LIC-12-0083, Responses to Requests for Additional Information License Amendment Request 10-07 to Adopt NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants, 2001 Edition
| ML12208A131 | |
| Person / Time | |
|---|---|
| Site: | Fort Calhoun  |
| Issue date: | 07/24/2012 |
| From: | Bannister D Omaha Public Power District |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| LIC-12-0083 | |
| Download: ML12208A131 (195) | |
Text
Opp" OmahalPubltclPowrDisbict 444 South 16Mh Street Mall Omaha, NE 68102-2247 LIC-1 2-0083 July 24, 2012 U. S. Nuclear Regulatory Commission Attn: Document Control Desk Washington, DC 20555-0001
References:
- 1. Docket No. 50-285
- 2. Letter from OPPD (J. A. Reinhart) to NRC (Document Control Desk),
License Amendment Request 10-07, Proposed Changes to Adopt NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition) at Fort Calhoun Station, dated September 28, 2011 (LIC-1 1-0099) (ML112760660)
- 3. Letter from the NRC (L. E. Wilkins) to OPPD (David J. Bannister), Fort Calhoun Station, Unit No. 1 - Request for Additional Information Re: License Amendment Request to Adopt National Fire Protection Agency Standard NFPA 805 (TAC No. ME7244), dated April 26, 2012 (NRC-12-0041)
SUBJECT:
Responses to Requests for Additional Information Re: License Amendment Request 10-07 to Adopt NFPA 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants,"
2001 Edition, at Fort Calhoun Station The Omaha Public Power District's (OPPD's) responses to the Nuclear Regulatory Commission's (NRC's) requests for additional information (RAIs) regarding the license amendment request (LAR) to adopt National Fire Protection Association (NFPA) 805 at the Fort Calhoun Station (FCS) are provided in the Enclosure to this letter.
In the Reference 2 LAR, OPPD requested an amendment to Renewed Facility Operating License No. DPR-40 for FCS, Unit No.1, to adopt NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition). The NRC staff reviewed the information provided in OPPD's application and determined that additional information was required in order to complete its review. Reference 3 provides the NRC's RAIs which were received on April 26, 2012.
Employment with Equal Opportunity
U. S. Nuclear Regulatory Commission LIC-12-0083 Page 2 of 4 The original RAI responses were due to the NRC within 60 days from receipt of the RAIs. It was determined that some of the RAIs will require additional planning and analysis (e.g., fire dynamic simulations) in order to complete the final RAI responses. The status of the RAI responses, delaying the original RAI response submittal, and proposed extension of select RAI responses were discussed during teleconferences between the NRC and OPPD staff on June 22, 2012, and June 26, 2012. The final schedule for submitting the remaining RAI responses was found acceptable by the NRC technical reviewers via email on July 3, 2012.
OPPD will submit the remainder of these NFPA 805 RAI responses on a schedule as reflected in Table 1 below. This is a one-time regulatory commitment and is being tracked by action request AR 52508.
Table 1 - Schedule for NFPA 805 LAR RAIs Responses Submittal RAI Topic RAI Number Submittal Date Fire Modeling RAI 01 c August 25, 2012 Fire Modeling RAI 01 f.v.(a) September 28. 2012 Fire Modeling RAI 05 c.i. September 28. 2012 Fire Modeling RAI 05 e. September 28. 2012 Probabilistic Risk Assessment RAI 01 h.i., ii. and iii. August 25, 2012 Probabilistic Risk Assessment RAI 07 a, b and c September 28. 2012 During the preparation of the RAI responses, it was identified that various portions of the LAR Attachments and Transition Report sections in Reference 2 would need to be revised to reflect the RAI responses. In some cases, these proposed markups are included in the Enclosure as part of the RAI responses. To preclude having to submit the changes as a complete LAR revision under current FCS administrative processes, all of the markups of the proposed LAR Attachment revisions are not included with the RAI responses. However, the proposed document changes have been placed on the NFPA 805 SharePoint Portal for the NRC technical reviewers to use in support of their technical review. These markups will be incorporated and reflected in the NFPA 805 LAR supplement.
Therefore, in an effort to minimize the number of times the NFPA 805 transition LAR submittal is revised and docketed due to its volume and security sensitive contents, OPPD will supplement the LAR to reflect the RAI responses identified in Table 2, as well as any subsequent RAIs identified during the NRC NFPA 805 transition LAR reviews, at a time mutually agreeable between the NRC technical reviewers and OPPD project management.
This is a regulatory commitment that is being tracked under AR 48249.
Table 2 provides the LAR sections from Reference 2 impacted by the RAIs which will be reflected in the NFPA 805 LAR supplement, as delineated in the enclosed RAI responses.
U. S. Nuclear Regulatory Commission LIC-12-0083 Page 3 of 4 Table 2 - LAR Sections Impacted by the RAIs to be Updated in NFPA 805 LAR Supplement RAI Topic RAI Number LAR Section to be Updated Fire Modeling RAI 01 f.iv. Attachment S, Table S-3 Fire Modeling RAI 03 c.i. Attachment J, Table J-1 Fire Protection Engineering RAI 01 Attachment I, Table I-1 Fire Protection Engineering RAI 05 Attachment A, Table B-1 Fire Protection Engineering RAI 07 Attachment A, Table B-1 Fire Protection Engineering RAI 09 Attachment A, Table B-1 Fire Protection Engineering RAI 10 Attachment A, Table B-1 Fire Protection Engineering RAI 11 Attachment A, Table B-1 References Section 6.0 Fire Protection Engineering RAI 12 Attachment A, Table B-1 Fire Protection Engineering RAI 13 Attachment A, Table B-1 Fire Protection Engineering RAI 15 Attachment A, Table B-1 Fire Protection Engineering RAI 16 Attachment A, Table B-1 Fire Protection Engineering RAI 17 Attachment A, Table B-1 Attachment C, Table B-3 Transition Report Table 4-3 Fire Protection Engineering RAI 18 Attachment A, Table B-1 Attachment L Fire Protection Engineering RAI 19 Attachment A, Table B-1 Attachment L Fire Protection Engineering RAI 20 Attachment C, Table B-3 Transition Report Section 4.2.3 Attachments K, 0, and T Monitoring Program RAI 01a Transition Report Section 4.6.2 Monitoring Program RAI 01 b Transition Report Section 4.6.2 Monitoring Program RAI 01c Transition Report Section 4.6.2 Monitoring Program RAI Old Transition Report Section 4.6.2 Monitoring Program RAI 01f Transition Report Section 4.6.2 Monitoring Program RAI Olg Transition Report Section 4.6.2 Monitoring Program RAI 01 h Transition Report Section 4.6.2 Monitoring Program RAI 01 i Transition Report Section 4.7.3 Probabilistic Risk Assessment RAI 13 Attachment S, Table S-3 Safe Shutdown Analysis RAI 01 Transition Report Section 4.2.1.1 Safe Shutdown Analysis RAI 02 Transition Report Section 4.2.1.2 Safe Shutdown Analysis RAI 04 Transition Report Section 4.2.3 Transition Report Table 4.3 Attachment C, Table B-3 Attachments G, K, and T Safe Shutdown Analysis RAI 06 Attachment C, Table B-3
U. S. Nuclear Regulatory Commission LIC-12-0083 Page 4 of 4 In conclusion, the two regulatory commitments being made in this letter associated with the NFPA 805 Transition LAR RAI responses are provided in Table 3.
Table 3 - Regulatory Commitments Regulatory Commitment Commitment Due Date OPPD will submit the remainder of the The schedule for completion is NFPA 805 RAI responses on a delineated in Table 1.
schedule as reflected in Table 1. [AR 52508]
OPPD will supplement the LAR to reflect the RAI responses identified in Table 2, as well as any subsequent The LAR supplement will be RAIs identified during the NRC NFPA submitted at a time mutually 805 transition LAR reviews, at a time agreeable between the NRC mutually agreeable between the NRC technical reviewers and OPPD technical reviewers and OPPD project project management. [AR 48249]
management.
If you should have any questions regarding this submittal or require additional information, please contact the Supervisor - Nuclear Licensing, Mr. Bill R. Hansher at 402-533-6894.
Sincerely, D. J. Bannister Site Vice President and CNO DJB/BJVS/dll
Enclosure:
Responses to Requests for Additional Information c: E. E. Collins, Jr., NRC Regional Administrator, Region IV L. E. Wilkins, NRC Project Manager J. C. Kirkland, NRC Senior Resident Inspector
LIC-1 2-0083 Enclosure Page 1 of 164 RESPONSES TO REQUESTS FOR ADDITIONAL INFORMATION RE:
LICENSE AMENDMENT REQUEST TO ADOPT NATIONAL FIRE PROTECTION ASSOCIATION STANDARD 805, "PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT WATER REACTOR GENERATING PLANTS," 2001 EDITION FORT CALHOUN STATION, UNIT NO.1 DOCKET NO. 50-285 By letter dated September 28, 2011 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML112760660), as supplemented by letters dated December 19 and 22, 2011, and March 20,2012 (ADAMS Accession Nos. ML113540334, ML11363A077, and ML12083A147, respectively), Omaha Public Power District (OPPD, the licensee), submitted a license amendment request (LAR) to transition the fire protection licensing basis at the Fort Calhoun Station, Unit 1, from paragraph 50.48(b) of Title 10 of the Code of FederalRegulations (10 CFR), to 10 CFR 50.48{c), National Fire Protection Association Standard NFPA 805. Portions of the letters dated September 28 and December 22, 2011, and March 20, 2012, contain sensitive unclassified non-safeguards information (security-related and proprietary) and, accordingly, those portions have been withheld from public disclosure.
The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the information provided in your application and determined that the following additional information is required in order to complete its review. The RAIs are grouped into the following technical review categories: fire modeling (FM), fire protection engineering (FPE),
monitoring program (MP), programmatic, safe shutdown analysis (SSA), radioactive release (RR), and probabilistic risk assessment (PRA).
Fire Modeling RAI 01:
National Fire Protection Association Standard 805, "Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants," 2001 Edition, (NFPA 805), Section 2.4.3.3, states: "The PSA [probabilistic safety assessment] approach, methods, and data shall be acceptable to the AHJ [authority having authority] ... "The NRC staff noted that fire modeling comprised the following:
Fire Dynamics Simulator (FDS) was used to assess the main control room (MCR) habitability and to model an air compressor oil fire scenario in Fire Area FC32 (also referred to as Room 19).
The algebraic equations implemented in FDTs [Fire Dynamics Tools] and Fire Induced Vulnerability Evaluation, Rev. 1 (FIVE) were used to characterize flame radiation (heat flux), flame height, plume temperature, ceiling jet temperature, and hot gas layer (HGL) temperature for various ignition source types and heat release rates (HRRs).
Section 4.5.1.2, "FPRA Quality" of the Transition Report states that fire modeling was performed as part of the Fire Probabilistic Risk Assessment (FPRA) development (NFPA 805 Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V," for a discussion of the acceptability of the fire models that were used. Regarding the acceptability of the Probabilistic Risk Assessment (PRA) approach, methods, and data:
LIC-1 2-0083 Enclosure Page 2 of 164
- a. In regard to fire location corner and wall proximity effects, which can affect air entrainment and flame height, as well as Zone of Influence (ZOI) and target impacts, the Location Factor was not applied in the ZOI calculations for ignition sources that are at a distance of 6-inch or greater from a wall or corner. The justification is provided in the Fire Scenario Selection Report (FSSR) (FC07823 and CN-RAM-1 0-013), and refers to data published by Zukoski et al., for a 55 kilowatt (kW) circular pool fire. Please explain why the conclusions of Zukoski's work are valid for larger fires that involve solid combustibles and that are rectangular in shape (e.g., 317 kW transient fires).
OPPD's Response to Fire Modeling RAI 01 a.:
In absence of additional published guidance for larger fires, a sensitivity study was performed to consider wall/corner effects for fire scenarios modeled by the FCS fire PRA where the flaming region is expected to be within two feet of a wall/corner. The selection of two feet is intended to be bounding, which is supported by NUREG-1934, (Second Draft for Public Comment, published July 2011), Section 3.1.1 explaining that wall/corner effects may occur when "...the fire is postulated either flush with or at most a few inches from the wall or the corner....
OPPD performed a plant walkdown April 2-6, 2012, to identify modeled ignition sources within two feet of a wall/corner. Many ignition sources were identified to be within two feet of a wall (in particular, wall-mounted cabinets) and a few were identified within two feet of a corner; however, in many cases the conservative ZOI approaches implemented (e.g., applying the flame radiation ZOI from floor to ceiling, forming a cylindrical ZOI) was equivalent to or bounded the ZOI created by the wall/corner interaction. In addition, many of the wall-mounted electrical cabinets were characterized as "well-sealed" by the fire PRA, and the wall/corner interaction is therefore irrelevant for these cabinets. The following ignition sources were identified to be within two feet of a wall or corner and not already bounded by the FCS fire PRA:
" FC06-3-IS1 (Wall)
- FC36A-IS1 (Corner)
- FC36B-IS29 (Wall)
" FC36B-IS30 (Wall)
" FC36B-IS33 (Wall)
For each of the above ignition sources, the relevant expanded ZOls, accounting for wall/corner effects, were applied to identify targets. The fire PRA model was then re-quantified for each of the above scenarios accounting for wall/corner effects. The core damage frequency (CDF) and large early release frequency (LERF) results for the base fire PRA quantification reported in the NFPA 805 transition LAR (LIC-1 1-0099, hereafter referred to as "the LAR") and this sensitivity study are reported in the following table.
LIC-12-0083 Enclosure Page 3 of 164 Base Case Sensitivity Base Case Sensitivity CDF (/yr) Study CDF (Iyr) LERF(/yr) Study LERF (/yr)
FC06-3-IS1 3.58E-10 3.26E-07 1.92E-11 2.68E-09 FC36A-IS1 7.32E-09 8.86E-09 4.70E-10 4.74E-10 FC36B-IS29 1.33E-08 2.71 E-07 6.93E-10 2.61 E-09 FC36B-IS30 2.48E-07 2.54E-07 1.88E-08 1.89E-08 FC36B-IS33 1.76E-07 1.76E-07 1.34E-08 1.34E-08 The following table extends this sensitivity study for the overall plant risk.
Base Fire PRA* Sensitivity Study**
Effects)
(Wall/Corner Net ACDF for NFPA 805 5.72E-06 6.31 E-06 Transition (/yr)
Net ALERF for NFPA 6.67E-07 6.72E-07 805 Transition (/yr)
Total CDF (internal, 6.01 E-05 6.07E-05 flood, fire) (/yr)
Total LERF (internal, 4.82E-06 4.82E-06 flood, fire) (/yr)
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174 when wall/corner effects are modeled for ignition sources within two feet of a wall or corner.
- b. During the audit walkdown of the compressor area (Fire Area FC32/Room 19), the NRC staff observed two raw water pipes with a type of neoprene rubber insulation. The rubber insulation was not considered as an intervening combustible in the ZOI calculations and FDS analysis for this area. The staff is concerned about the possibility that non-cable intervening combustibles were also missed in other areas of the plant.
Please provide information how noncable intervening combustibles were accounted for in the fire modeling analyses.
OPPD's Response to Fire Modeling RAI 01 b.:
Non-cable intervening combustibles (e.g., foam pipe insulation, HVAC duct insulation, etc.) were assumed to negligibly contribute to the fire scenario development, with the primary generation of fire effects caused by the ignition source itself and overhead cable trays. Non-cable intervening combustibles were therefore not modeled by the FCS fire PRA supporting the LAR.
LIC-1 2-0083 Enclosure Page 4 of 164 A bounding sensitivity study assessing the impact of non-cable intervening combustibles is documented in this RAI response. Plant walkdowns were performed April 2 - 6, 2012, to identify non-cable intervening combustibles in fire compartments in which detailed fire modeling was performed. The walkdown identified the following non-cable intervening combustibles that could contribute to its associated scenario.
Scenario Ignition Source Intervening Combustible FC06-3-1S2 MCC-4B3 Insulated pipe FC06-3-1S3 MCC-3A3 Insulated pipe FC06-3-1S4 AI-205 Insulated pipe FC06-3-1S8 LP-5 Small bundled plastic tubing FC06-3-1S12 LP-1 Small bucket FC20-1 -1S8 CH-4B (30 hp) Blue painted insulation (demin. waste)
FC20-1-1S9 Al-182 Drain hose, 3/4 "good all", 1 or 2 bundles Blue demin. pipe insulation (4.75 ft) from skid -
cons. -> 1 pipe overhead approximately 5 ft FC20-7-1S6 AI-80 vertically from tray FC20-7-1S12 AI-284A Blue demin water insulation.
FC20-7-1S22 LP-9 Transients, drywall (cart with electronics)
FC20-7-1S25 CH-12-MS Small amount of insulation FC20-7-1S26 DW-41 B Black soft foam pipe insulation FC20-7-1S27 DW-41A Black soft foam pipe insulation FC20-7-1S32 DW-43A Black soft foam pipe insulation FC20-7-1S33 DW-43B Black soft foam pipe insulation 2 groups of 2 batteries. Each group is on FC31-1S35 FP-1 B BATTERY plywood. ( 2 plywood sheets x 4' x 1' x 1")
MCC-3B3 AND FC31-1S12 MCC-4C4 Transients PC31-1S17 HE-5 Power Switch Transients PC34B-1-IS8 AI-209 HVAC insulation FC34C-IS1 MCC-3C1 HVAC insulation FC34C-IS3 MCC-3B1 HVAC insulation FC34C-IS5 RC-4A VA-86 duct insulation 1.5 ft diameter x 11.9 ft FC34C-IS6 MCC-3A1 VA-86 duct insulation 1.5 ft diameter x 17.75 ft FC36A-IS1 CAB-SWYD-CONN Black pipe insulation FC36A-IS12 1B3C-4C Insulation to heater FC36A-IS25 1A1 Pipe insulation FC36B-IS38 1A4 Pipe insulation During the walkdown, ignition of intervening combustibles associated with the following ignition sources was judged to not be capable of failing targets beyond the modeled ZOI.
LIC-12-0083 Enclosure Page 5 of 164
- FC06-3-1S4
- FC06-3-1S8
- FC06-3-1S12
- FC20-1 -1S9
- FC20-7-1S12
- FC20-7-1S25
- FC31 -1S12 (No PRA targets nearby if ZOI expanded)
- FC31 -1S17 (No PRA targets nearby if ZOI expanded)
- FC36A-IS1
- FC36A-IS12
- FC36A-IS25
- FC36B-IS38 In Fire Compartment FC34C, the following combustibles are assumed to not ignite as the ductwork insulation is not directly over the ignition source and furthermore the ductwork is outside the flame radiation ZOI.
- FC34C-IS1
- FC34C-iS3
- FC34C-IS5 A conservative, bounding sensitivity study was performed on the remaining scenarios (i.e., scenarios listed in the table above, excluding those from the bulleted lists above). The conditional core damage probability (CCDP) and conditional large early release probability (CLERP) associated with full compartment burnout (i.e., hot gas layer) were conservatively applied to the following scenarios. Note that this is a conservative assumption for the purpose of this sensitivity study.
- FC20-7-1S6
- FC20-7-1S22
- FC20-7-1S26
- FC20-7-1S27
- FC20-7-1S32
- FC20-7-1S33
- FC31-1S35
- FC34B-1-IS8
- FC34C-IS6 When the above scenario frequencies are conservatively applied by this sensitivity study to their respective hot gas layer CCDP and CLERP values, the total fire CDF and LERF increase by 9.57E-07 /yr and 7.59E-09 /yr, respectively.
Regarding FC06-3-1S2 and FC06-3-1S3, the intervening combustible is pipe insulation. If either one of these scenarios were to ignite the pipe insulation, the target set obtained by combining the ZOI for each scenario is considered to be bounding. However, there are no PRA targets located within the ZOI for either ignition source, and the current fire PRA therefore bounds the potential contribution of the pipe insulation.
