LAR Attachment C, Table C-2.
c) Discuss the approach for reviewing safety margins using the NEI 04-02, Revision 2, criteria for assessing safety margin in the FREs.
NFPA 805 PRA RAI 18 - Impact of a Key Source of Uncertainty on Application Section 2.4.3.3 of NFPA 805 states that the PRA approach, methods, and data shall be acceptable to the NRC. Section 2.4.4.1 of NFPA 805 further states that the change in public health risk arising from transition from the current FPP to an NFPA 805 based program, and all future plant changes to the program, shall be acceptable to the NRC. RG 1.174 provides quantitative guidelines on CDF, LERF, and identifies acceptable changes to these frequencies that result from proposed changes to the plants licensing basis and describes a general framework to determine the acceptability of RI changes. With regard to model uncertainty, Section 2.5.3 of RG 1.174 states that In many cases, the industrys state of knowledge is incomplete, and there may be different opinions on how the models should be formulated. It also states that understanding the impact of key assumptions may be addressed by performing the appropriate sensitivity studies. , Disposition of Key Assumptions/Sources of Uncertainty, of your application to adopt 10 CFR 50.69, provides dispositions for candidate key assumptions and sources of uncertainty for RI categorization. One uncertainty identified that may impact the NFPA 805 application concerns the assumed conditional probability of 1E-02 used account for the loss of net positive suction head (NPSH) following emergency containment venting which leads to failure of the of low pressure emergency core cooling system pumps. The 50.69 LAR did not explain the basis for the 1E-02 value or indicate how much uncertainty may exist in this assumption. NFPA 805 LAR Attachment C, Table C-1 identifies VFDRs associated with spurious opening of the Safety Relief Valves (SRVs) which suggests that assumptions made regarding loss of NPSH following containment venting might have an impact on the estimated change-in-risk. Accordingly, the NRC staff observes that the sensitivity of the fire change-in-risk results for the NFPA 805 application may be sensitive to the same modeling uncertainty as the 10 CFR 50.69 application. In light of these observations:
a) Describe the basis for the assumed conditional probability of 1E-02 for loss of NPSH given containment venting and indicate the degree of uncertainty that exists.
b) Justify why uncertainty in the assumed probability for loss of NPSH following containment venting has a minimal impact on fire risk estimates (i.e., CDF, LERF, CDF, and LERF).
If it cannot be qualitatively justified that the impact from the assumed probability for loss of NPSH following emergency venting has a minimal impact the fire risk estimates, then perform a sensitivity study on the integrated analysis provided in response to PRA RAI 03 demonstrating that the uncertainty associated with the assumed conditional probability of 1E-02 does not impact the NFPA 805 application.
NFPA 805 Fire Protection Engineering (FPE) RAI 01 The compliance strategy for NFPA 805 Section 3.3.5.1, in LAR Attachment A, Table B-1, is identified as "Complies with Required Action, with the actions being revising plant documentation and submitting for NRC approval. In LAR Attachment L Approval Request 3, the
licensee indicated that fire zones contain wiring above suspended ceilings that is not in compliance with NFPA 805, Section 3.3.5.1.
The licensee indicated that one of the basis for their approval request is that, there are small quantities of low voltage video, communication, and data cables, which are not susceptible to self-ignition. However, the licensee provided no justification for its statements that these types of cables are not susceptible to self-ignition.
Provide the technical basis for this statement and include whether the quantity and material properties of the cables impact the basis for the request.
NFPA 805 FPE RAI 02 The compliance strategy for NFPA 805, Section 3.3.5.2, "Electrical Raceway Construction Limits" in LAR Attachment A, Table B-1, is identified as Complies. However, in LAR Attachment L, Approval Request 4, the licensee is requesting approval for the use of polyvinylchloride (PVC) coated flexible conduit in lengths up to 6 feet and embedded non-metallic conduit.
It is not clear whether there are any non-embedded, non-metallic conduits installed at the plant.
Describe whether there are any non-embedded, non-metallic conduits installed at the plant. In addition, discuss whether approval is being requested for future installations or if future installations will be installed in accordance with the requirements of NFPA 805.
NFPA 805 FPE RAI 03 Fire protection systems and features that require NFPA code compliance are reflected in NFPA 805, Chapter 3, "Fundamental Fire Protection Program and Design Elements of NFPA 805."
The LAR does not contain a list of codes of record that establishes whether or how the licensee meets Chapter 3 of NFPA 805, therefore, the NRC staff requests that the licensee provide a complete list of the applicable NFPA codes and standards designated as the code of record, including identification of the edition (years), that will be in place post transition. For codes and standards with numerous editions, identify which editions pertain to which particular plant areas and systems.
NFPA 805 FPE RAI 04 In LAR Attachment A, Table B-1, the compliance strategies for NFPA 805 Sections 3.3.7.1, 3.3.8, and 3.4.1(a)(1) (for example), are identified as Complies with the use of existing engineering equivalency evaluation (EEEE) but does not describe whether or how non-compliances were resolved. Therefore, the NRC staff requests that the licensee describe how any non-compliances identified during these evaluations were addressed. If any non-compliances are still outstanding, describe how these will be addressed prior the completion of NFPA 805 implementation.