LIC-1 2-0083 Enclosure Page 6 of 164 Regarding FC20-1 -1S8, the intervening combustible is an insulated demineralized water pipe located two meters above the ignition source and just beneath the overhead cable tray. Using the severity factor concept from NUREG/CR-6850 Task 8, the pump fire heat release rate probability density function, per Appendix G of NUREG/CR-6850 and a vertical distance of two meters, results in a severity factor of 0.02. To conservatively assess the risk associated with ignition of the secondary combustible in scenario FC20-1-1S8, 98% (per calculated severity factor) of the fire frequency (3.69E-05 /yr) is applied to the baseline scenario CCDP (7.03E-04) and CLERP (3.78E-05), and 2% of the fire frequency is conservatively applied to the hot gas layer CCDP (1.0) and CLERP (1.05E-01).
The CDF and LERF for fire scenario FC20-1 -1S8 are re-calculated below:
- CDF = 3.69E-05 * [(0.98
- 7.03E-04) + (0.02
- 1.00)] = 7.63E-07 /yr The baseline CDF for this scenario is 1.29E-08 /yr, per fire PRA quantification supporting the LAR, and the bounding estimate of CDF increase when considering the pipe insulation is therefore calculated as (7.63E-07) - (1.29E-08) = 7.50E-07 /yr. The impact of this numerical value is assessed against the NFPA 805 acceptance criteria at the end of this RAI response.
- LERF = 3.69E-05 * [(0.98
- 3.78E-05) + (0.02
- 1.05E-01)] = 7.89E-08 /yr The baseline LERF for this scenario is 6.96E-10 /yr, per fire PRA quantification supporting the LAR, and the bounding estimate of LERF increase when considering the pipe insulation is therefore calculated as (7.89E-08) - (6.96E-10) = 7.82E-08 /yr. The impact of this numerical value is assessed against the NFPA 805 acceptance criteria at the end of this RAI response.
Additionally, the presence of intervening combustibles can contribute to the overall scenario heat release rate profile, which can affect whether a hot gas layer is expected to form, and its associated timing. For the purpose of this sensitivity study, the heat release rate contribution of each intervening combustible (primarily foam pipe insulation) is approximated as one square meter of polyurethane foam. Using a heat release rate per unit area of 710 kW/m2, per NUREG-1805 Chapter 8, and an assumed surface area of one square meter, yields 710 kW.
The following table summarizes the peak upper layer temperature and whether a hot gas layer is expected to form when this heat release rate is incorporated into the fire growth profiles for each ignition source. This is performed for all scenarios identified to have intervening combustibles, with the exception of those determined not to ignite the intervening combustible and those conservatively assumed to cause a hot gas layer (HGL).
LIC-1 2-0083 Enclosure Page 7 of 164 Does HGL form if 710 Scenario kW added? Notes FC06-3-1S2 NO FC06-3-1S3 NO FC06-3-1S4 NO FC06-3-lS8 NO FC06-3-IS12 NO FC20-1 -IS8 NO FC20-1-IS9 NO FC20-7-1S12 NO FC20-7-1S25 NO FC31-1S12 NO FC31-1S17 NO FC36A-IS1 NO FC36A-IS12 NO Black foam insulation area is much smaller than one square meter and will therefore not contribute 710kW. A minimum of 626 additional kW is required for hot gas layer formation, and there is not sufficient foam in this scenario to FC36A-IS25 YES contribute that heat release rate.
Black foam insulation area is much smaller than one square meter and will therefore not contribute 710kW. A minimum of 582 additional kW is required for hot gas layer formation, and there is not sufficient foam in this scenario to FC36B-IS38 YES contribute that heat release rate.
The following table extends this sensitivity study for the overall plant risk.
Sensitivity Study**
Base Fire PRA* (Intervening Combustibles)
Net ACDF for NFPA 805 5.72E-06 7.43E-06 Transition (/yr)
Net ALERF for NFPA 805 6.67E-07 7.53E-07 Transition (/yr)
Total CDF (internal, flood, 6.01 E-05 6.18E-05 fire) (/yr) III Total LERF (internal, flood, 4.82E-06 4.91 E-06 fire) (/yr) II
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
LIC-1 2-0083 Enclosure Page 8 of 164 In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, when conservative approximations of the contribution of intervening combustibles are modeled by the fire PRA.
- c. Please explain how the HRR of cable tray fires was calculated. Please provide justification for the use of a "characteristic length" of 1 foot instead of the cabinet length as specified in NUREG/CR-6850, "EPRI/NRC-RES, Fire PRA Methodology for Nuclear Power Facilities," page R-9, and explain (i.e.,
a sensitivity study to obtain a quantitative measure) the effect of this discrepancy on the ZOI and HGL calculations and target damage assessment. In addition, please provide a justification for the fact that the effect of the cable tray fire on the radius of the ZOI was not considered.
OPPD's Response to Fire Modeling RAI 01 c.:
OPPD plans to have a quantitative sensitivity study performed for comparing characteristic length assumption (generic one foot) from base fire PRA against using the actual characteristic length per fire scenario. Full plant walkdown and data collection is about 75% complete in support of this effort. As discussed during the June 22, 2012, teleconference with the NRC, OPPD will submit its response to this RAI on August 25, 2012. [AR 52508]
- d. In the Method of McCaffrey, Quintiere, and Harkleroad (MQH) correlation (NUREG-1805, "Fire Dynamics Tools (FDTs) Quantitative Fire Hazard Analysis Methods for the U.S. Nuclear Regulatory Commission Fire Protection Inspection Program"), a vent opening area of 1 square meter (M 2 ) and vent height of 1 meter (m) were used in all compartment fire modeling calculations. Changing the vent area to 0.5 m 2 and the vent height to 0.5 m, everything else being equal, would increase the MQH HGL temperature results by 41 percent. This appears to indicate that the HGL temperature results are very sensitive to the compartment ventilation openings. In addition, the NRC staff has concerns regarding the applicability of the MQH correlations for vents of this size that may be located in the upper part of a wall or in the ceiling of the compartment.
L. Please provide justification for the use of the MQH correlation for compartment vents of the order of 0.5 in to 1 M2 in size.
OPPD's Response to Fire Modeling RAI 01 d.i.:
OPPD's response to Fire Modeling RAI 03 b. demonstrates that the equivalence ratios for fire scenarios in which the MQH correlation was implemented are generally within validated range for fires up to 1,000 kW.
However, for the few cases in which heat release rates exceeding 1,000 kW were postulated, the equivalence ratio exceeds the validated range, and this primarily is due to generically assuming a one square meter ventilation opening area. This assumption is generally conservative as
LIC-1 2-0083 Enclosure Page 9 of 164 smaller ventilation opening areas yield higher predicted hot gas layer temperatures. This approach also conservatively neglects the temperature reducing effect of forced ventilation. Furthermore, the analysis is conservative because, for high heat release rate scenarios where this approach indicates the fire may be ventilation limited, the resulting reduced rate of combustion is not credited.
ii. Please explain (i.e., a sensitivity study) the effect of the vent opening area and height on the HGL temperature results.
OPPD's Response to Fire Modeling RAI 01 d.ii.:
Plant walkdowns were performed April 2 - 6, 2012, to identify the opening areas (length and width), opening heights, and ceiling heights for each fire compartment in which the MQH model was employed. A sensitivity study was then performed using the compartment-specific values for these parameters rather than the generically assumed one square meter opening area, one meter opening height, and three meter ceiling height.
The sensitivity study identified a slight increase in total fire-induced CDF from 2.68E-05 /yr to 2.69E-05 /yr and LERF from 2.06E-06 /yr to 2.07E-06
/yr. While the majority of fire scenario risk remained unchanged, the slight increase was attributed to several cable spreading room scenarios in which the actual vent opening area was less than the generically assumed value, increasing the propensity of these scenarios to hot gas layer formation. For the remaining compartments, modeling of the actual vent areas did not significantly affect the propensity for HGL formation. The risk increase for these scenarios was only slight because the CCDP and CLERP values associated with localized fire damage is similar to the values associated with hot gas layer generation.
The net CDF and LERF increases, between the base case and sensitivity study case, are 4.78E-08 /yr and 8.30E-09 /yr, respectively. The following table extends this sensitivity study for the overall plant risk.
Base Fire PRA* Sensitivity Study**
(Generic MQH (Compartment-Values) Specific MQH Values)
Net ACDF for NFPA 805 Transition 5.72E-06 5.77E-06
(/yr)
Net ALERF for NFPA 805 Transition 6.67E-07 6.75E-07
(/yr)
Total CDF (internal, flood, fire) (/yr) 6.01 E-05 6.01 E-05 Total LERF (internal, flood, fire) 4.82E-06 4.83E-06
(/yr)
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
LIC-12-0083 Enclosure Page 10 of 164 In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, when using the compartment-specific opening areas, opening heights, and ceiling heights for each compartment in which the MQH model was applied.
iii. Please compare the results with the HGL temperature results for closed compartments based on Beyler's correlation in NUREG-1805.
OPPD's Response to Fire Modeling RAI 01 d.iii.:
The method of Beyler calculates hot gas layer temperature for closed compartments (Reference NUREG-1805, Chapter 2 published December 2004). Plant walkdowns were performed April 2 - 6, 2012, to identify the opening areas (length and width) and opening heights for each fire compartment in which the MQH model was employed. Compartment-specific ceiling heights were obtained from plant documentation and used in this analysis. While the walkdown indicated that each of these compartments had openings beyond the minimum required to prevent pressure buildup, the following two compartments had the lowest total opening areas (approximately one square foot) and were identified for a sensitivity study comparing MQH against the method of Beyler:
" FC03 - Spent Regenerant Tank & Pump Area
" FC34C - Group 1 MCC Area Utilizing a bounding approach in which all fire scenarios in each of the above fire compartments are assumed to result in hot gas layer, with a non-suppression probability of 1.0, causes fire-induced CDF and LERF increases of 2.98E-06 /yr and 2.34E-07 /yr, respectively. The following table extends this sensitivity study for the overall plant risk.
Base Fire PRA* Sensitivity Study**
(MQH) (Beyler)
Net ACDF for NFPA 805 5.72E-06 8.70E-06 Transition (/yr)
Net ALERF for NFPA 805 6.67E-07 9.01 E-07 Transition (/yr)
Total CDF (internal, 6.01 E-05 6.31 E-05 flood, fire) (/yr)
Total LERF (internal, 4.82E-06 5.05E-06 flood, fire) (/yr)
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, when applying the Beyler method to fire modeled compartments with minimal (approximately one square foot) identified ventilation openings.
LIC-1 2-0083 Enclosure Page 11 of 164 iv. Please justify why Beyler's correlation for closed compartments was not used.
OPPD's Response to Fire Modeling RAI 01 d.iv.:
The method of Beyler calculates hot gas layer temperature for closed compartments (Reference NUREG-1 805, *Chapter 2 published December 2004). The FCS fire PRA does not implement the method of Beyler because FCS fire compartments are generally not closed (e.g., forced ventilation, natural ventilation openings, etc.). This was confirmed via plant walkdown performed April 2-6, 2012. Nonetheless, a sensitivity study was performed in response to Fire Modeling RAI 01 d.iii. and concluded that the NFPA 805 acceptance criteria would still be met if the method of Beyler were applied to compartments with minimal identified ventilation openings.
- e. It is stated, in the FSSR (Assumption 3 on page 53), that all cabinets are assumed to contain qualified cable, although some cabinets may contain nonqualified cable. This is justified on the basis that the 98th percentile HRR of cabinets with qualified cable in Table G-1 of NUREG/CR-6850, is higher than that of cabinets with non-qualified cable and closed doors.
However, the 98th percentile HRR is higher for cabinets with non-qualified cable and open doors. Please provide justification for not considering the possibility of cabinets with open doors.
OPPD's Response to Fire Modeling RAI 01 e.:
Walkdowns performed the week of April 2, 2012 confirmed that, outside the main control room, all electrical cabinets for which the heat release rate was modeled (e.g., for calculating ZOI and collecting source-target data) could be classified as closed-door. No open-door configurations were discovered, consistent with the initial fire PRA walkdowns and modeling.
Inside the main control room envelope, specifically in the computer room, the following four communication / server racks were identified with the potential to behave as open door cabinet fires:
" FC42-1S20 AI-312
" FC42-1S21 Al-311
- FC42-1S22 AI-316
- FC42-1S23 PS-CPU-1 These ignition sources are modeled in the main control room analysis as closed-door, qualified cable, multiple bundle cabinets.
LIC-12-0083 Enclosure Page 12 of 164 A sensitivity study was performed where the control room abandonment CDF and LERF were recalculated modeling the above cabinets as open-door, unqualified cable, multiple bundle with a heat release rate probability density function having a 98th percentile value of 1,002 kW, alpha of 0.46, and beta of 386. Note that NUREG/CR-6850 does not provide a heat release rate probability density function for open-door cabinets containing qualified cable. Note also that this sensitivity study was performed up to 1,000 kW (i.e., 97.998th percentile).
The CDF and LERF results for each the baseline (closed door) and sensitivity study (open door) cases are summarized in the following table.
Scenario Closed Door Open Door Closed Door Open Door (Baseline) CDF (/yr) CDF (/yr) (Baseline) LERF (/yr) LERF (/yr)
FC42-1S20 2.30E-08 2.52E-08 2.30E-09 2.52E-09 FC42-1S21 2.30E-08 2.52E-08 2.30E-09 2.52E-09 FC42-1S22 2.30E-08 2.52E-08 2.30E-09 2.52E-09 FC42-1S23 2.30E-08 2.52E-08 2.30E-09 2.52E-09 The net CDF and LERF increases for FC42 between the open door and closed door cases are 8.98E-09/yr and 8.98E-10/yr, respectively. The following table extends this sensitivity study for the overall plant risk.
Base Fire PRA* Sensitivity Study**
(Closed Door Case) (Open Door Case)
Net ACDF for NFPA 805 Transition (/yr) 5.72E-06 5.73E-06 Net ALERF for NFPA 805 Transition (/yr) 6.67E-07 6.68E-07 Total CDF (internal, flood, fire) (/yr) 6.01 E-05 6.01 E-05 Total LERF (internal, flood, fire) (/yr) 4.82E-06 4.82E-06
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, if the identified ignition sources were modeled as open door cabinets.
- f. Regarding the use of FDS:
- i. Please provide the input files in electronic format (*.fds) for all FDS runs that were conducted (15 in the MCR and 1 in the air compressor area (Fire Area FC32/Room 19)).
OPPD's Response to Fire Modeling RAI 01 f.i.:
The Fire Dynamics Simulator input files were provided to the NRC in electronic format (*.fds) as Enclosure 1 to the letter from OPPD to NRC, "Submittal of Fire Dynamic Simulator (FDS) Input Files in Response to
LIC-1 2-0083 Enclosure Page 13 of 164 NRC NFPA 805 Transition License Amendment Request Audit Team Request at Fort Calhoun Station," (LIC-12-0033) dated March 20, 2012.
ii. Please justify the grid spacing in the MCR and air compressor area (Fire Area FC32/Room 19) fire simulations based on a Characteristic Fire Diameter (D*) analysis. Please describe the size and location of the compartment vent(s) to the outside that is (are) specified in the FDS input files for the MCR and air compressor area (Fire Area FC32/Room 19) fire simulations.
OPPD's Response to Fire Modeling RAI 01 f.ii.:
The grid spacing of each FDS run is assessed here using the D*/dx ratio.
The characteristic fire diameter, D*, is determined using the following equation:
D* Q 2/5 poocpG0 g Where:
Q is the heat release rate (kW, varies per simulation)
Poo is the density of air (1.2 kg/m 3) cP is the specific heat of air (1.005 kJ/kg-K)
Too is the ambient air temperature (293 K) g is the acceleration of gravity (9.81 m/s 2) dx is the grid size in the x direction. Note that dy and dz are not evaluated here, since the mesh was setup such that each individual control volume is generally cubic. In addition, for the main control room analysis, a separate mesh was defined for each: the general main control room area and the plenum area. This analysis assesses D*/dx for the lower mesh containing the fire location.
The following table assesses D*/dx for the main control room fire simulations.
Main Control Room Analysis (kW) D* dx D*/dx 100 0.38 0.15 2.55 200 0.50 0.15 3.36 300 0.59 0.15 3.95 400 0.67 0.15 4.44 500 0.73 0.15 4.85 600 0.78 0.15 5.22 700 0.83 0.15 5.55 800 0.88 0.15 5.85 900 0.92 0.15 6.14 1000 0.96 0.15 6.40
LIC-1 2-0083 Enclosure Page 14 of 164 NUREG-1824 Volume 7 assessed a range of D*/dx ratios of 4-16. The upper heat release rate scenarios (400 - 1,000 kW) fall within the validate range, while the lower heat release rate scenarios (100 - 300 kW) fall slightly below the validated range, suggesting the grid implemented was relatively coarse. Using a finer grid for these fire scenarios is not expected to significantly alter the results and conclusions of the main control room abandonment analysis, which is a function of global hot gas layer temperature, heat flux, and smoke density, all of which are less sensitive to grid size than, for example, detailed examination of the plume dynamics.
The following table assesses D*/dx for the instrument air compressor fire simulation.
InstrumentAir Com pressor Anal sis Q(kW) D* dx D*Idx 1794 1.21 0.10 12.13 D*/dx falls within the validated range for the instrument air compressor analysis.
Regarding size and location of the specified vents to outside the control volume, the main control room simulations included a -0.5 meter wide opening spanning the entire length of the control volume, in the above-ceiling plenum space, which is connected to the main control room occupied area by a handful of small open ceiling vents. While this opening is arbitrary, review of the fire simulations suggests that the ventilation modeling does not significantly impact the resulting abandonment times, which occur early in the fire and primarily due to smoke obscuration. The small vent size between the occupied area and above-ceiling plenum does not provide a significant removal of smoke or heat from the occupied area.
For the instrument air compressor analysis, erroneously, no openings were specified to outside the control volume. A review of the simulation results indicated that this caused the modeled fire to extinguish (due to oxygen starvation) at approximately 800 seconds. While the simulation is likely conservative in the sense that no openings were modeled to vent and cool the upper layer, the simulation will be re-run with appropriate ventilation modeled in response to Fire Modeling RAIs Olf.v(a) and 05e, which also involve re-running the FDS simulations.
iii. FAQ 08-0052 "Transient Fires-Growth Rates and Control Room Non-suppression" specifies a time to peak heat release rate of 2 min and 8 min for MCR trash bags and trash cans, respectively. Please justify why a time to peak heat release rate of 5 min was used for transient fires in the MCR.
LIC-12-0083 Enclosure Page 15 of 164 OPPD's Response to Fire Modeling RAI 01 f.iii.:
FAQ 08-0052 (Reference NUREG/CR-6850 Supplement 1) provides recommended transient fire growth times for common trash cans and trash bags (unconfined to a can). A trash can is the expected transient combustible for the FCS main control room, and FAQ 08-0052 recommends a growth time of eight minutes for a common trash can.
FAQ 08-0052 recommends a growth time of two minutes for an unconfined trash bag, which is generally not expected in the FCS main control room. While trash cans are the expected transient combustible, the fire PRA conservatively assumes a growth time of five minutes, which helps bound uncertainty associated with predicting the exact configuration of the trash.
iv. Please justify why scenarios with an ignition source in the kitchen, computer room and other areas connected to the MCR were not considered.
OPPD's Response to Fire Modeling RAI 01 f.iv.:
Aside from the main control board and back panel areas, the main control room envelope includes a mechanical equipment room, computer room, and small kitchen. Ignition sources within the main control room envelope, with exception of the kitchen, are included in the ignition frequency calculation and modeled by the main control room abandonment analysis.
The fire frequency for each of these ignition sources is applied to the fire location postulated in the FDS runs.
Regarding the kitchen, although NUREG/CR-6850 does not identify kitchen appliances as a generic ignition source, even when attached to the MCR, OPPD recognizes that the kitchen in the FCS main control room contains ignition sources that may present fire risk. OPPD will evaluate and take action to minimize the risk contribution of this equipment as part of the 10 CFR 50.48(c) LAR implementation. Some options currently being considered are removing higher risk equipment, enclosing the kitchen within a fire rated boundary, or enhanced administrative controls while equipment is in use. The updated LAR Attachment S, including this implementation item will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
- v. Regarding the FDS simulation of the Fire Area FC32/Room 19:
(a) Please justify why the potential contribution from raw water pipe insulation to the HRR was not considered. Determine whether the insulation would be involved in the compressor oil scenario, and, if so, rerun FDS for this scenario accounting for the contribution of the pipe neoprene rubber insulation.
LIC-1 2-0083 Enclosure Page 16 of 164 OPPD's Response to Fire Modeling RAI 01 f.v.(a):
OPPD plans to re-run the FDS simulation including potential contribution of RW pipe insulation and assess impact on meeting NFPA 805 acceptance criteria (sensitivity study). The walkdown is complete but additional time is needed for analysis. As discussed during the June 26, 2012, teleconference with the NRC technical reviewers and subsequently agreed upon via email on July 3, 2012, OPPD will submit the response to this RAI on September 28, 2012.
[AR 52508]
(b) For the compressor oil fire scenario, please explain why the heat flux threshold for cables does not appear to have been considered and only the temperature threshold was evaluated.
OPPD's Response to Fire Modeling RAI 01 f.v.(b):
Cable failure due to exceeding its damage temperature threshold was selected for the instrument air compressor analysis.
Temperature, as opposed to heat flux, was judged to be the primary and most likely failure mechanism, and this judgment was based on the significant distance between the flaming region and the targets, and the target location in the plume and hot gas layer regions.
(c) Please demonstrate how a turbine-driven aux feedwater pump oil fire, including the potential contribution of the raw water pipe insulation in the vicinity of the pump would not result in further propagation or damage to additional targets. Please provide a summary description, assumption and basis, of the fire modeling conducted and if FDS was used, provide the input files in electronic format (*.fds).
OPPD's Response to Fire Modeling RAI 01 f.v.(c):
FDS was not used to evaluate this scenario. Turbine-driven auxiliary feedwater pump (FW-10) lubricating oil fires are divided into two scenarios in accordance with NUREG/CR-6850. In the first scenario (FC32-1S7-Oi110), 98% of the oil fire frequency is assumed to involve 10% of the oil inventory. In the second scenario (FC32-IS7-OillOO), 2% of the oil fire frequency is assumed to involve 100% of the oil inventory. (Note that NUREG/CR-6850, Section E.3 recommends that 2% of the oil fire frequency be assumed to involve 98% of the oil inventory, while the FCS fire PRA conservatively models 100% of the oil inventory.) FW-10 contains approximately 5.5 gallons of lubricating oil.
LIC-12-0083 Enclosure Page 17 of 164 The 10% oil case involves 0.55 gallons of oil and is assumed to only damage the pump itself (FW-10). This assumption is based on the small quantity of oil involved, the distance between the pump and overhead targets, and the wet pipe sprinkler heads located directly above the pump cage (above the pump but below overhead targets). With the same bases, the 10% oil case is not expected to ignite nearby pipe insulation, which is routed on the other side of the partial height wall separating FW-6 and FW-1 0.
The 100% oil case is conservatively assumed to cause hot gas layer formation if unsuppressed by the area-wide, smoke detector actuated, pre-action sprinkler system. This conservative approach bounds the potential contribution of nearby raw water pipe insulation to the fire scenario heat release rate.
Fire Modeling RAI 02:
NFPA 805, Section 2.5, 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, or components. Appropriate temperature and critical heat flux criteria must be used in the analysis.
Assumption 1 in Section 4.4 on page 53 of the FSSR (FC07823 and CN-RAM-10-013) states that, "All PRA targets are assumed to have a radiant heat flux damage threshold of 11 kW/m 2 and a temperature damage threshold of 3300 C. These damage thresholds are consistent with those for electrical cables with thermoset insulation."
Please provide the following information:
- a. Characterize the installed thermoset and thermoplastic cabling in the power block specifically with regard to the critical damage threshold temperatures and critical heat flux threshold as described in NUREG/CR-6850. Please provide a statement regarding the extent of installed thermoset cable insulation.
OPPD's Response to Fire Modeling RAI 02 a.:
FCS primarily uses cables having thermoset insulation, with use of thermoplastic insulation limited to fire detection and suppression applications and limited use in the main control room. This statement is based on the following OPPD assessments:
" Report No. 04-4035, Fort Calhoun Fire Induced Circuit Failure Assessment, Nexus Technical Services Corporation, December 30, 2004.
- EA-FC-00-102, Electrical Cable Scoping, Screening, & Aging Management Review for License Renewal, Revision 2.
LIC-1 2-0083 Enclosure Page 18 of 164 Therefore, the thermoset damage criteria specified by NUREG/CR-6850 (11 kW/m 2 and 3300C) was used for the purpose of identifying cable targets failed for each fire scenario evaluated using the ZOI approach.
- b. If thermoplastic cabling is present, please discuss the additional targets created/identified using the lower critical temperature damage threshold and/or critical heat flux damage threshold criteria of NUREG/CR-6850.
OPPD's Response to Fire Modeling RAI 02 b.:
The limited use of thermoplastic cable insulation did not necessitate identifying additional targets beyond those identified using the thermoset damage criteria.
Outside the main control room, thermoplastic cable insulation is used for detection and suppression applications. The fire PRA utilized in the LAR assumed fire detection and suppression cables are not considered fire PRA targets, as it is assumed that fire will not cause cable damage such that it prevents fire detection and suppression systems from performing their design functions.
Inside the main control room, thermoplastic cable insulation is used inside electrical cabinets. The main control room analysis for the fire PRA conservatively applied the heat release rate distribution for electrical cabinets containing qualified cable, despite the presence of thermoplastic cable. This approach is conservative because the qualified cable heat release rate distribution is biased toward higher values than the distribution for unqualified cable. For example, per NUREG/CR-6850 Table G-1, the 98th percentile for electrical cabinets with multi-bundle qualified cable is 702 kW, as opposed to 464 kW for electrical cabinets with multi-bundle unqualified cable.
- c. If thermoplastic cabling is present, please discuss impact on ZOI size due to increased HRR and fire propagation.
OPPD's Response to Fire Modeling RAI 02 c.:
The use of thermoplastic cable is sufficiently limited such that the zones of influence associated with thermoset damage criteria were applied to all fire scenarios and target sets.
- d. If thermoplastic cabling is present, please discuss self-ignited cables and their impact to additional targets created.
OPPD's Response to Fire Modeling RAI 02 d.:
The use of thermoplastic cable is sufficiently limited that self-ignited cable fires are not modeled by the FCS fire PRA.
LIC-1 2-0083 Enclosure Page 19 of 164
- e. If more targets are identified please describe the impact to core damage frequency (CDF) and large early release frequency (LERF), as well as changes in CDF (ACDF) and changes in LERF (ALERF) for those fire areas affected.
OPPD's Response to Fire Modeling RAI 02 e.:
Not applicable. The use of thermoplastic cable is sufficiently limited such that fire-induced CDF and LERF are calculated using thermoset damage criteria.
Fire Modeling RAI 03:
NFPA 805, Section 2.7.3.2, "Verification and Validation," states: "Each calculational model or numerical method used shall be verified and validated through comparison to test results or comparison to other acceptable models."
4.5.1.2, "FPRA Quality," of the Transition Report states that fire modeling was performed as part of the FPRA development (NFPA 805, Section 4.2.4.2). Reference is made to Attachment J, "Fire Modeling V&V," for a discussion of the verification and validation (V&V) of the fire models that were used. Furthermore Section 4.7.3 "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805" of the Transition Report 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."
Regarding the V& V of fire models:
- a. Attachment J of the Transition Report states that the algebraic equations implemented in FDTs and FIVE, Rev. 1, were used to characterize flame radiation, flame height, plume temperature, ceiling jet temperature, HGL temperature for various ignition source types and HRRs. However, the FDTs and/or FIVE, Rev. 1, spreadsheets were not used to perform the calculations, but selected algebraic models from NUREG-1805 and FIVE, Rev. 1, were used in a new spreadsheet (or set of spreadsheets). Please describe how this new spreadsheet (or set of spreadsheets) was verified, (i.e., how was it 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 NUREG-1805 or FIVE, Rev. 1).
OPPD's Response to Fire Modeling RAI 03 a.:
The fire modeling calculations supporting the FCS fire PRA, including the NUREG-1805 and EPRI FIVE, Revision 1, algebraic models that were codified into new spreadsheets were performed under the Westinghouse Quality Management System (QMS). The Westinghouse QMS describes Westinghouse commitments to the quality assurance requirements of ISO 9001, ISO 90003, 10 CFR 50 Appendix B, ASME NQA-1 Edition, and other national and
LIC-1 2-0083 Enclosure Page 20 of 164 international regulatory requirements. Fire modeling calculations supporting the FCS fire PRA were subject to full independent verification per the Westinghouse QMS, and this process assured that the fire modeling calculations were implemented correctly.
- b. For V&V of the aforementioned algebraic models reference is made to NUREG-1824, "Verification and Validation of Selected Fire Models for Nuclear Power Plant Applications." Please provide technical details to demonstrate that the algebraic models have been applied within the validated range of input parameters, or to justify the application of the equations outside the validated range reported in NUREG-1824.
OPPD's Response to Fire Modeling RAI 03 b.:
The fire modeling input parameters used by the FCS fire PRA have been either demonstrated to be within the range of applicability for fire models implemented or their use has been justified if outside the range of applicability.
The following table documents the algebraic fire model, aforementioned in response to RAI 03a, input values and their ranges as implemented by the FCS fire PRA.
Input Range for FCS Q (HRR (kW)) 69- 1,500 p- (ambient density of air (kg/m 3)) 1.2 cp (specific heat of air (kJ/kg-K)) 1.01 T- (ambient air temperature (K)) 298 g (acceleration of gravity (m/s 2 )) 9.81 D (fire diameter (m)) 0.344 - 0.609 Lf (flame length (m)) N/A H (ceiling height (m)) 3 rcj (ceiling jet radial distance (m)) 0.30 - 9.45 Hcj (height of ceiling jet relative to fire 1 elevation (m))
A. (area of the opening (M2 )) 1 Ho (height of the opening (m)) 1 V (ventilation rate (m 3/sec)) N/A L (length of compartment (m)) N/A W (width of compartment (m)) N/A Oxygen heat of consumption (kJ/kg) 13,100 kJ/kg
LIC-12-0083 Enclosure Page 21 of 164 The following table assesses the range of algebraic fire model, aforementioned in response to RAI 03a, input values against their NUREG-1824 validation ranges, using selected normalized parameters. The normalized parameter definition and validation ranges are taken from Table 2-5 of NUREG-1 934, second draft report for public comment, dated July 2011.
Normalized Eti Validation FCS Comments Parameter quaion Range Range For large HRRs, the generically assumed fire diameter (0.61 m for transient fire sources and 0.34 m for all other fire sources) is unrealistically small. This conservative and simplifying assumption was made to preclude the need to speculate the actual fire diameters Q for each fire scenario. This approach is Fire Froude = 0.40-2.40 0.43-19.11 conservative since smaller assumed fire Number PocpT.D 2g-' diameters results in taller flame heights and more aggressive plume temperature profiles.
Note that applying the 9 8 th percentile heat release rates specified by NUREG/CR-6850 Table G-1 to many typical electrical cabinet vent areas will similarly exceed the Froude number validation range.
The FCS fire PRA calculates severity factors Lf based on plume temperature and ceiling jet H profiles, which bound severity factors Flame Length calculated by flame lengths.
relative to 0.20 -1.00 N/A Ceiling Height Lf Similarly, the FCS fire PRA calculates zones of
= 3.7Q 2 /5 - 1.02 influence based on plume temperature, ceiling D jet, and flame radiation, which bound zones of influences associated with flame length.
LIC-1 2-0083 Enclosure Page 22 of 164 The FCS fire PRA exceeds the validation range at both the low and high ends.
Exceeding the range on the low end indicates a small critical ceiling jet radius relative to the ceiling height. As the non-dimensional parameter decreases, the critical ceiling jet radius approaches the plume centerline, for a fixed ceiling height. This effect does not impact the FCS fire PRA, which assumes all targets are failed within a cylindrical volume surrounding the ignition source, whose radius Ceiling Jet is defined by the radiant heat flux required to Radial Distance 7j 1.20- 1.70 0.10-3.15 cause target damage that is conservatively H-relative to applied from floor to ceiling. This conservative Ceiling Height approach bounds cases where the validation range is exceeded on the low end.
Exceeding the range on the high end is a result of the relatively high heat release rates specified by NUREG/CR-6850 Table G-1 and the conservatively low ceiling height (three meters) generically assumed as a simplification by the FCS fire PRA. Because exceeding the validation range is a result of conservative input parameters, it is not expected to non-conservatively change the results and conclusions of the NFPA 805 application.
LIC-12-0083 Enclosure Page 23 of 164
-I 1 T The effect of mechanical ventilation is conservatively not modeled by the FCS fire PRA.
Regarding natural ventilation, for fire scenarios with heat release rates up to 1,OOOkW, the FCS fire PRA equivalence ratios are within the validation range (or reasonably close, as the Mech: N/A equivalence ratio is no higher than 0.66).
-AH 0o2 7hO2 However, for postulated heat release rates Equivalence above 1,000 kW, the equivalence ratio exceeds Ratio as an 0.04 - 0.60 the validated range, and this primarily is due to indicator of the i = 0.23 X1 Aoi* (Natural) generically assuming a one square meter Natural: 0.14-Ventilation Rate ventilation opening area (Ao). This assumption 0.23 poo% (Mechanical) 1.00 is generally conservative as smaller ventilation opening areas yield higher predicted hot gas layer temperatures. This approach also conservatively neglects the temperature reducing effect of forced ventilation.
Furthermore, the analysis is conservative because, for high heat release rate scenarios where this approach indicates the fire may be ventilation limited, the resulting reduced rate of combhustion is conservatively not credited.
I I combustionis conservativelynotcredited.
LIC-12-0083 Enclosure Page 24 of 164 Most (just over half) of the compartments in which fire modeling was performed (Group 'A')
fell within the validated range of compartment aspect ratios.
Some compartments (Group 'B') fell outside the validated range, with their geometries trending toward corridors rather than cubes. The aspect ratio is relevant to FCS fire PRA calculation of Compartment L W A: 1.16 - 4.88 the time to hot gas layer formation using the H and H1 0.60- 5.70 "MQH" model. As compartment geometry Aspect Ratios B: 0.30- 17.21 approaches a corridor, time to hot gas layer formation becomes delayed by the time required for transport of hot gases from the fire origin to the compartment boundaries (in particular along the "long" dimension). In this application, use of the "MQH" model outside its validated range of aspect ratios introduces conservatism as this delay time is conservatively not quantified or credited by the time to hot gas layer (and therefore time available for suppression) calculations.
This non-dimensional parameter is relevant to the FCS calculation of zone of influence due to flame radiation. The FCS input parameters are generally within the validated range, although Radial Distance the range is slightly exceeded on both the r" lower and upper ends. This slight deviation is relative to Fire 15 2.2-5.7 1.77- 6.20 not considered significant in the overall context Diameter of the source-target data collection and fire PRA. Note the reported FCS range is associated with the 9 8 th percentile heat release rates, which were used for source-target identification (i.e., the 2 5 th, 5 0 th, and 7 5 th values were not used for scenario develoDment).
LIC-12-0083 Enclosure Page 25 of 164
- c. Attachment J of the Transition Report states that National Institute of Standards and Technology (NIST) FDS Version 5 was used to assess MCR habitability:
- i. For V&V, reference is made to NUREG-1824. However, NUREG-1824 provides V&V of FDS Version 4, while FDS Version 5 was used.
Please explain why the V&V reports developed by NIST for FDS Version 5 were not included in the V&V basis described in Attachment J of the Transition Report.
OPPD's Response to Fire Modeling RAI 03 c.i.:
Excluding the following NIST FDS Version 5 verification and validation reports from the LAR, Table J-1, V & V Basis, was an oversight:
- NIST Special Publication 1018-5 "Fire Dynamics Simulator (Version 5) Technical Reference Guide Volume 2: Verification",
October 29, 2010.
- NIST Special Publication 1018-5 "Fire Dynamics Simulator (Version 5) Technical Reference Guide Volume 3: Validation",
October 29, 2010.
These reports document a comprehensive assessment of the FDS Version 5 mathematical modeling and encoding (verification) as well as comparison of predicted results against a variety of experimental data (validation). These documents were added to the LAR Attachment J, Table J-1 under the column entitled V & V Basis, and will be included in the NFPA 805 transition LAR supplement. [AR 48249].
ii. Please provide technical details to demonstrate that FDS has been applied within the validated range of input parameters, or to justify the application of the model outside the validated range reported in NUREG-1824 and the aforementioned NIST reports.
OPPD's Response to Fire Modeling RAI 03 c.ii.:
The fire modeling input parameters used by the FCS fire PRA have been either demonstrated to be within the validated range of input parameters or their use has been justified.
LIC-12-0083 Enclosure Page 26 of 164 Input Range for Notes FCS 0 (HRR (kW)) 100- 1,000
- p. (ambient density of air Note that this term is not an explicit input to the (kg/m 3)) 1.2 Fire Dynamics Simulator model for the main control room analysis.
cp (specific heat of air 1.01 Note that this term is not an explicit input to the (kJ/kg/K)) Fire Dynamics Simulator model for the main control room analysis.
Default value used by Fire Dynamic Simulator T. (ambient air 293 applied to FCS main control room analysis.
temperature (K)) Default value specified by NIST Special Publication 1019-5 "Fire Dynamics Simulator (Version 5) User's Guide, April 8, 2009.
g (acceleration of gravity Note that this term is not an explicit input to the (m/s 2)) 9.81 Fire Dynamics Simulator model for the main control room analysis.
D (fire diameter (m)) 1.13 Converted from one square meter burner surface area specified by Fire Dynamics Simulator input file Lf (flame length (in)) N/A No targets in flame region.
rcj (ceiling jet radial N/A distance (m)) No targets in ceiling jet region.
A, (area of the opening (M2)) N/A See discussion of equivalence ratio in below table.
H, (height of the opening N/A See discussion of equivalence ratio in below (m)) table.
LIC-1 2-0083 Enclosure Page 27 of 164 Input Range for Notes FCS Vdot (ventilation rate N/A (m3/sec)) Mechanical ventilation is conservatively not modeled.
L (length of compartment 24.4 i(m))
W(width of compartment 21.6 (m))
H (height of compartment 6.4 (m))
The "target"for the MCR abandonment r (radial distance (in)) N/A scenario is the operator in relation to the hot upper layer.
AH 43,600
LIC-12-0083 Enclosure Page 28 of 164 The following table assesses the range of Fire Dynamics Simulator Version 5 input values against their NUREG-1824 validation ranges, using selected normalized parameters, for the main control room fire model. The normalized parameter definition and validation ranges are taken from Table 2-5 of NUREG-1 934, second draft report for public comment, dated July 2011.
Normalized Equation Validation FCS Comments Parameter Range Range The FCS model is within the validated range for scenarios above 600 kW and beneath the validated range for scenarios below 600 kW. This is a result of specifying the fire source area to be one square meter, generically over the range of Q_ heat release rates modeled. The assumed Fire Froude Q*r pocpTooD2,-f. 0.40 - 2.40 0.07-0.66 fire source area is unrealistically large for heat release rates less than 600 kW.
This abnormality in the fire geometry is judged not to significantly affect the predicted abandonment times, which are a function of upper layer temperature, heat flux, and smoke density, all of which should minimally be affected by the fire geometry.
Lf H
Flame Length 0.20-1.00 N/A relative to There are no targets in the flame region.
Ceiling Height Lf = 3.7Q 2 /5 1.02 D
Ceiling Jet Radial Distance H-- 1.20 - 1.70 N/A There are no targets in the ceiling jet relative to region.
Ceiling Height
LIC-12-0083 Enclosure Page 29 of 164 Normalized Equation Validation FCS Comments Parameter Range Range The fire heat release rate profiles for each modeled scenario has been verified in the Fire Dynamics Simulator output files to be consistent with those specified in the respective input files. This indicates that
ýP = the modeled fire scenarios are not Equivalence ZH 02 mo 2 ventilation-limited (i.e.,,p <1.0). Since the Ratio as an 1 0.04 - 0.60 N/A Fire Dynamics Simulator uses a simplified indicator of the . 0.23 x A 0 0 (Natural) combustion model that "turns off" the fire Ventilation Rate h0 2 = 2 when oxygen concentration reduces 0.23 poV (Mechanical) beneath a pre-determined threshold, and the scenarios modeled by the FCS fire PRA are not ventilation-limited, the equivalence ratio does not impact the Fire Dynamics Simulator results as applied to the FCS fire PRA.
L W Compartment - and - 0.60 - 5.70 3.81 FCS main control room analysis is within Aspect Ratio H H the validated range.
Radial Distance r The "target"for the MCR abandonment relative to Fire 2.2 - 5.7 N/A scenario is the operator in relation to the Diameter I hot upper layer.
LIC-1 2-0083 Enclosure Page 30 of 164
- d. FDS was also used to model an instrument air compressor oil fire scenario in Fire Area FC32/Room 19. This is not mentioned in Attachment J of the Transition Report. Please provide technical details for this scenario to demonstrate that FDS has been applied within the validated range of input parameters, or to justify the application of the model outside the validated range reported in NUREG-1824 and the aforementioned NIST reports.
OPPD's Response to Fire Modeling RAI 03 d.:
The fire modeling input parameters used by the FCS fire PRA have been either demonstrated to be within the range of applicability for fire models implemented or their use has been justified if outside the range of applicability.
The following table documents the Fire Dynamics Simulator Version 5 input values and their ranges as implemented by the FCS fire PRA assessment of the FC32 instrument air compressor lube oil fire scenarios.
Range FCS Reference Input for (HRR (kW)) 1,005 CN-RAM-1 0-013 Attachment 14 CN-RAM-09-045 Section 4.4.4 (CN-RAM-10-013 Table
- p. (ambient density 1.2 4-12). Note that this term is not an explicit input to the of air (kg/m 3)) Fire Dynamics Simulator model for the main control room analysis.
CN-RAM-09-045 Section 4.4.4 (CN-RAM-10-013 Table cp (specific heat of air 1.01 4-12). Note that this term is not an explicit input to the (kJ/kg/K)) Fire Dynamics Simulator model for the main control room analysis.
Default value used by Fire Dynamic Simulator applied T., (ambient air to FCS main control room analysis. Default value temperature (K)) 293 specified by NIST Special Publication 1019-5 "Fire Dynamics Simulator (Version 5) User's Guide, April 8, 2009.
CN-RAM-09-045 Section 4.4.4 (CN-RAM-10-013 Table g (acceleration of 4-12). Note that this term is not an explicit input to the gravity (m/s2)) 9.81 Fire Dynamics Simulator model for the main control room analysis.
Converted from 0.56 square meter total oil fire surface D (fire diameter (m)) 0.84 area specified by Fire Dynamics Simulator input file (CN-RAM-1 0-013 Attachment 14)
L,(flame length (m)) N/A No targets in the flame region.
No targets in ceiling jet region. Cable Trays 3 and 4 S(ceiling jet radial N/A (Reference Fire Dynamics Simulator input file in d(istangjet ri N/A Attachment 14 to CN-RAM-1 0-013) are effectively in distance (in)) the plume region, while the remaining cable trays are effectively in the hot gas layer region.
LIC-1 2-0083 Enclosure Page 31 of 164 A,, (area of opening the (M2)) N/A See equivalence ratio discussion in the following table.
H, (height of the N/A See equivalence ratio discussion in the following table.
opening (m))
Mechanical ventilation is conservatively not modeled (M3V/sec)) N/A per Fire Dynamics Simulator input file (CN-RAM 013 Attachment 14)
Fire Dynamics Simulator input file (CN-RAM-10-013 Attachment 14). Note that the model included the vicinity of the compressors and conservatively excluded L (length of 10.8 a large volume of the room (which would dissipate hot compartment (m)) gas above the compressors), and this simplification was primarily made due to computational limitations associated with modeling a very large control volume with the Fire Dynamics Simulator.
W(width of 10.4 Fire Dynamics Simulator input file (CN-RAM-10-013 compartment (m)) Attachment 14)
H (height of 6.1 Fire Dynamics Simulator input file (CN-RAM-10-013 compartment (m)) Attachment 14)
Parameter relevant to point-source flame radiation heat r (radial distance (m)) N/A flux model.
AH N/A See equivalence ratio discussion in the following table.
LIC-1 2-0083 Enclosure Page 32 of 164 The following table assesses the range of Fire Dynamics Simulator Version 5 input values against their NUREG-1824 validation ranges, using selected normalized parameters, for the instrument air compressor fire model. The normalized parameter definition and validation ranges are taken from Table 2-5 of NUREG-1934, second draft report for public comment, dated July 2011.
Normalized Validation FCS Parameter Equation Range Range Comments Fire Froude _ Q The FCS model is within the validated
-* 0.40 -2.40 1.40 Number p.cpT.DD2Jf' range.
Lf No targets in the flame region. Cable Trays Flame Length H 3 and 4 (Reference Fire Dynamics relative to 0.20 - 1.00 N/A Simulator input file in Attachment 14 to CN-RAM-10-013) are effectively in the plume Ceiling Height Lf 3.7Q 2 /5 . 1.02 region, while the remaining cable trays are D effectively in the hot gas layer region.
No targets in ceiling jet region. Cable Ceiling Jet Trays 3 and 4 (Reference Fire Dynamics Radial Distance Tci 1.20-1.70 N/A Simulator input file in Attachment 14 to CN-relative to H-/ RAM-10-013) are effectively in the plume Ceiling Height region, while the remaining cable trays are effectively in the hot gas layer region.
LIC-12-0083 Enclosure Page 33 of 164 Normalized Validation FCS Equation Range Range Comments Parameter The fire heat release rate profile has been verified in the Fire Dynamics Simulator output files to be consistent with those specified by the input file. This indicates that the modeled fire scenario is not e (ventilation-limited (i.e., qp <1.0). Since the Equivalence = Fire Dynamics Simulator uses a simplified Ratio as an 1A.-0° 2m° 2 combustion model that "turns off" the fire indicator of the 0.23 x 1A0.4-0 (Natural) when oxygen concentration reduces Ventilation Rate in0 2 -2 beneath a pre-determined threshold, and 0.23 pofV (Mechanical) the instrument air compressor oil fire scenario modeled by the FCS fire PRA is not ventilation-limited, the equivalence ratio does not impact the Fire Dynamics Simulator results as applied to the FCS fire PRA.
Compartment L and W
-- 0.60 -5.70 1.70 -1.77 The FCS model is within the validated Aspect Ratio H H range.
Radial Distance r Parameter relevant to point-source flame relative to Fire 2.2 - 5.7 N/A Diameter D radiation heat flux model.
LIC-1 2-0083 Enclosure Page 34 of 164 Fire Modeling RAI 04:
NFPA 805, Section 2.7.3.4, "Qualification of Users," states: "Cognizant personnel who use and apply engineering analysis and numerical models (e.g., fire modeling 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."
Section 4.5.1.2 of the Transition Report states that fire modeling was performed as part of the FPRA development (NFPA 805, Section 4.2.4.2). This requires that qualified fire modeling and PRA personnel work together. Furthermore, Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," of the Transition Report states that "For personnel performing fire modeling or FPRA development and evaluation, OPPD develops and maintains qualification requirements for individuals assigned various tasks. Position specific training 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."
Regarding qualifications of users of engineering analyses and numerical models:
- a. Please describe what constitutes the appropriate qualifications for the OPPD staff and consulting engineers to use and apply the methods and fire modeling tools included in the engineering analyses and numerical models.
OPPD's Response to Fire Modeling RAI 04 a.:
Personnel who perform fire modeling in support of the FCS NFPA 805 program must 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" in accordance with NFPA 805 Section 2.7.3.4. Example attributes of a qualified fire model user include an engineering degree, formal training or experience in fire behavior and fire modeling techniques, and experience performing or supporting nuclear power plant fire modeling applications.
- b. Please describe the process/procedures for ensuring adequate qualification of the engineers/personnel performing the fire analyses and modeling activities.
OPPD's Response to Fire Modeling RAI 04 b.:
The fire PRA was performed by contract personnel under the Westinghouse Electric Company QMS and will be maintained under the OPPD Quality Assurance Program, both of which provide specific requirements for ensuring that work is performed by qualified personnel.
LIC-1 2-0083 Enclosure Page 35 of 164
- c. Please explain how the necessary communication and exchange of information between fire modeling analysts and FPRA personnel was accomplished.
OPPD's Response to Fire Modeling RAI 04 c.:
The FCS fire PRA was developed by PRA analysts who were also qualified fire modeling users. The fire PRA analysts performed the fire modeling. This approach inherently assures that the boundary conditions for PRA and fire modeling are compatible.
Fire Modeling RAI 05:
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."
Section 4.7.3, "Compliance with Quality Requirements in Section 2.7.3 of NFPA 805," of the Transition Report states that "Uncertainty analyses were performed as required by Section 2.7.3.5 of NFPA 805 and the results were considered in the context of the application. This is of particular interest in fire modeling and FPRA development."
Regarding the uncertainty analysis:
- a. Please explain the uncertainty analyses for fire modeling that was performed. Describe how the uncertainties of the input parameters (compartment geometry, HRR, radiative fraction, etc.) were determined. In addition, please substantiate the statement in Appendix J of the Transition Report that states, " ...the predictions are deemed to be within the bounds of experimental uncertainty... "
OPPD's Response to Fire Modeling RAI 05 a.:
Fire modeling sources of uncertainty were identified by reviewing each fire PRA calculation and supplemented by the generic sources of uncertainty in NUREG/CR-6850, Appendix V. Each source of uncertainty is either qualitatively or quantitatively dispositioned. The following paragraphs briefly discuss sources of uncertainty relevant to fire scenario development.
There is some uncertainty associated with the ability of compartment boundaries to confine fire (i.e., random failure of fire barriers, random failure of active suppression systems credited as partitioning feature, etc.). These compartment boundary failures are modeled quantitatively in the multi-compartment analysis, and uncertainty associated with barrier failure probabilities is judged to negligibly impact the overall conclusions of the fire PRA as applied to NFPA 805 due to the low frequency of severe fires physically capable of multi-compartment propagation.
LIC-1 2-0083 Enclosure Page 36 of 164 Ignition source frequencies modeled by the FCS fire PRA are based on generic fire frequencies developed from nuclear industry fire event data. A review of FCS fire events has been performed, and no outlier experience was identified, suggesting the generic frequencies are applicable to FCS. There is some data uncertainty associated with the generic frequencies themselves, and they are therefore provided as frequency distributions. The FCS fire PRA implements the mean values of these distributions.
The heat release rates modeled by the FCS fire PRA are a large source of uncertainty and a potentially significant source of conservatism. The conservative treatment of heat release rates minimizes the non-conservative potential impact of this uncertainty. Scenario peak heat release rate values are provided as gamma distributions by NUREG/CR-6850, and the full spectrum of these distributions is modeled (up to the 98th percentile) for scenarios implementing the severity factor concept. A sensitivity study assessing the impact of the severity factor concept as applied to the FCS fire PRA was performed. While it is not practical to perform a Monte Carlo sampling study on heat release rate uncertainty impact on target identification, and therefore CCDP and CLERP calculations, the potential impact of this uncertainty on the overall conclusions of the fire PRA as applied to NFPA 805 is again minimized by conservatisms in heat release rate selection.
Manual non-suppression probabilities are approximated as point estimates as a function of time available for suppression. While generic probability distributions are not available for manual non-suppression probability, its uncertainty is judged to not significantly impact the overall conclusions of the fire PRA as applied to NFPA 805 due to the low frequency of severe fires in which the FCS fire PRA credits manual suppression (primarily control room scenarios that could lead to abandonment and ex-control room scenarios leading to hot gas layer).
Automatic suppression system failure probabilities are similarly provided as point estimates by NUREG/CR-6850. The fire PRA identified the switchgear room Halon system as an important risk-reducing system for the plant, and a sensitivity study postulating lower and upper bound failure probabilities for this system was performed.
Uncertainty associated with initial fire modeling treatment of the FCS instrument air compressors was identified to have a potentially significant impact on the results and conclusions of the fire PRA. The conclusions for the importance of this scenario are consistent with earlier PRA analysis that prompted the installation of an oil collection system to improve safety. As a result, a more detailed fire modeling assessment of the air compressors was performed using the FDS. While this more detailed analysis confirmed the original fire modeling assessment that overhead cables would not be damaged, the relevant scenarios were subdivided into two cases, one in which overhead cables are damaged at a low likelihood, and the other in which overhead cables are not damaged at a higher likelihood. This approach considers a more complete spectrum of outcomes in consideration of the model and parameter uncertainty associated
LIC-1 2-0083 Enclosure Page 37 of 164 with the fire modeling assessment.
In addition, response to Fire Modeling RAI 05 b. includes sensitivity studies assessing electrical cable thermophysical property uncertainty. These sensitivity studies include cable heat release rate per unit area, cable heat of combustion, and cable insulation density. A qualitative assessment of cable damage temperature and heat flux is also provided.
Finally, in response to the final question of Fire Modeling RAI 05 b., the LAR Appendix J Table J-1 statements regarding predictions generally being within experimental uncertainty are related to FDS calculation of hot gas layer temperature and heat flux. These statements were based on the NUREG-1824 Chapter 7 FDS model validation. Regarding hot gas layer temperature, the NUREG summarizes that, 'The FDS predictions of HGL temperature and height are, with a few exceptions, within experimental uncertainty." Regarding heat flux, the NUREG summarizes that, "FDS predictions of heat flux and surface temperature are generally within experimental uncertainty..." The referenced statement in Table J-1 was intended to be general in nature.
- b. Cables are the primary target in FPRA. The FPRA peer review team raised the question of uncertainty due to these target's thermophysical properties, and their importance to FPRA results. Please provide technical justification for the statement in the FSSR (FC07823 and CN-RAM-1 0-013) that "thermophysical property uncertainty is not expected to significantly affect the PRA results." In addition, provide a quantitative assessment of the impact of this uncertainty on the FPRA results.
OPPD's Response to Fire Modeling RAI 05 b.:
The cable thermophysical properties utilized by the FPRA include:
- Damage temperature
- Damage radiative heat flux
- Heat release rate per unit area
- Heat of combustion
" Cable insulation density The fire PRA assumes the damage temperature of 330 0 C and heat flux of 11 kW/m 2 as specified by NUREG/CR-6850 for thermoset cable. OPPD has a high degree of confidence that primarily cables with thermoset insulation are installed at FCS. The thermoset failure thresholds are provided by NUREG/CR-6850 as point estimates, as opposed to distributions, which the NUREG provides for parameters with significant uncertainty (e.g., heat release rate). If the actual failure thresholds were larger than prescribed, then the fire PRA would be conservative. While it is theoretically possible that the actual failure thresholds could be smaller than prescribed, OPPD has a high degree of confidence that the NUREG/CR-6850 thresholds are conservative and their application results in conservative ZOls, especially when compared to the actual fire event data upon which the fire frequencies are based.
LIC-12-0083 Enclosure Page 38 of 164 The cable tray fire propagation analysis, used for assessing the potential for hot gas layer formation, assumes a heat release rate per unit area of 328 kW/m 2, which is the average bench scale value for thermoset insulation types (XPE/FRXPE, XPE/Neoprene, and XPE/XPE) provided in Table R-1 of NUREG/CR-6850. The values range from 178 kW/m 2 to 475 kW/m 2 in Table R-1 of NUREG/CR-6850. A sensitivity study was performed by varying the heat release rate per unit area between 178 kW/m 2, 328 kW/m2, and 475 kW/m2, and this study concluded:
" CDF and LERF results remain unchanged between the 178 kW/m 2 and 328 kW/m 2 cases.
- CDF and LERF increased for the following five scenarios between the 328 kW/m 2 and 475 kW/m 2 cases, as specified in the following table. These increases are due to these scenarios creating the potential for hot gas layer formation if 475 kW/m 2 is assumed; whereas there is no potential for hot gas layer formation if 328 kW/m 2 is assumed.
328 kW/m 2 475 kW/m 2 Scenario CDF (/yr) LERF (/yr) CDF (/yr) LERF (/yr)
FC41-PW-CFWC-4 6.71 E-08 3.54E-08 2.14E-07 4.60E-08 FC41-PW-TFWC-3 3.28E-10 3.28E-10 9.37E-10 3.72E-10 FC41-PW-TFWC-4 3.67E-10 2.65E-1 1 9.37E-10 6.76E-1 1 FC41-PW-TRAN-3 '1.59E-09 1.59E-09 4.56E-09 1.81 E-09 FC41-PW-TRAN-4 1.79E-09 1.29E-10 4.56E-09 3.29E-10 TOTAL 7.12E-08 3.75E-08 2.25E-07 4.86E-08 The net CDF and LERF increases for FC41 between the 328 kW/m 2 and 475 kW/m 2 cases are 1.54E-07 /yr and 1.11 E-08 /yr, respectively. The following table extends this sensitivity study for the overall plant risk.
Base Fire PRA* Sensitivity Study**
(328 kW/m 2 Case) (475 kW/m 2 Case)
Net ACDF for NFPA 805 5.72E-06 5.87E-06 Transition (/yr)
Net ALERF for NFPA 805 6.67E-07 6.78E-07 Transition (/yr)
Total CDF (internal, flood, 6.01 E-05 6.03E-05 fire) (/yr)
Total LERF (internal, flood, 4.82E-06 4.83E-06 fire) (/yr)
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
LIC-1 2-0083 Enclosure Page 39 of 164 In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, for the range of cable heat release rates per unit area assessed by this sensitivity study.
The cable tray fire propagation analysis, used for assessing the potential for hot gas layer formation, assumes a cable heat of combustion of 12,000 BTU/Ib. The heat of combustion is specifically used to calculate cable tray burnout time (i.e.,
extinction due to consumption of all fuel in the tray). The assumed value of 12,000 BTU/Ib is specified and used by the FCS Combustible Loading Calculation, which also identifies a range of possible values from 9,500 to 13,250 BTU/Ib. A sensitivity study was performed by varying the heat of combustion between 9,500, 12,000, and 13,250 BTU/Ib, and this study concluded:
" The resulting cable tray burnout times varied between 8, 10, and 11 minutes for the 9,500, 12,000, and 13,250 BTU/Ib cases, respectively.
Cable tray burnout times increase with increasing assumed heats of combustion, since a higher heat of combustion indicates a higher energy content per unit mass, which requires more time to burn completely, assuming a constant mass burning rate.
- Varying the heat of combustion over the specified range did not change the resulting CDF and LERF values for any scenario. The fire-induced CDF and LERF are relatively insensitive to assumed heat of combustion.
This is reasonable since slight changes in the cable tray burnout time (for which the assumed heat of combustion affects) will alter slightly the growth profile but will usually not change the peak heat release rate that a given scenario achieves.
" The NFPA 805 acceptance criteria are met over the range of heats of combustion assessed by this sensitivity study.
The cable tray fire propagation analysis, used for assessing the potential for hot gas layer formation, assumes a cable insulation and jacket density of 0.02 pounds per linear foot of cable. This cable insulation/jacket density is specifically used to calculate cable tray burnout time (i.e., extinction due to consumption of all fuel in the tray). The assumed value of 0.02 lb/ft is specified and used by the FCS combustible loading calculation.
For the sensitivity study, the cable insulation/jacket densities for a variety of cables listed in Table 9-B of NUREG/CR-7010 (Draft Report for Comment, issued October 2010) were reviewed and summarized in the following table. The insulation/jacket densities ranged from 0.07 to 0.60 lb/ft, with an average value of 0.19 lb/ft. Note that the 0.60 lb/ft for Cable ID 11 appears to be a high outlier.
Nonetheless, the lower end of this range is a factor of 3.5 higher than the value assumed for the FCS FPRA.
LIC-1 2-0083 Enclosure Page 40 of 164 Cable Plastic Mass Plastic (kg/m) Fraction (kg/kg) (Ib/ft) 11 1.985 0.45 0.60 16 0.671 0.48 0.22 23 0.253 0.69 0.12 43 0.357 0.62 0.15 219 0.296 0.61 0.12 269 0.24 0.53 0.09 367 0.441 0.73 0.22 700 0.322 0.33 0.07 701 0.366 0.42 0.10 Average 0.55 0.54 0.19 A sensitivity study was performed by varying the cable insulation/jacket density between 0.02, 0.19, and 0.60 lb/ft, and this study concluded:
" The resulting cable tray burnout times varied between 10, 97, and 307 minutes for the 0.02, 0.19, and 0.60 lb/ft cases, respectively. Cable tray burnout times increase with increasing assumed cable insulation/jacket density, since a higher densities a higher energy content per unit mass, which requires more time to burn completely, assuming a constant mass burning rate. The tray burnout time is sensitive to the assumed cable insulation/jacket density.
" CDF and LERF increased slightly for the following four scenarios between the 0.02, 0.19, and 0.60 lb/ft cases, as specified in the following table.
These slight increases are due to these scenarios creating the potential for hot gas layer formation as higher cable insulation/jacket densities are assumed; whereas there is no potential for hot gas layer formation with the 0.02 lb/ft assumed for the base fire PRA.
0.02 lb/ft 0.19 lb/ft 0.60 lb/ft Scenario CDF (/yr) LERF (/yr) CDF (/yr) LERF (/yr) CDF (/yr) LERF (/yr)
FC41-PW- 3.28E-10 3.28E-10 3.33E-10 3.28E-10 3.33E-10 3.28E-10 TFWC-3 FC41-PW- 3.67E-10 2.65E-11 8.69E-10 6.28E-11 8.69E-10 6.28E-11 TFWC-4 FC41-PW- 1.59E-09 1.59E-09 1.62E-09 1.60E-09 1.62E-09 1.60E-09 TRAN-3 FC41-PW- 1.79E-09 1.29E-10 4.23E-09 3.05E-10 4.23E-09 3.05E-10 TRAN-4 TOTAL 4.08E-09 2.08E-09 7.05E-09 2.29E-09 7.05E-09 2.29E-09
- Note that the fire PRA supporting the LAR conservatively modeled the cable tray stack above FC20-1-1S18 as containing five trays due to difficulties in verifying the configuration at the time analysis was performed. The revised sensitivity study models the stack as containing three trays. This is based on plant walkdowns and additional research conducted in response to this RAI.
LIC-12-0083 Enclosure Page 41 of 164 The net CDF and LERF increases for FC41 between the 0.02 lb/ft (assumed by base fire PRA) and upper bound 0.60 lb/ft cases are 2.97E-09 /yr and 2.10E-10 /yr, respectively. The following table extends this sensitivity study for the overall plant risk.
Base Fire PRA* Sensitivity Study** Sensitivity Study**
(0.02 lb/ft Case) (0.19 lb/ft Case) (0.60 lb/ft Case)
Net ACDF for NFPA 805 5.72E-06 5.72E-06 5.72E-06 Transition (/yr)
Net ALERF for NFPA 805 Trniin(y)6.67E-07 6.67E-07 6.67E-0)7 Total CDF (internal, 6.01 E-05 6.01 E-05 6.01 E-05 flood, fire) (/yr)
Total LERF (internal, 4.82E-06 4.82E-06 4.82E-06 flood, fire) (/yr) I I
- Base Fire PRA results as reported in Section W.2 of LIC-11-0099.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, for the range of cable insulation/jacket densities assessed by this sensitivity study.
C. It appears that in many cases the most conservative model input parameter values were not used. Two specific examples are described below:
- i. The soot yield in the FDS analysis for MCR habitability study was assumed to be 0.06 while higher values are reported in the fire protection literature for the same type of cables. Please provide a summary report of the impact of the uncertainty of the soot yield on the Fire PRA results (i.e., perform a quantitative assessment).
OPPD's Response to Fire Modeling RAI 05 c.i.:
As discussed during the June 26, 2012, teleconference with the NRC technical reviewers and subsequently agreed upon via email on July 3, 2012, OPPD will submit its response to this RAI on September 28, 2012.
[AR 52508]
ii. A HRR per unit area (HRRPUA) of 328 kW/m 2 was used to calculate the HRR of cable tray fires in FDS analysis to model an air compressor oil fire scenario in Fire Area FC32/Room 19. This value is the average of all values for thermoset cables listed in Table R-1 of NUREG/CR-6850.
However, the exact composition of the cables in each tray is not known and a tray might be filled with cables that have a higher HRRPUA. In addition, the licensee assumed that each tray contained 50 cables and that the weight of insulation and jacket is 0.02 pounds
LIC-1 2-0083 Enclosure Page 42 of 164 per foot (lb/ft). During the audit walkdowns, the NRC staff found numerous trays that had more than 50 cables. The cables in these trays appeared to contain more insulation than what was assumed.
Please provide a summary report of the impact of the uncertainty of the HRRPUA and combustible loading of the cables on the Fire PRA results (i.e., perform a quantitative assessment).
OPPD's Response to Fire Modelinq RAI 05 c.ii.:
Response to Fire Modeling RAI 05 b. documents a sensitivity study regarding the assumed cable tray heat release rate per unit area of 328 kW/m 2 . Note that this parameter is used in calculating the overall fire growth profiles for scenarios modeled using the ZOI approach. This parameter is not used in the FDS evaluation of the instrument air compressor oil fire scenarios. The purpose of the FDS is to determine if the overhead cables would fail due to the oil fire exposure and not to quantify the heat release rate contribution of the cables should they ignite.
Response to Fire Modeling RAI 05 b. documents a sensitivity study regarding the assumed 0.02 pounds of cable insulation per linear foot.
The assumed number of cables per tray is used to calculate cable tray burnout time, which affects the overall fire scenario growth profile. The burnout time increases as the number of cables increases. The current fire PRA assumes 50 cables per tray.
A review of the FCS cable management database, referred to as FACTS, provided the following characterization of the number of cables installed per tray at FCS:
- Average = 26
- Median=12
- Minimum = 1
- Maximum =345 The following figure plots the number of cable trays per tray, assuming the data is normally distributed.
LIC-1 2-0083 Enclosure Page 43 of 164 Number of Cables per Tray (Assumed Data Normally Distributed) 0.014 0.012-0.008-0.004 ____
0~o-* ----- - ..... -- ---.
Jj .
0.002 0 20 40 60 80 100 120 140 160 Number of Cables per Tray The 98th percentile of this distribution is 96 cables per tray. As a sensitivity study, the fire PRA model was re-quantified assuming 96 cables per tray, and the total fire-induced CDF and LERF remained unchanged from the fire PRA quantification reported in the LAR. The NFPA 805 acceptance criteria would therefore still be met, within the same Regulatory Guide 1.174 regions, if the 98th percentile (96 cables per tray) were assumed for all fire modeled scenarios involving cable trays.
- d. Please provide justification for the assumption that in the Fire Area FC321 Room 19 compressor area (only) 10 percent of the compressor oil fires result in failure of the overhead cable trays while, based on the FDS calculations, the overhead cable trays in 90 percent of the fires are not damaged.
OPPD's Response to Fire Modeling RAI 05 d.:
The fire PRA uncertainty analysis included a sensitivity study of the initial fire PRA treatment of the instrument air compressor oil fire scenarios. This initial treatment qualitatively assumed that the compressor oil fire scenarios (each involving seven gallons of lubricating oil) were not capable of damaging the overhead cable trays, and this judgment was based on the oil collection system and the distance from the compressor pedestal to the overhead cable trays.
Because failure of the overhead cable trays has the potential to broadly impact the electrical distribution system, a sensitivity study was performed in which it was assumed the compressor oil fire scenarios could damage the overhead trays, and this study indicated the overall results of the initial fire PRA were very sensitive to the simplified modeling assumptions initially made. As a result of this sensitivity study, a more detailed fire model using the computational fluid dynamics code FDS was performed.
LIC-1 2-0083 Enclosure Page 44 of 164 The results of the FDS study confirm the original assumption that overhead cable trays will not reach their failure threshold. However, as a result of this study identifying a large CDF sensitivity, and the known model and parameter uncertainties associated with fire modeling, the compressor large oil fire cases were each subsequently divided into two scenarios. In the 'a' scenario, 90% of the large oil fire frequency is assumed to not be capable of damaging overhead trays, consistent with the FDS prediction. In the 'b' scenario, 10% of the large oil fire frequency is assumed to be capable of generating environmental conditions exceeding the failure threshold of the cable trays, and those trays are therefore assumed failed if their automatic deluge system also fails. This approach has several benefits, including fire PRA postulation of a more complete spectrum of outcomes, accounting for the possibility, albeit remote, of an oil fire that cannot be confined by the oil collection system (e.g., atomize spray fire), and the consideration of model and parameter uncertainty associated with the above fire modeling calculation.'
- e. During the audit, the NRC staff reviewed Attachment 14 to the FSSR (FC07823 and CN-RAM-10-013). The staff noted that cable tray obstructions appeared to have been omitted in the FDS fire modeling analysis for Fire Area FC32/Room 19. Please justify why cable tray obstructions were not considered in the FDS fire modeling analysis for Fire Area FC32/Room 19.
OPPD's Response to Fire Modeling RAI 05 e.:
This RAI response will be addressed coincident with the response to Fire Modeling RAI 01 f.v.(a). Potentially re-run FDS simulations considering cable tray obstructions and assess impact on meeting NFPA 805 acceptance criteria (sensitivity study). As discussed during the June 26, 2012, teleconference with the NRC technical reviewers and subsequently agreed upon via email on July 3, 2012, OPPD will submit its response to this RAI on September 28, 2012. [AR 52508]
LIC-12-0083 Enclosure Page 45 of 164 Fire Protection Engineering RAI 01:
The LAR Table 4.3 and the fire hazards analysis (FHA) include fire area "50" ("outdoor gas storage") within their scope but the LAR Table I-1 (power block definition) does not.
The NRC staff understands that NEI 04-02, "Guidance for Implementing A Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," Rev. 2, Attachment K, was used; however, please provide justification for not having the outdoor gas storage area included within the power block definition under Attachment I.
OPPD's Response to Fire Protection Engineering RAI 01:
LAR Attachment I has been revised to add a discussion of the Outdoor Gas Storage, Fire Area
- 50. The following information has been added to the "Excluded Structures" section of Attachment I:
Fire Area 50, Outdoor Gas Storage, is not a structure and is not included in the Power Block. The area does not affect the nuclear safety or radiological release performance criteria. However, it is a plant FHA fire area and is included in the B-3 table. It is also included in the B-1 table based on the NFPA 50A code review performed for this area.
The updated LAR Transition Report Attachment I, with this revised information, will be included in the NFPA 805 transition LAR supplement. [AR 48249]
Fire Protection Engineering RAI 02:
Please provide further justification detailing why substation 1251 control building should not be considered within the power block since the 161 kiloVolt (KV) switchyard supplies the plant with offsite power.
OPPD's Response to Fire Protection Engineering RAI 02:
The substation #1251 control building located within the 161KV switchyard has been excluded from the power block in LAR Attachment I. Although the 161 KV switchyard (inclusive of #1251 control building) is contained within the owner controlled area, it is not contained within the protected area. The design authority for the switchyard does not reside within the FCS plant engineering organization(s).
Calibration Procedure EM-CP-06-FDZ-SWYD, "Calibration of Switchyard Buildings Fire Detection System" identifies that there is no annunciation for the 161 kV switchyard building fire detection system provided to the FCS main control room. The switchyard'building detection system actuates an alarm at the Energy Control Center (ECC). The offsite fire department is the first responder for fire alarms associated with the switchyard.
The NFPA 805 NSCA does include control, power, and metering cables from the power block structures up to the boundary of the 161 KV switchyard as necessary to support the analysis for availability of 161 KV offsite power to the station. However, the 161 KV switchyard (inclusive of
- 1251 control building) is not considered by FCS to be a structure within the scope of NFPA 805. Offsite power transmission lines, associated transmission equipment, and structures located outside of the owner controlled area also support offsite power for the NFPA 805 NSCA; however, these are similarly excluded from the power block.
LIC-1 2-0083 Enclosure Page 46 of 164 Fire Protection Engineering RAI 03:
Please provide further justification detailing why the 345 KV switchyard should not be considered within the power block since this switchyard carries the power output of the main generator as well as provides a portion of the supply to the plant for offsite power.
OPPD's Response to Fire Protection Enqineering RAI 03:
The 345KV switchyard has been excluded from the power block in LAR Attachment I.
Although the 345KV switchyard is contained within the owner controlled area, it is not contained within the protected area. The design authority for the switchyard does not reside within the FCS plant engineering organization(s).
Calibration Procedure EM-CP-06-FDZ-SWYD, "Calibration of Switchyard Buildings Fire Detection System" identifies that there is no annunciation for the 345KV switchyard buildings fire detection systems provided to the FCS main control room. The switchyard building detection system actuates an alarm at the Energy Control Center (ECC). The offsite fire department is the first responder for fire alarms associated with the switchyard.
The NFPA 805 NSCA does include control, power, and metering cables from the power block structures up to the boundary of the 345KV switchyard as necessary to support the analysis for availability of 161KV offsite power to the station. However, the 345KV switchyard is not considered by FCS to be a structure within the scope of NFPA 805. Offsite power transmission lines, associated transmission equipment, and structures located outside of the owner controlled area also support offsite power for the NFPA 805 NSCA; however, these are similarly excluded from the power block.
Fire Protection Engineering RAI 04:
The FHA, design basis document (DBD), and code compliance analysis (EC-FA-95-022) are missing reference codes that are in the LAR Table B-i, or missing the code edition as stated in the LAR. It is not clear if the LAR contains the complete list of reference codes required. It is not clear why there is a difference in codes referenced between the LAR and the other documents. Please clarify whether or not the LAR Table B-i, and Section 6.0 contain a complete and accurate list of required codes. Also, please clarify why there is a difference between the LAR Section 6.1, LAR Table B-i, DBD, and EA-FC-95-022.
OPPD's Response to Fire Protection Engineering RAI 04:
The LAR Transition Report Attachment A, Table B-1 and LAR Transition Report Section 6.0 contain the complete and accurate list of required codes for the NFPA 805 transition.
Implementation Item REC-102, "Development of the FCS NFPA 805 FP basis document (as described in Section 2.7.1.2 of NFPA 805)," is in place to update the fire protection-related documents to reflect the applicable codes of record.
LIC-1 2-0083 Enclosure Page 47 of 164 EA-FC-95-022 is the station's code compliance review and has performed reviews of codes that are not required by NFPA 805; therefore, these codes have been omitted from the LAR Table B-1 and Section 6.0. The design basis document PLDBD-NU-61 references several NFPA codes but does not provide the justification for inclusion in the design basis. These codes, included in PLDBD-NU-61 but not required for NFPA 805 transition, are not included in the LAR Transition Report Attachment A, Table B-1 or the LAR Transition Report Section 6.0.
The following table identifies the NFPA codes and editions, with conflicts noted between these documents and with the explanation for the difference:
NFPA _Code Edition EC-FA SectionB-16.0/
Table PLDBD-NU-61 022 NFPA 10 1994 No 1994 Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 12 1993 No 1993 Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 12A 1992 1985 1992 NFPA 12A-1985 edition is only mentioned in PLDBD. Per EA-FC 022, NFPA 12A-1992 is the appropriate code of record. PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 13 1996 1985 1996 NFPA 13-1985 edition is only mentioned in PLDBD. Per EA-FC 022, NFPA 13-1996 is the appropriate code of record. PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 13A N/A No N/A Code is only mentioned in PLDBD.
Edition However, compliance with this code is not required by NFPA 805.
NFPA 14 1996 No 1996 Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 15 N/A 1985 No Edition NFPA 15-1985 edition is only mentioned in PLDBD. LAR Table 4-3 did not identify any NFPA 15 systems required for NFPA 805 compliance, therefore, the code was not referenced.
LIC-1 2-0083 Enclosure Page 48 of 164 Code Edition NFPA Table B-1 I PLDBD- EC-FA Section 6.0 NU-61 022 NFPA 20 1996 No 1996 Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 24 1984 1984 1984 No conflicts.
NFPA 25 N/A N/A 1995 A previous code review was performed for NFPA 25 in EA-FC 022. However, the review is not required by NFPA 805 and is therefore not referenced in the LAR.
EA-FC-95-022 is the station's code compliance review and was not limited only to codes required by NFPA 805.
NFPA 26 N/A 1976 No Edition NFPA 26-1976 edition is only mentioned in PLDBD. However, compliance with this code is not required by NFPA 805.
NFPA 27 N/A No No Edition Code is only mentioned in PLDBD.
Edition EA-FC-95-022 discusses that compliance with NFPA 600 is demonstrated in the LAR in lieu of NFPA 27. PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 30 1966 / 1987/ 1987 1966 / 1987 Code edition in PLDBD conflicts with 2000 / 2008 / 2000 / additional code editions in LAR and 2008 EA-FC-95-022. PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 31 N/A 1987 N/A NFPA 31-1987 edition is only mentioned in PLDBD. However, compliance with this code is not required by NFPA 805.
NFPA 37 N/A 1984 N/A NFPA 37-1984 edition is only mentioned in PLDBD. However, compliance with this code is not required by NFPA 805.
NFPA 49 N/A No N/A Code is only mentioned in PLDBD.
Edition However, compliance with this code is not required by NFPA 805.
NFPA 50A 1969 No 1969 Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
LIC-1 2-0083 Enclosure Page 49 of 164 Code Edition Code Table B-1 / PLDBD- EC-FA Section 6.0 NU-61 022 NFPA 51 N/A No N/A Code is only mentioned in PLDBD.
Edition However, compliance with this code is not required by NFPA 805.
NFPA 51B 1999 No 1999 Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 70 N/A 1975/ 1996 NFPA 70-1975, 1984, and 1987 1984 / editions are only mentioned in 1987 PLDBD. However, compliance with this code is not required by NFPA 805.
NFPA 72 1990 /2002 N/A 1990 /2002 NFPA 72A-1985, 72D-1976, and
/ 2007 / 2007 72E-1984 are only mentioned in NFPA 72A N/A 1985 N/A PLDBD. Per EA-FC-95-022, these NFPA 72D 1975 / 1979 1976 1975 / 1979 codes are not applicable to the NFPA 72E 1978/1982 1984 1978/1982 systems required for NFPA 805
/ 1990 / 1990 compliance. EA-FC-95-022 determined that the editions referenced in that code review and in the LAR are the appropriate editions to review for NFPA 805 compliance.
PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 80 1968 No 1968 Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 80A 1996 N/A 1996 Code is missing from PLDBD.
PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 90A 1985 1985 1985 No conflicts.
NFPA 92M N/A N/A No Edition Code is only mentioned in EA-FC 022. However, compliance with this code is not required by NFPA 805.
NFPA 101 2000 No 2000 Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 204 N/A N/A No Edition Code is only mentioned in EA-FC 022. However, compliance with this code is not required by NFPA 805.
LIC-1 2-0083 Enclosure Page 50 of 164 Code Edition NFPA Table B-1 / PLDBD- EC-FA Section 6.0 NU-61 022 NFPA 220 1999 N/A 1999 Code is missing from PLDBD.
PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 231 N/A No N/A Code is only mentioned in PLDBD.
Edition However, compliance with this code is not required by NFPA 805.
NFPA 232 N/A No N/A Code is only mentioned in PLDBD.
Edition However, compliance with this code is not required by NFPA 805.
NFPA 241 2000 N/A 2000 Code is missing from PLDBD.
PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 251 1999 No 1999 Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 252 1999 No N/A Code edition is missing from PLDBD.
Edition PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
Code is missing from EA-FC-95-022 because a code review is not required for this code. It is only referenced in the NFPA 805 definition of flame spread rating.
NFPA 253 N/A No N/A Code is only mentioned in PLDBD.
Edition However, compliance with this code is not required by NFPA 805.
NFPA 256 1998 N/A 1998 Code is missing from PLDBD.
PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 600 2000 N/A 2000 Code is missing from PLDBD (identified in PLDBD as NFPA 27).
PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA 701 1999 N/A 1999 Code is missing from PLDBD.
PLDBD-NU-61 is to be revised post-LAR to reference the NFPA codes used in the LAR and EA-FC-95-022.
NFPA N/A No N/A Code is only mentioned in PLDBD.
1962 Edition However, compliance with this code is not required by NFPA 805.
LIC-12-0083 Enclosure Page 51 of 164 Fire Protection Engineering RAI 05:
LAR Table B-i, Section 3.3.1.2, currently states "N/A"; however, NFPA 805 contains an overall requirement in this section. Please provide the appropriate compliance statement for Section 3.3.1.2 of Table B-I. In addition, please ensure Table B-i, for other instances, where an overall requirement is contained within a high level item description, is provided with the appropriate compliance statement.
OPPD's Response to Fire Protection Engineering RAI 05:
The correct compliance statement for Section 3.3.1.2 is "Complies". The updated LAR Transition Report Attachment A, Table B-1 with this correction to the compliance statement is provided in the table below and will be included in Attachment A of the NFPA 805 transition LAR supplement.
Additional sections of Table B-1 have been corrected to include the appropriate compliance statement to address an overall requirement contained within the high level item description.
These sections are: Section 3.2.3, Section 3.3, Section 3.3.1.1, and Section 3.4.3. The updated LAR Transition Report Attachment A (Table B-i), with these corrections to the Compliance Statements, Compliance Basis, and/or Reference Document are provided below and will be included in the NFPA 805 transition LAR supplement. [AR 48249]
NFPA 805 Compliance Compliance Basis Reference Ch 3 Ref Statement Document 3.2.3 Procedures Complies Plant procedures for implementation of the fire protection program have been developed. The procedures address, at a minimum, the fire protection program elements identified in the sections below, but are not limited to these elements. Upon review of these procedures, FCS concludes that the NFPA 805 code requirements in the following subsections are satisfied.
3.3 Prevention Complies No additional Clarification FCSG-15-11, "Fire Prevention Plan,"
Rev. 2 / All 3.3.1.1 General Complies Plant procedures for general fire prevention activities have Fire Prevention been developed and implemented. The procedures address, Activities at a minimum, the fire protection program elements identified in the sections below, but are not limited to these elements.
Upon review of these procedures, FCS concludes that the NFPA 805 code requirements in the following subsections are satisfied.
3.3.1.2 Control of Complies Combustible Materials 3.4.3 Training Complies No Additional Clarification Fire Brigade and Drills Training Program Master Plan, Revision 23 / All
LIC-1 2-0083 Enclosure Page 52 of 164 Fire Protection Engineering RAI 06:
Table B-i, Section 3.2.3(1) Inspection, Testing, and Maintenance: the compliance basis for element 3.2.3(1) of Attachment A to the Transition Report (page A-4) states "complies." During the audit, the plans to use Electric Power Research Institute (EPRI) topical report EPRI TR1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide," at a future date (post-LAR) were discussed. If using performance-based methods in lieu of the deterministic requirements of NFPA 805, Chapter 3, this must be identified and approved by the NRC via a license amendment. In addition, if performance-based methods are used, a discussion on how performance based surveillance frequencies will be established and monitored in the future needs to be provided. In this regard, please provide the following:
- a. Please state whether EPRI TR1006756 will be followed. If not, state which guidance will be followed.
OPPD's Response to Fire Protection Engineering RAI 06 a.:
In response to this RAI, OPPD currently has no plans to revise surveillance and maintenance procedures to utilize performance based methods. For NFPA 805 implementation, existing procedures and frequencies will remain unchanged except as previously noted in Attachment S-3 of the Transition Report. If changes are made in the future to incorporate performance based methods, it is planned that the changes would be made in accordance with EPRI TR1006756.
- b. Please describe how EPRI TR1006756, or other committed guidance, will be utilized to establish performance-based surveillance frequencies. Specifically discuss how the initial surveillance frequencies will be established; discuss the plant program for obtaining/using performance monitoring information to revise surveillance frequencies post-transition, and discuss the fire protection systems/features to be included in this program.
OPPD's Response to Fire Protection Engineering RAI 06 b:.
OPPD currently has no plans to revise surveillance and maintenance procedures to utilize performance based methods. For NFPA 805 implementation, existing procedures and frequencies will remain unchanged except as previously noted in Attachment S-3 of the Transition Report. If changes are made in the future to incorporate performance based methods, it is planned that those changes would be made in accordance with EPRI TR1 006756.
- c. Please describe the changes to be made to inspection, testing, and maintenance procedures during the transition period to implement performance-based " surveillance frequencies, including identifying the procedures to be changed or developed. Specifically, please discuss the elements of EPRI TR1006756, or other committed guidance, that will be implemented, including identifying the applicable fire protection systems.
LIC-1 2-0083 Enclosure Page 53 of 164 OPPD's Resp~onse to Fire Protection Engineering RAI 06 c.:
OPPD currently has no plans to revise surveillance and maintenance procedures to utilize performance based methods. For NFPA 805 implementation, existing procedures and frequencies will remain unchanged except as previously noted in Attachment S-3 of the Transition Report. If changes are made in the future to incorporate performance based methods, it is planned that the changes would be made in accordance with EPRI TR1006756.
- d. If performance-based procedure changes are/will be made, please clarify which implementation item in Table S-2 applies to these procedure changes/developments.
OPPD's Response to Fire Protection Engineering RAI 06 d.:
OPPD currently has no plans to revise surveillance and maintenance procedures to utilize performance based methods. For NFPA 805 implementation, existing procedures and frequencies will remain unchanged except as previously noted in Attachment S-3 of the Transition Report. If changes are made in the future to incorporate performance based methods, it is planned that the changes would be made in accordance with EPRI TRI 006756.
Fire Protection Engineering RAI 07:
Table B-i, Section 3.3.8 Bulk Storage: The compliance basis statement does not specifically address the prohibition on bulk storage in certain areas of the plant as stated in the first sentence of NFPA 805, Section 3.3.8. Please provide further justification for how this requirement is met.
OPPD's Response to Fire Protection Engineering RAI 07:
The compliance statement for Section 3.3.8 has been updated to include "Complies by Previous NRC Approval". There is no bulk storage of flammable or combustible liquids inside structures containing systems, equipment, or components important to nuclear safety, with the exception of that which is previously approved. The compliance basis has been revised to include a discussion of the prior NRC approval of the diesel fuel oil and lubricating oil in the diesel generator rooms and in the turbine building. (Reference NRC letter NRC-78-0104 from NRC (Reid) to OPPD (T. E. Short) dated August 23, 1978, Enclosure 2, Safety Evaluation Report, Sections 5.22 and 5.29). The compliance basis was also revised to discuss the modifications in the diesel generator rooms (MR-FC-87-42 and FC77-30) which modified the DG rooms' dry pipe sprinkler system to resolve deficiencies in sprinkler clearance and coverage and provided DG room separation with a 3-hour rated rollup door, respectively.
These corrections to the compliance statement, compliance basis, and references will be reflected in the updated NFPA 805 transition LAR supplement. [AR 48249]
LIC-12-0083 Enclosure Page 54 of 164 Fire Protection Engineering RAI 08:
Table B-i, Section 3.4.1 (c): During the audit, it was stated that the fire brigade will be reorganized to contain a fire brigade leader and at least two brigade members that will have "sufficient training and knowledge" of nuclear safety systems. Please clarify what is meant by "sufficient." Describe the level of training and knowledge these three members receive. Describe any operations training the members will receive. For example, please describe whether these members will be licensed or non-licensed operators.
OPPD's Response to Fire Protection Engineering RAI 08:
In response to this RAI, Implementation item REC-009 in the LAR will reconfigure the current fire brigade composition of a brigade leader from operations, a brigade member from operations, and three brigade members from security. This will ensure that the brigade leader and at least two brigade members have sufficient training and knowledge of nuclear safety systems to understand the effects of fire and fire suppressants on nuclear safety performance criteria. Operations personnel would complete Equipment Operator - Nuclear Auxiliary non-license classroom training. Similar training is not required for the non-operations personnel.
Fire Protection Engineering RAI 09:
Table B-i, Sections 3.4.2.1, 3.4.2.4, 3.4.3(a)(2), and 3.4.3(c)(2) indicate "complies" without any required action. Please clarify that the Radiation Release implementation items from Attachment E have been incorporated into the fire brigade pre-plans, training, and drills.
OPPD's Response to Fire Protection Engineering RAI 09:
The correct compliance statement for Sections 3.4.2.1, 3.4.2.4, 3.4.3(a)(2), and 3.4.3(c)(2) is "Complies, with Required Action." These corrections to the compliance statement, as well as the corrected compliance basis and the reference document for the LAR Transition Report Attachment A (Table B-i), are provided in the table below. These corrections will also be reflected in the updated NFPA 805 transition LAR supplement. [AR 48249]
NFPA 805 Compliance Reference Ch 3 Ref Statement Compliance Basis Document 3.4.2.1 Complies, Per section 5.2.2 of standing order SO-G-28, "Station Fire Plan," "Each pre-fire with plan consists of a minimum of two pages..."
Required Action See implementation item identified below.
IMPLEMENTATION ITEMS (see Attachment S):
REC 100 Standing Order SO-G-28, "Station Fire Plan," and associated fire brigade training materials will be revised to reflect changes required to meet the NFPA 805 radioactive release performance criteria as identified in the NFPA 805 radioactive release review, EA10-043.
3.4.2.4 Complies, Section 4 of standing order SO-G- 28, "Station Fire Plan," identifies the roles
[Pre-Fire with and responsibilities of personnel during fire emergencies and includes Plan Required coordination between each plant group.
Coordinati Action on Needs] See implementation item identified below.
LIC-12-0083 Enclosure Page 55 of 164 IMPLEMENTATION ITEMS (see Attachment S):
REC 100 Standing Order SO-G-28, "Station Fire Plan," and associated fire brigade training materials will be revised to reflect changes required to meet the NFPA 805 radioactive release performance criteria as identified in the NFPA 805 radioactive release review, EA10-043.
3.4.3(a)(2) Complies, Per section 5.3.2 of standing order SO-G-1 02, "Fire Protection SO-G-28, Training with Program Plan," "The Fire Brigade training program shall meet or exceed the "Station and Drills Required requirements of section 600 Standard on Industrial Fire Brigades NFPA Code- Fire Plan,"
Action 1992, except that the meeting frequency may be quarterly." Per section 6.1.1 Rev. 77/
of Fire Brigade Training Program Master Plan, "Fire Brigade meetings are All intended to present pertinent new information to Fire Brigade personnel on a continuing basis. These meetings shall be conducted quarterly. Examples of meeting topics include plant modifications, procedural changes, industry events, drill critiques, and review of self-evaluations. All Fire Brigade Members are required to attend each quarterly meeting to remain qualified."
See implementation item identified below.
IMPLEMENTATION ITEMS (see Attachment S):
REC 100 Standing Order SO-G-28, "Station Fire Plan," and associated fire brigade training materials will be revised to reflect changes required to meet the NFPA 805 radioactive release performance criteria as identified in the NFPA 805 radioactive release review, EA10-043.
3.4.3(c)(2) Complies, Per section 6.1.2 of the Fire Brigade Training Program Master Plan, "The area, SO-G-28, Training with type of fire and simulated use of equipment for each quarterly drill should differ "Station and Drills Required from the previous quarterly drill." Fire Plan,"
Action Rev. 77 /
The Fire Drill Evaluation Form in appendix B of Fire Brigade Training Program All Master Plan assesses the fire brigade's selection of firefighting equipment, fire attack techniques, communications, and teamwork.
See implementation item identified below.
IMPLEMENTATION ITEMS (see Attachment S):
REC 100 Standing Order SO-G-28, "Station Fire Plan," and associated fire brigade training materials will be revised to reflect changes required to meet the NFPA 805 radioactive release performance criteria as identified in the NFPA 805 radioactive release review, EA10-043.
In conclusion, the Radiation Release implementation items from the LAR Attachment E will be incorporated into the fire brigade pre-plans, training, and drills.
Fire Protection Engineering RAI 10:
Table B-1, Section 3.6.3, Hose Nozzle Clarification: NFPA 805, Section 3.6.3, states the requirements for hose nozzles supplied to each power block area. The compliance statement in the Transition Report, Table 8-1, Section 3.6.3, states "complies with EEEE"; however, the compliance basis only mentions a few nozzles and only the Auxiliary and Turbine buildings. It is unclear if the rest of the plant complies with this requirement. Please provide clarification for the following:
- a. Please state if these are the only nozzles used throughout the plant.
LIC-1 2-0083 Enclosure Page 56 of 164 OPPD's Response to Fire Protection Engineering RAI 10 a.:
These are not the only nozzles used throughout the plant. The other power block areas that contain nozzles are the Intake Structure and Radwaste Building.
- b. Please state the compliance statement for the other power block fire areas.
OPPD's Response to Fire Protection Engineering RAI 10 b.:
The compliance statements for the Intake Structure and Radwaste Building hose nozzles are as follows:
The nozzles in fire hose cabinets FP-4N and FP-4P in the Intake Structure are AKRON Turbojet Model Style 1720. There is no UL listing/label identified on the nozzles. However, the nozzles are identified as FM approved and conform to NFPA 1964 per the Akron Brass website (www.akronbrass.com). The nozzles have adjustable stream shape patterns from full fog to straight stream.
The two fire hose cabinets in the Radwaste Building are provided with small brass nozzles attached to the hoses. The nozzles are Giacomini Brass Model 647 X. The nozzle has adjustable fog, straight-stream, and shut-off features.
The nozzles are UL listed and are clearly marked as "NOT TO BE USED ON ELECTRICAL FIRES."
The LAR Transition Report Attachment A, Table B-1 has been updated with this additional information included in the Compliance Basis for Section 3.6.3. These changes to the Table B-1 will be reflected in the updated NFPA 805 transition LAR supplement. [AR48249]
- c. Please describe compliance regarding the prohibition of use of straight stream nozzle capability in high voltage settings. For example, please describe if the 250,000 volt criteria listed will satisfy all power block areas and applicable transformers.
OPPD's Response to Fire Protection Engineering RAI 10 c.:
The 250,000 volt criteria satisfy all power block areas; the yard transformers are protected by automatic deluge systems and manual fire suppression is available from hydrants and hose houses located within 500 feet of the transformers. Per the NFPA 24 code review, REC-037 will revise surveillance test OP-ST-FP-0001A, Fire Protection System Inspection and Test, to show that hose houses should be equipped with the following: two approved adjustable spray-solid stream nozzles equipped with shutoffs for each size of hose provided, as part of implementation.
- d. Please describe the meaning of "consistent with other listed nozzles" and describe why this is acceptable per NFPA 805.
LIC-1 2-0083 Enclosure Page 57 of 164 OPPD's Response to Fire Protection Engineering RAI 10 d.:
The nozzle is an adjustable fog-type "consistent with other listed nozzles" and is an integral part of the Norris Model 411 hose rack assembly that does bear a UL label.
Although the nozzle itself does not bear a UL mark, the construction and use of the nozzle is similar to that of UL listed adjustable fog-type nozzles. Manufacturers do not necessarily mark all parts of the standard assembly with the UL mark but instead classify the entire assembly as UL listed. These assemblies typically utilize UL listed components, including hose nozzles, and therefore, if the assembly remains as-is from the manufacturer, the hose rack assembly, including hose and nozzle, would be considered UL listed.
Fire Protection Engineering RAI 11:
Table B-i, Sections 3.10.1 and 3.10.3, reference engineering analysis document EA-FC-93-047, "Halon System Operability Evaluation." However, Section 3.10.1 refers to revision 3 while Section 3.10.3 refers to revision 4, and LAR Section 6.139 references revision 3. Please provide justification for allowing two different revisions to be referenced within LAR or clarify the correct revision.
OPPD's Response to Fire Protection Engineering RAI 11:
This was a typographical error in the LAR. The correct revision of engineering analysis document EA-FC-93-047, "Halon System Operability Evaluation," is Revision 4. The updated LAR Transition Report Attachment A, Table B-1 with the corrected revision number for the Basis Document in Section 3.10.1(2) and the updated LAR Section 6.0 with correct revision number for this reference are provided below and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
The LAR Section 6.0 (References) Item 6.139 has been revised to indicate the corrected EA revision number as follows:
6.139. EA-FC-93-047, Halon System Operability Evaluation, Revision 4 The LAR Attachment A, Table B-i, Section 3.10.1(2) has been revised to reflect Revision 4 of the EA as follows:
Reference Document EA-FC-93-047, "Halon System Operability Evaluation," Rev. 4 / All Fire Protection Engineering RAI 12:
Table B-i, Section 3.4.4: The compliance basis points to an implementation item; however, there is no implementation item listed in Section 3.4.4. Please provide a new implementation item or provide clarification of which implementation item is part of this compliance basis.
LIC-1 2-0083 Enclosure Page 58 of 164 OPPD's Response to Fire Protection Engineering RAI 12:
The correct compliance statements for Section 3.4.4 are "Complies" and "Complies with Use of EEEEs." The compliance basis pointing to an implementation item ("See implementation item identified below") has been removed and the correct compliance basis and additional reference document provided. The updated LAR Transition Report Attachment A (Table B-i),
with this correction to the compliance statement, the corrected compliance basis, and updated reference document will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
Fire Protection Engineering RAI 13:
Table B-i, Section 3.3.12: Provide the missing reference to OPPD Letter [LIC-88-1066]
which requested NRC-88-0457. Please clarify whether LIC-97-0073 dated September 30, 1997, should be referenced in Table B-i, Section 3.3.12.
OPPD's Response to Fire Protection Engineering RAI 13:
The "Reference Documents" section for the LAR Transition Report Attachment A (Table B-i)
Sections 3.3.12, 3.3.12(1), 3.3.12(2), 3.3.12(3), 3.3.12(4), and 3.3.12(5) has been updated to provide reference to the following OPPD letters, which requested the exemption for the RCP lube oil collection system:
Letter LIC-97-0073 from OPPD (S. K. Gambhir) to NRC (Document Control Desk) dated September 30, 1997 / All Letter LIC-88-1066 from OPPD (K. J. Morris) to NRC (Document Control Desk) dated November 28, 1988 / All In addition, the compliance basis and reference document sections have been updated for the LAR, Transition Report Attachment A (Table B-i), Sections 3.3.12, 3.3.12(1), 3.3.12(2),
3.3.12(3), 3.3.12(4), and 3.3.12(5), to identify the document number for the NRC letter from NRC (R. Wharton) to OPPD (S. K. Gambhir) dated May 21, 1998, as letter NRC-98-0083. The updated Transition Report Attachment A (Table B-i), with these corrections to the reference documents and the corrected compliance basis will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
Fire Protection Engineering RAI 14:
Table B-i, Section 3.3.9, requires implementation item REC-039 which requires inspection of transformer oil collection basins to ensure the basins are free of debris. If applicable:
- a. Please provide the frequency used and the justification for the basis of this frequency of inspections.
LIC-12-0083 Enclosure Page 59 of 164 OPPD's Response to Fire Protection Engineering RAI 14 a.:
Where rock-filled oil collection basins are used, inspections will be performed in accordance with the guidance in IEEE Std 980-1994, IEEE Guide for Containment and Control of Oil Spills in Substations. This will include periodic inspection to ensure there is no filling of void spaces by dirt, dust or silt. The initial frequency will be set to coincide with the testing of the transformer suppression systems which are tested each refueling outage. This frequency may be adjusted based on trending of results.
- b. Please indicate how this frequency is related to the fire protection system inspection frequencies.
OPPD's Response to Fire Protection Engqineering RAI 14 b.:
The inspection frequency will be set to coincide with the testing of the transformer suppression systems. These systems are tested each refueling outage.
- c. Please provide any plans on how frequency changes to the inspection frequency will be monitored (if frequency changes are planned/ expected).
OPPD's Response to Fire Protection Engineering RAI 14 c.:
OPPD has no plans to alter this inspection frequency at this time.
Fire Protection Engineering RAI 15:
Table B-i, Sections 3.3.7 and 3.3.8, discuss storage requirements for flammable gas and flammable and combustible liquids including requirements to meet NFPA 30, "Flammable and Combustible Liquids Code," and NFPA 50A, "Standard for Gaseous Hydrogen Systems at Consumer Sites," respectively. Please clarify if there are non-hydrogen flammable gas or combustible liquids stored on-site (e.g., acetylene, propane, etc.) and whether the compliance basis statements include these non-hydrogen materials. Also, please clarify if these materials are controlled under the combustible control program and meet the requirements of NFPA 30 and/or NFPA 50A.
OPPD's Response to Fire Protection Engineering RAI 15:
The LAR Transition Report, Attachment A (Table B-i) Section 3.3.7.1 has been revised to include an updated compliance basis to identify the non-hydrogen flammable gas stored and used at Fort Calhoun. Station guideline FCSG-15-36, "Compressed Gas Cylinder Safety" provides guidelines to specify requirements for the safe handling, use, and storage of compressed gas cylinders at FCS. Procedure SO-G-91, "Control and Transportation of Combustible Materials," establishes requirements for the storage and handling of combustible materials.
LIC-1 2-0083 Enclosure Page 60 of 164 The LAR Transition Report, Attachment A, Table B-1 Section 3.3.8 has been revised to include an updated compliance basis which identifies that there is no bulk storage of flammable or combustible liquids inside structures containing systems, equipment, or components important to nuclear safety, with the exception of those previously approved. The compliance statement has been updated to include "Complies by Previous NRC Approval" and the compliance basis has also been revised to include a discussion of the prior approval of the diesel fuel oil and lubricating oil in the Diesel Generator Rooms and the turbine lube oil in the Turbine Building.
The updated LAR Transition Report, Attachment A, Table B-1 with these corrections to Sections 3.3.7.1 and 3.3.8 will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
General storage and use of combustible liquids complies with NFPA 30 as identified in the NFPA 30 checklists in EA-FC-95-022. Non-hydrogen flammable gas does not need to meet the requirements of NFPA 50A as NFPA 50A is only applicable to bulk hydrogen storage.
Fire Protection Engineering RAI 16:
Table B-i, Item 3.3.5.1, indicates a reliance on implementation item REC-050 to modify SO-G-21; however, SO-G-21 is not listed in the referenced document's column. Please clarify why SO-G-21 is not referenced and clarify whether SO-G-21 and the FHA will include any restrictions to ensure cabling is kept to a minimum regardless of the type/classification of loading.
OPPD's Response to Fire Protection Engineering RAI 16:
The LAR Transition Report, Attachment A (Table B-i), Section 3.3.5.1 has been revised to include reference to the following procedure under the column entitled "Reference Documents.":
- SO-G-21, "Modification Control," Rev. 86 / All The updated LAR Transition Report Attachment A (Table B-i), with the correction to the reference document for Section 3.3.5.1, will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
The LAR Transition Report Attachment A (Table B-i) Section 3.3.5.1 includes Implementation Item REC-050 which states the following:
Standing Order SO-G-21, "Modification Control," will be revised to require the use of plenum-rated wiring above suspended ceilings. This revision is being tracked by EC 50741 and will be completed as part of LAR implementation.
This revision to SO-G-21 will include a requirement to ensure that combustible loading in the area above the suspended ceiling is kept to a minimum.
LIC-12-0083 Enclosure Page 61 of 164 Fire Protection Engineering RAI 17:
Table 4-3 and the appropriate section of Table B-3 for Fire Area 36B: Page L-15, Attachment L Approval #7, refers to an overhead horizontal cable chase (i.e. fire area 36C) in room 56W (west switchgear room) that was previously approved as a radiant energy shield. Note: this Pyrocrete shield is not the same feature which seeks approval under Attachment L, Approval #7, but is in reference as similar in construction to the room 56E (east switchgear room) new approval submitted feature. Table 4-3 and the appropriate section of Table B-3 do not reflect this feature for Fire Area 36B (room 56W, west switchgear room). Please provide justification for not including this credited feature in Table 4-3 and the appropriate section of Table B-3.
In addition, Table 4-3 states "Pyrocrete enclosures protecting vertical cable trays and horizontal conduit bank" are required electrical raceway fire barrier systems (ERFBS) features for fire area 36A (east switchgear room). However, Table 4-3, Table B-i, and Table B-3 do not clearly identify which ERFBS are credited for fire area 36A.
Specifically, please identify, by location and configuration type, all ERFBS features to be credited within Fire Area 36A. In addition, please clarify if the two vertical Pyrocrete ERFBS features inside fire area 36A between column lines 6d and 7a are credited for the NFPA 805 Fire Protection Program (FPP) and are accounted for within the Table 4-3 (fire area 36A) statement "Pyrocrete enclosures protecting vertical cable trays ... >
OPPD's Response to Fire Protection Engineering RAI 17:
LAR Transition Report Table 4-3 and LAR Transition Report Attachment C, Table B-3 for Fire Area 36B have been updated to include an entry for "Features" to identify the following ERFBS feature previously identified as fire area 36C:
0 Pyrocrete enclosure protecting cable tray 54S (fire area 36C)
LAR Transition Report Table 4-3 and LAR Transition Report Attachment C, Table B-3 for Fire Area 36A have been updated to include three separate entries for "Features" to specifically identify the following ERFBS features credited in the fire area:
" Pyrocrete enclosure protecting overhead horizontal Train B cabling encased in conduit between column lines 3a and 4a, and terminating at panel AI-109B
- Pyrocrete enclosure protecting vertical Train B cable tray sections 22S to 5-4A from fire area 32 (below) to fire area 41 (above) between column lines 6d and 7a
" Pyrocrete enclosure protecting vertical Train A cable tray sections 10S to 5-4B from fire area 32 (below) to fire area 41 (above) between column lines 6d and 7a The updated LAR Transition Report Table 4-3 and Attachment C, Table B-3 with the correction to the ERFBS features in Fire Areas 36A and 36B will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
LIC-1 2-0083 Enclosure Page 62 of 164 LAR Transition Report Attachment A, Table B-1 Section 3.11.2 has been revised to remove the Prior Approval compliance bases and associated reference documents of the Pyrocrete enclosure that "forms the credited barrier between fire areas 31 and 31A in the intake structure" and the compliance basis of the Pyrocrete enclosure that 'forms the barrier between fire areas 36B and 36C in the west switchgear room". The Prior Approval compliance bases and associated reference documents have been added to LAR Transition Report Attachment A, Table B-1 Section 3.11.5 as these Pyrocrete enclosures are considered ERFBS features as they are not separating credited fire areas.
The updated LAR Transition Report, Attachment A, Table B-1 with the correction to Sections 3.11.2 and 3.11.5 will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
Fire Protection Engineering RAI 18:
Attachment L, Approval #7: Please provide further details for why the ERFBS does not meet Generic Letter (GL) 86-10, Supplement 1, "Fire Endurance Test Acceptance Criteria for Fire Barrier Systems Used to Separate Redundant Safe Shutdown Trains within the Same Fire Area." During the audit, it was discussed that this ERFBS does not meet any specific fire tests under Supplement 1, and the use of NFPA 805, Section 3.11.5, was also discussed. Please clarify if the intention is to comply with NFPA 805, Exception 2 to Section 3.11.5, or to seek a new approval under Attachment L.
- a. If seeking Exception 2 from Section 3.11.5, then please provide the basis for the conclusion for successfully meeting the pre-GL-86-10 letter and the limiting end point temperature requirements as specified by the NRC at the time of acceptance. Include a summary of the testing, the limiting end point temperature used, and any reference to the engineering analysis and testing.
Also, please include any similar configurations that have received similar pre-GL-86-1 0 approval.
- b. If seeking a new approval, then please provide the basis for the conclusion of equivalency to GL 86-10, Supplement 1. Include a summary of any testing performed, a figure depicting the as-installed and as-tested/analyzed configuration, a description of the materials utilized, and a reference to the engineering analysis document(s).
OPPD's Response to Fire Protection Engineering RAI 18 a. and b.:
The LAR Transition Report, Attachment L has been revised to delete Approval #7 as compliance with NFPA 805 Section 3.11.5 is being met through Exception No. 2 of Section 3.11.5. The LAR Transition Report, Attachment A, Table B-1 Section 3.11.5 has been revised to remove the compliance statement, "Submit for NRC Approval," and the associated compliance basis for the approval request.
LIC-12-0083 Enclosure Page 63 of 164 The compliance basis under "Complies with Clarification" has been revised to identify that the following Pyrocrete enclosure configurations were employed prior to the issuance of Generic Letter 86-10, Supplement 1. They were tested against the end point temperature requirements similar to the acceptance criteria of NFPA 251 that are identified in Generic Letter 86-10, Supplement 1. The configurations are therefore acceptable as 3-hour rated enclosures in accordance with Exception No. 2 of Section 3.11.5 of NFPA 805 per the time-temperature data supplied by the Pyrocrete manufacturer, which is included as Attachment B to letter LIC 0062.
Overhead horizontal Train B cabling encased in conduit, wrapped in metal lath, and surrounded by 2 inches of Pyrocrete in fire area 36A between column lines 3a and 4a, from fire area 36B and terminating at panel Al-109B in fire area 36A.
Vertical Train B cable tray sections 22S to 5-4A from fire area 32 (below) to fire area 41 (above) within fire area 36A between column lines 6d and 7a is wrapped in metal lath, and surrounded by 2 inches of Pyrocrete.
Vertical Train A cable tray sections 10S to 5-4B from fire area 32 (below) to fire area 41 (above) within fire area 36A between column lines 6d and 7a is wrapped in metal lath, and surrounded by 2 inches of Pyrocrete.
The updated LAR Transition Report Attachment A, Table B-1 with the correction to Section 3.11.5 and the updated LAR Transition Report Attachment L, with the Approval #7 removed will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
Fire Protection Engineering RAI 19:
Table B-i, Item 3.5.14(b), and Attachment L, Approval #6: Please discuss receiving approval for certain curb valves within the fire water main loop which do not meet the locked or supervised requirements stated in NFPA 805, Section 3.5.14.
OPPD's Response to Fire Protection Engineering RAI 19:
Table B-i, Section 3.5.14 has been revised as follows to provide detail of the specific gate valves with roadway boxes (curb valves) to which the request for approval is applicable:
NRC approval of the three (3) valves northeast of the machine shop and the two (2) valves east of the service building that are underground gate valves with roadway boxes (curb valves) that are sectionalizing the underground yard fire main loop that are not supervised, locked, or sealed, is being requested in Attachment L of the Transition Report, Item 5.
The LAR Transition Report, Attachment L, Approval #5 has also been revised to include this detail. The updated LAR Transition Report, Attachment A, Table B-1 with this correction and the updated Attachment L with this corrected information will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
LIC-1 2-0083 Enclosure Page 64 of 164 Fire Protection Engineering RAI 20:
Attachment K: Please clarify the need to rely on the exemption regarding no suppression within the MCR in order to meet the NFPA 805 FPP requirements.
OPPD's Response to Fire Protection Engineering RAI 20:
The Attachment K transition and Attachment T clarification for this prior approval is not required on the basis that the Main Control Room (fire area 42) is transitioned to NFPA 805 as a Performance Based Fire Area, and is not subject to the NFPA 805 deterministic requirement for full area suppression.
The risk informed performance based assessment for fire area 42 has established the fire risk as being within the acceptance criteria of RG 1.174, and that adequate defense-in-depth and safety margin are maintained without full area suppression.
The revisions to LAR Transition Report Section 4.2.3, Table B-3, and LAR Attachments K, 0, and T pertaining to the deletion of the prior approval lack of full area suppression in fire area 42 are provided below and will also be reflected in the NFPA 805 transition LAR supplement.
[AR 48249]
Excerpts from the LAR Transition Report and Attachments (markups):
LAR Section 4.2.3 Licensing Action Transition Results DELETED BULLETED ITEMS [transition of this prior approval is not required on the basis that the Main Control Room is transitioned to NFPA 805 as a Performance Based Fire Area, and is not subject to the NFPA 805 deterministic requirement for full area suppression]:
" Fire Area 42: control room, lack of area wide supesin In alternt shutdown arca (july 31985)
" L~ack of full suppression in the conRtrol room LAR Attachment C Table B-3 Fire Area Transition Fire Area 42 Licensing Actions DELETED IN ENTIRETY [clarification of this prior approval is not required on the basis that the Main Control Room is transitioned to NFPA 805 as a Performance Based Fire Area, and is not subject to the NFPA 805 deterministic requirement for full area suppression]:
LIC-1 2-0083 Enclosure Page 65 of 164 LAR Attachment K Existing Licensing Action Transition Licensing Action - Appendix R Exemption, Control Room, Lack of area wide suppression in alternate shutdown area (III.G.3 criteria)
DELETED IN ENTIRETY (transition of this prior approval is not required on the basis that the Main Control Room is transitioned to NFPA 805 as a Performance Based Fire Area, and is not subject to the NFPA 805 deterministic requirement for full area suppression):
Licensing Action Appondix R Exomption, Control Room, Lack of aroa wide sprsini altFrnatc shutdown area (llGF3cIteria)
Basis Date; July 3, 1985
- - ,*.,,, -- -- ^I --.-
Transetioned?: Yes Basis! EXGmn~tiGn roauested nor OPPD submittal LIC 83-0219 to theNR reques~ts approVal for the lack of an area-wide automatic firo SUPpr.so SytemF in an area for which alternate shu toFwei*, provded, cOna tth requirements Of Section 1IILG.3 of Appendix R. Approval was granted inthe RCExemptio NRC 85-0200, which pro'.idos the followin;g bases for acceptability:
1..The fire hazard in; this area is loW-.
- 2. A petential fire would tend to develop slowly because of the wide dispersion of the com.ustible materials that may ignite.
- 3. Because of the smoke detection systemsG and the contiuous manngi the cOntro room, a fire would be detected inisiiilstages and extinguished before serious damage ocurred.
- 4. If serious dlamage shoul1d occu-r befo-re the arrival of the plant fire brigade, an alternate shutdown capability exists that is independent of the room In concIlusion, the bases for prevou accptance r-emain valid and have be Clarification isrqurd regarding the roUting of cables, from the cable spreadin-g room to the cabinlýts i the ma*in Eo)nRtl room no*t routed in cnhduit, for the
- , is of this exemption.
in addition, clarificatio ireured regarding the presence Of safety relate equipment in the comute roo, although the impact of a fire on the safet related equpeti not ad-versoe to the establishment of rsafe and stable plan Although the exact number and conRfiguration Of com~bustibles have nrad over tim~e (per SER NIRC 85-02200, combus61tible loadring 1,000 E3T U/f2 Q) 1Im
LIC-1 2-0083 Enclosure Page 66 of 164 per E=A-FC 97-001, "Firo Hazards Analysis Manual," Rov. 15, comnbustibl loadiRg "Leow" per Calculation F*C681I4, "UF^HA Combustible Loading," Rev. 1-,
combu6t*ble loading 59,41,44 BTU/ft 2 @ 45 mm, tho combu.stible loadig* ; remains low and oxomption basos remain valid.
Additional information i6 provided in Attachmont T, ClarificationR Of Prior R Applic~ablc Firc Arcac:
FirA irrr o,-r Ar 42- Contrtol Room Complex Area LU~nsigAe ton Doeume~nta Loffor LIC 8-3-0-219 from Jonos6 (OPPID) to Clark (NRC) datod August 30, 1-983&
ATTACHM=ENTI A, SECTION V111 "B. Evxomptior Reque6t Tho District requiests an exemption p rsat to Sections 50.12(a) and 50.48Wc of"OCER, from the requiremen.ts S+e*tion "t .G.3 of A.pp..dix R. Specifically, exemption is requested from the reurmet that a fixed fire suppression sys,9tem be installed in the con)trol room.
Thi"s requirement is unnecossar,' to assur+.e the c . to safely shutdoew the
+apability plant in the event of any credible fire in this fire area for the folloEwing reasons-.
Fire Protection The control room c~omplex area is protec~ted by fire detecators loc~ated as follows:
see ionzation type in the control room (F=A 42A), two ionization type inth computer room (42B), n inztion type and one duet type in the elevatorshf (42G), five ioiaintpDnteFpeRsonel corridor (42FE), one thermnal type in the kitchen (41)D), two duet-installed ionization types in the control room ventilto dueting, and onRe ionization type detector in the shift suporv'sor'Es offic~e (42D).
following: -a20 lb. G02, 10( lb. G02, and a 10 lb. drFy chemical in tho conrol1 roo (F=A 42A); a 22 lb. Halen and a 10 lb. C02 in the comFputer room (429); a 10lb dr', c~hemic~al in the corridor (42E); and a 10 l b. dry chem~ical in the kitchen (42D).
In addition, there are two fire hose8 cabinets locEated in the corridorF to serge ars a fire suppression backup9.
LIC-1 2-0083 Enclosure Page 67 of 164 T-horo are two .d.,oor-"ays to the Control Room Complex area from the turbine building. Eac-.Jr-h of theosoi-.rEtocted by 3-hour fire Fated doors and an automatic wator curtaiR nemuntod over the turbine building side of the dooevay.
Additionally, thoro is a fixed Halon Fire SuppressionR Systom installed inthemi walk panos. ;n1control Postulated Fire Fire Area 42A Co+ntrol Room:. A fire is postulated in a console containi instrumentation adontrl cEirc;ut-s'*of safety related cable diVisions EA, EF, EC, Fire Area 42B CoGmputer Room: Afire ispostulated inthe com~puter circulitry-.
Fire Area 42C Stai ell. NO combustibles.
Fire ,A~ea 42D -Shift Superiser's Area and Kitchen: A paper fire is postulatedi the Shift SUPerMSEr's, offic.
Fire Area 42FE Personnel Corridor and Storage Closets: A rag fire is postulated in the Janitorial Closet.
Consequen~es, of F~ire Without Active Protection Fire Area 42A ConGtrol Room: The cabinets in the contro~l room form a metal enclosure with c~ables entering the bottom of the cabinets through conduit4fo the cable room below. FoFr some plant func~tions the redundant control and display information is totally separated by virtue of being in adjacent cabines Even in the cases where redundant divisionRs enter the panel they are separate by metal bharriters within the panel's At the back of each panel where the termina boards are located. ThGidviul oductorFs are FRu in wiring harnesses to the front of the panel. ThereforFe, it is only at the front of the panel where wirin harnssesmay be in clos6e proximity that the separation criteria is not satisfied.
The initiationR of a fire in those cabinets is extremely unlIkl since the voltage.
and amperage is low. The cable insu~lation is fire retard-ant anRd oxidatio proceeds slowly. The control roo icotinuously manned and an overheae electrical component, would be readily detected both by the personnel and by the ion~ization detectors. The portable fire extinguishers are available insideth conRtrol room to allow rapid extinguishmnent Of the fire.
Additionally, the fixed Halon Fire Suppression System in the main conro cabinets will adequately con~trol and extinguish any fire in the panels. There are no oil resr.ir or ther significant stockpilesm Of com~bustibles to aggravata small c~able insulation fire. The control room ventilatio is capble of mana contro~l so that the once through ventilation moede can be used to rapidly clear h
LIC-1 2-0083 Enclosure Page 68 of 164 room of smoke. Thoroforo safo shutdown is not precludod by a fire inthe control Fire Area 42B Cmputor Room.:-The plant com.puter do. not po...r" any safety related fuRntions. The postulated fire in the comiputerFroom will be an annoyance, but will not affect the ability of the plant to aIhieve safe shutdown.
The maximum fire soverity is estimated as 12 minuters whic~h is based on stacks of cmputerF printout papeF. This m-aterial w*uld be vn,' difficult to Ignite in its compact form. Therefore a fire of any severity On this room is hypothetical.
Fimre Area 42GC Staim~ell: No fire postulated due to lack Of com~bustibles.
Fir*e Area 42D Shift SUPerVOr,,i'S Office and Kitchen. The postulated fire in this area is due to the paper in files and Rmiscellaneous flammable furniture coverings.
This space does not sorwo an" safety related fnc~tions. No safety-rel-ated c-ables pass through this area. It *an be is;ol-ated from the control room by mFoan Of the 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> fire rated doors and the 1 1,'2 hour2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> fire rated walls. The postulated fire safe smhutdown.
Fire Area 42E Personnel Corridor and Strage C postulated fireisI TheIsets:
assumed to be rags in the janitorial closet. There are no significant comnbustibles inR the corrid-or itsGelf. No safetyrlated cables pass3 through the corridor area. Th ventilationR duc~ting for the control room passes through the cr~eridoer, however,th maxium eveityfire for this area could nOt affec this ducting. Therefore,th pos~tulated fire does not affect the ability of the plant to achio've safe shutdown-.
Alternate rshutdown capability which has been provided also lessensm the effect of a ver' unlikely control room fire. Based on the above, the District reqluests an exemnption from the rqientsC I. of. those portions of Sections ll!.G.3 of Appendix R which require that additional fire protection) features be prov~idedfo fire area 42 of the Fort Calhoun Station."
NRC SER dated July 3, 1985 (NRC 85 0200):.
SECTION 7.0, "CONTROL= ROOM (FIRE AREA 42)"
SECTION 8.3, "EVALUATION" "The tec;Ahnical requieeemlents of Section lll.G are net met inthis area becaus o the absence of an area w;ideC~I~ fixed fire supprsso system.
The fire hazard in this area is low. Because of the wide dfis~persion Ofth combustible mnaterials that may ignite, a potential fire would tend to dev~elep slowly. Beaueof the smoke detection systems6 and the con~tinuu anigi the centFOI room, a fire would be detected in tR nta stages and extinguise before swerius damage occurred.
LIC-1 2-0083 Enclosure Page 69 of 164 If serious damage should occur before the arrival of the plant fire brigade, an alteFRatc shutdown capability cXis~ts that is indcpondont of the room. Thrcforc-,
safo shutdoWn could bo achieved and mnaintained."
SECTION 8.4, "CONCLUSION" "Bacod on our ovaiuation, We conclude that the Xisting fire proitection provides an eto acceptable 1 level of safety equivalent to that achieved by comFpliance wt ntlr~rtelcnesrqettreepin~raTXCTr suppression system in the con.trol room , should be approved."
Assoemaa rId EnQlnnpanl Er*l..enll Ea Iu nameni Engineering analysis EA FC -90 088, "PFie Resistance Rlating of Doors 1036 1, 1036- 2, and 1011-28," Rle'.. 0, justifies the acceptability of three Control Room Comnplex doors that are not installed in a tes~ted (single) coenfiguration, but rahe in apair* Jdcofiguration. This e*eMption*- cdits "walls, floor and ceiling of 3_hour fire-r"*fat;ed construction with openings protected by either fire rated doors, dam:pers Or penetration seals-i." Teacpability of this exemption is jusGtifiedo the basis of the content of EA-FC 0 88 LAR Attachment 0 Orders and Exemptions Exemptions DELETED BULLETED ITEMS [transition of this prior approval is not required on the basis that the Main Control Room is transitioned to NFPA 805 as a Performance, Based Fire Area, and is not subject to the NFPA 805 deterministic requirement for full area suppression]:
0 Fire Area 42: control room, lack of area-wide supesini alternt shutdown area (July 3, 198&)
0 Fire Area 42: control room, lack of area-wide supeso)n alternt shutdown area (July 3, 1985)
LAR Attachment T Clarification of Prior NRC Approvals Prior Approval Clarification Request 12 DELETED IN ENTIRETY [clarification of this prior approval is not required on the basis that the Main Control Room is transitioned to NFPA 805 as a Performance Based Fire Area, and is not subject to the NFPA 805 deterministic requirement for full area suppression]:
LIC-1 2-0083 Enclosure Page 70 of 164 Monitoring Program RAI 01:
NFPA 805, Section 2.6, "Monitoring," states that, "a monitoring program shall be established to ensure that the availability and reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria" and that "Monitoring shall ensure that the assumptions in the engineering analysis remain valid."
Specifically, NFPA 805, Section 2.6 states that:
(2.6.1) "Acceptable levels of availability, reliability, and performance shall be established."
(2.6.2) "Methods to monitor availability, reliability, and performance shall be established. The methods shall consider the plant operating experience and industry operating experience."
(2.6.3) "If the established levels of availability, reliability, or performance are not met, appropriate corrective actions to return to the established levels shall be implemented. Monitoring shall be continued to ensure that the corrective actions are effective."
In addition, Section 4.6, "Monitoring Program" of the Transition Report states that the NFPA 805 monitoring program will be implemented "after the safety evaluation issuance as part of the fire protection program transition to NFPA 805" (Table S-3, Implementation Items, item 11-805-089 of the Transition Report).
Furthermore, the licensee has committed to comply with Frequently Asked Question (FAQ) 10-0059. The NRC staff noted that the information provided in Section 4.6, "Monitoring Program," of the Transition Report is insufficient for the staff to complete its review of the monitoring program. Please provide the following additional information for the NRC staff to complete its review:
- a. A description of the process by which systems, structures, and components (SSCs) will be identified for inclusion in the NFPA 805 monitoring program, including the approach to be applied to any 'fire protection SSCs that are already included within the scope of the Maintenance Rule program.
OPPD's Response to Monitorinq Progqram RAI 01 a.:
The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 describes the process by which systems, structures, and components (SSCs) will be identified for inclusion in the NFPA 805 monitoring program, including the approach to be applied to any fire protection SSCs that are already included within the scope of the Maintenance Rule program. The revised LAR Section 4.6.2 is provided in Attachment 1 (MP RAI 01-1) to this Enclosure and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
LIC-12-0083 Enclosure Page 71 of 164
- b. A description of the process that will be used to assign availability, reliability, and performance goals to SSCs within the scope of the monitoring program including the approach to be applied to any SSCs for which availability, reliability, and performance goals are not readily quantified.
OPPD's Response to Monitoring Program RAI 01 b.:
The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 provides a description of the process that will be used to assign availability, reliability, and performance goals to High Safety Significant (HSS) SSCs within the scope of the monitoring program. Low Safety Significant (LSS) SSCs do not specifically require assignment of availability, reliability, and performance goals.
Programmatic elements such as fire brigade performance, fire watches, combustible controls, etc., will be evaluated using the existing program health process. It is not practical to assign target values of reliability and availability to these attributes so their effectiveness is based on objective and anecdotal evidence evaluated by plant personnel in charge of the fire protection programs as is currently practiced. The revised Section 4.6.2 is provided in Attachment 1 to this Enclosure and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
- c. A demonstration of how the monitoring program will address a response to programmatic or training elements that fail to meet performance goals (examples include fire brigade response or performance standards and discrepancies in programmatic areas such as combustible controls programs).
OPPD's Response to Monitoring Program RAI 01 c.:
The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 provides a description of how the monitoring program will address response to programmatic elements that fail to meet performance goals. Training is implicitly included within the performance regarding programmatic elements. The revised Section 4.6.2 is provided in Attachment 1 to this Enclosure and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
- d. A description of how the monitoring program will address fundamental fire protection program elements.
OPPID's Response to Monitoring Program RAI 01 d.:
The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 provides a description of how the monitoring program addresses fire protection systems and features and programmatic elements. The revised Section 4.6.2 is provided in Attachment 1 to this Enclosure and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
LIC-1 2-0083 Enclosure Page 72 of 164
- e. A description of how the guidance in EPRI TR1006756, "Fire Protection Equipment Surveillance Optimization and Maintenance Guide," if used, will be integrated into the monitoring program.
OPPD's Response to Monitoring Program RAI 01 e.:
As described in the response to FCS Fire Protection Engineering RAI 06, and as identified in LAR Transition Report Attachment A, Table B-i, Section 3.2.3(1), the frequency at which inspections, testing and maintenance of the fire protection systems and features are performed will be evaluated using the EPRI TR1006756.
EPRI TR1006756, Section 11 contains the guidance which ensures that reliability levels established are consistent with the FPRA and Maintenance Rule Program.
The EPRI TR1006756 may be used as input for establishing reliability targets, action levels, and monitoring frequency. When establishing the action level threshold for reliability and availability, the action level will be no lower than the fire PRA assumptions.
- f. A description of how periodic assessments of the monitoring program will be performed taking into account, where practical, industry wide operating experience including whether this process will include both internal and external assessments and the frequency at which these assessments will be performed.
OPPD's Response to Monitoring Program RAI 01 f.:
The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. The revised Section 4.6.2 provides a description of how periodic assessments of the monitoring program will be performed, including consideration of internal and external operating experience (OE), nominally every three years. The periodic assessment frequency will be adjusted based on inputs from industry OE and internal audits. The revised Section 4.6.2 is provided in Attachment 1 to this Enclosure and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
- g. Section 4.6.2 of LAR describes an overview of the post-transition NFPA monitoring program to be implemented after safety evaluation issuance.
Phase 2, in Section 4.6.2 is entitled "Establishing Risk Criteria." FAQ 10-0059 identifies this phase as "Screening Using Risk Criteria." Please describe the changes that are being proposed in regards to the NFPA 805 monitoring program compared to the monitoring program described in FAQ 10-0059.
OPPD's Response to Monitoring Program RAI 01 g.:
The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. Phase 2 will be implemented consistent with the approved FAQ 10-0059. The revised LAR Section 4.6.2 is provided in Attachment 1 to this Enclosure and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
LIC-1 2-0083 Enclosure Page 73 of 164
- h. The discussion of phases 2, 3, and 4 of the proposed NFPA 805 monitoring program discussed in Section 4.6 of the LAR appears to omit Nuclear Safety Capability Assessment (NSCA) equipment. Please describe how the NFPA 805 monitoring program will address NSCA equipment.
OPPD's Response to Monitoring Program RAI 01 h.:
The LAR Transition Report Section 4.6.2 has been revised to align with the approved FAQ 10-0059 and its related closure memo. Nuclear Safety Capability Assessment (NSCA) equipment will be assessed and included in the monitoring program consistent with the approved FAQ 10-0059. The revised LAR Section 4.6.2 is provided in Attachment 1 to this Enclosure and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
- i. Please describe the requirements for periodic NFPA 805 assessments (audits) of the fire protection program under the existing Fire Protection Quality Assurance Program. If these assessments are not conducted under the existing Fire Protection Quality Assurance Program, please describe the program and process that will be used to conduct them.
OPPD's Response to Monitoring Program RAI 01 i.:
Quality Assurance requirements (audits) for the existing Fire Protection Program are documented in the following: QAP-6.7 - Fire Protection, QAP-10.1 - Audit Program and Audits, and USAR Appendix A - Quality Assurance Program. FCS intends to maintain the same quality assurance requirements (audits) for NFPA 805 as are defined in the documents above. The revised LAR Section 4.7.3 is provided below and will be reflected in the NFPA 805 transition LAR supplement.
[AR 48249]
LAR Section 4.7.3 Compliance with Quality Requirements in Section 2.7.3 of NFPA 805 Fire Protection Program Quality REVISED TEXT FROM:
FCS will maintain the existing Fire Protection Quality Assurance program.
During the transition to 10CFR50.48(c), FCS performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805.
REVISED TEXT TO:
The existing FCS Fire Protection QA program requirements are contained in the following documents:
LIC-12-0083 Enclosure Page 74 of 164
" USAR Appendix A - Quality Assurance Program
- QAP-6.7 - Fire Protection
- QAP-10.1 -Audit Program and Audits QA Program Utilized During Transition During the transition to 10 CFR 50.48(c), FCS performed work in accordance with the quality requirements of Section 2.7.3 of NFPA 805 and the existing FP QA Program described above. This included requirements that each analysis, calculation, or evaluation performed to support compliance with 10 CFR 50.48(c) be independently reviewed.
Post Transition QA Program FCS will utilize the existing Fire Protection Quality Assurance program with the following changes.
In addition to editorial and administrative changes (i.e. replacing references to previous NRC guidelines with those associated with the NFPA 805 transition and ensuring the features required for a performance based program under NFPA 805 are addressed), the components and systems currently considered within the scope of the Fire Protection QA Program will be expanded to include those components and systems that are in the power block and are required by Chapter 4 of NFPA 805. This means that certain FP systems and features in some buildings not currently considered under the FP QA Program that are required by NFPA 805 Chapter 4 will now fall under the Fire Protection QA program. As such, any future modifications to these systems will be conducted under the design controls required by the FP QA program.
The FP QA Program includes a requirement to conduct independent audits of the FP Program by the Quality Department to ensure that the requirements of the fire protection program are being effectively implemented. The audit requirements contained in QAP-6.7 and/or QAP-10.1 will be revised as applicable to include the periodic review of the Monitoring Program.
LIC-1 2-0083 Enclosure Page 75 of 164 Programmatic RAI 01:
NFPA 805 Section 2.7.1.1 states that "the analyses performed to demonstrate compliance with this standard shall be documented for each nuclear power plant (NPP).
The intent of the documentation is that the assumptions be clearly defined and that the results be easily understood, that results be clearly and consistently described, and that sufficient detail be provided to allow future review of the entire analyses.
Documentation shall be maintained for the life of the plant and be organized carefully so that it can be checked for adequacy and accuracy either by an independent reviewer or by the AHJ."
NFPA 805 Section 2.4.3.3 states that "the PSA approach, methods, and data shall be acceptable to the AHJ. They shall be appropriate for the nature and scope of the change being evaluated, be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant."
NFPA 805 Section 3.3.1.2 states that "procedures for the control of general housekeeping practices and the control of transient combustibles shall be developed and implemented."
FPRA analyses assume combustible loading will be maintained at or below certain values. Please provide a description of how the combustible controls program will be administered to ensure that FPRA assumptions regarding combustible loading will be met.
OPPD's Response to Programmatic RAI 01:
Transient combustibles within the station protected area are controlled by Standing Order SO-G-91, Control and Transportation of Combustible Materials. Changes to the standing order would require review by a Society of Fire Protection Engineers member. This individual would ensure that the change was consistent with the FPRA assumptions regarding combustible loading.
LIC-12-0083 Enclosure Page 76 of 164 Safe Shutdown Analysis RAI 01:
Table B-2, NEI 00-01, Rev. 1, Alignment Basis: Regulatory Guide 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants,"
Rev. 1, identifies NEI 00-01, "Guidance for Post-Fire Safe Shutdown Circuit Analysis,"
Rev. 2, Chapter 3, as the guidance document to be used to ensure alignment with current NRC guidance for the application of NFPA 805. LAR Table B-2 identifies a comparison with NEI 00-01, Rev. 1. Please provide a gap analysis of differences between the alignments using NEI 00-01, Rev. 2, as the basis for transitioning compared to NEI 00-01, Rev. 1.
OPPD's Response to Safe Shutdown Analysis RAI 01:
FCS EA10-064, "Nuclear Safety Performance Analysis Methodology Review," provides a comparison or gap analysis between NEI 00-01, Revision 1, and NEI 00-01, Revision 2.
Based on the gap analysis, there are no significant differences between alignment with NEI 00-01, Revision 1 and NEI 00-01, Revision 2 for FCS. LAR Transition Report Section 4.2.1.1 is revised to provide a discussion of these results. The proposed revision to the LAR Transition Report Section 4.2.1.1 is provided below and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
NFPA 805 Transition LAR Section 4.2.1.1 Compliance with NFPA 805 Section 2.4.2 Overview of Process:
ADDED TEXT beginning with "Description of Gap..." after the last paragraph.
Description of Gap Analysis between NEI 00-01 Revisions 1 and 2:
In May 2009, Revision 2 to NEI 00-01 was issued. Revision 2 in large measure implements the Required for Hot Shutdown (RHSD; "Green box") and Important to Safe Shutdown (ISSD; "Orange box") criteria for classifying SSD-credited devices and determining the allowable tools to address their failure. In general, Revision 2 incorporates the guidance associated with multiple spurious operations (MSOs) as related to those plants not transitioning to NFPA 805. Specifically, the methodology in Revision 2 reflects insights gained from, not only the EPRI/NEI Cable Fire Testing, but also the CAROLFIRE Cable Fire Testing, the outcome of meetings with the NRC Staff and information provided within SECY 08-0093 and a draft revision to Regulatory Guide 1.189. These changes were made to address NRC comments related to segregating those components necessary for post-fire hot shutdown ("green box", defined in 10 CFR 50, Appendix R, Section IIl.G.1.a as one train of systems necessary to achieve and maintain hot shutdown conditions) and those whose mal-operation could provide a potential impact to post-fire safe shutdown ("orange box", defined 10 CFR 50, Appendix R, Section III.G.1 as components important to safe shutdown that could adversely affect safe shutdown capability or cause mal-operation of safe shutdown systems). However, when RG 1.205, Revision 1 was issued in support of NEI 04-02, Revision 2, it endorsed the methodology contained in NEI 00-01, Revision 2 as one acceptable approach to circuit analysis for a plant implementing a Fire Protection Program under 10 CFR 50.48(c).
LIC-1 2-0083 Enclosure Page 77 of 164 A paragraph-by-paragraph review was conducted of the criteria and assumptions presented in NEI 00-01, Revision 1 and Revision 2 to address any potential gaps to demonstrate FCS meets the guidelines of Revision 2, where applicable. The paragraph-by-paragraph review is documented in FCS Engineering Analysis EA10-064, "Nuclear Safety Performance Analysis Methodology Review."
The review concluded that the circuit failure criteria set forth in NEI 00-01, Revision 2 does not differ significantly from that in Revision 1, and specifically from that implemented through the use of project procedures. Differences between NEI 00-01, Revisions 1 and 2 not applicable to circuit analysis include:
" The FCS NFPA 805 transition project evaluated MSOs consistent with the process outlined in' FAQ 07-038, Revision 3, "Lessons Learned on Multiple Spurious Operations" [ML110140242].
" Manual operation of valves must consider the effect of a fire on stem lubrication, where applicable. An evaluation should be conducted to justify those instances where hand wheel operation of a valve in the area of the fire is credited. FCS does not credit Recovery Actions for valves located in the fire affected area.
Additional information for evaluating circuit coordination for power sources has been implemented at FCS by use of loss of overcurrent trip methodology outlined in FCS Engineering Analysis EA10-036, "Fort Calhoun Station Automation and Update of Safe Shutdown Analysis.'
Safe Shutdown Analysis RAI 02:
Safe and Stable: LAR Section 4.2.1.2 describes the capability for achieving and maintaining safe and stable conditions. The licensing basis is to achieve and maintain hot shutdown (Mode 3) conditions with the minimum plant operating shift staff "to demonstrate that FCS can achieve and maintain Mode 3 (Hot Shutdown Condition) ... for a coping time of 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />." LAR Table B-2, Section 3.1.1.9, Criteria/Assumptions states that the 24-hour coping time has been selected based on the design capacity for the backup nitrogen supply that is relied upon to maintain positive remote control over the turbine driven auxiliary feedwater pump, and based on the ability of the Emergency Response Organization (ERO) to respond to the event with adequate time allowed for the ERO personnel to muster, assess the extent of fire damage, and assist the plant operating staff with implementation of the required actions to sustain Mode 3 (Hot Shutdown Condition), beyond 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, or to assess the extent of fire damage, and assist the plant operating staff with implementation of cold shutdown actions and/or cold shutdown repairs for the plant to transition to, and enter, Mode 4 (Cold Shutdown Condition).
Actions required to sustain Mode 3 (Hot Shutdown Condition) are included in the results part of Section 4.2.1.2 of the LAR including reference to appropriate procedures.
LIC-1 2-0083 Enclosure Page 78 of 164 Please provide a description of the risk impact(s) of any of these actions including the nature of the actions (i.e.; routine, abnormal, or repair) that may be required to maintain safe and stable.
OPPD's Response to Safe Shutdown Analysis RAI 02:
The fire PRA, consistent with the internal events PRA, models human failure to isolate LCV-1190 (hotwell level control valve) upon loss of instrument air and subsequent depletion of nitrogen backup accumulators in four hours. The consequence of this human failure event (XlS01 190) is failure of LCV-1 190 in the open position, drain down of the CST to the condenser, and loss of suction to FW-54. The CDF, LERF, ACDF, and ALERF results reported in the LAR include failure to perform this action.
Failure to isolate LCV-1 190 is important from a fire risk perspective. Approximately one third of total fire CDF is attributed to switchgear room fire scenarios, whose dominant sequences involve station blackout. FW-54 provides one of two AC-independent methods of secondary side heat removal. This action is also important for fire scenarios that degrade, or fail, secondary side heat removal from FW-6 and/or FW-10, although such scenarios generally pose less of a contribution to the overall fire risk profile than sequences involving station blackout. Isolation of LCV-1 190 is generally a reliable action, with its internal events human error probability in the mid E-03 range, due to the long time window available, low execution complexity, and general operator sensitivity to maintaining the decay heat removal key safety function.
The fire PRA, consistent with the internal events PRA, models human failure to refill the EFWST prior to its depletion (eight hours following a complete loss of main feedwater). The consequence of this human failure event (AHFFEFWST) is failure of long term secondary side heat removal when use of FW-54 (including its suction from the CST) is also failed. The CDF, LERF, ACDF, and ALERF results reported in the LAR include failure to perform this action.
This is an abnormal action directed by an abnormal operating procedure.
Failure to refill the EFWST is important from a fire risk perspective. It is particularly important for fire scenarios in which FW-54 is failed or degraded, and it therefore has some interaction with human failure to isolate LCV-1 190. The dependency between failure to isolate LCV-1190 and failure to refill the EFWST is quantitatively modeled by the fire PRA. Failure of EFWST refill frequently occurs in the dominant cutsets for fire sequences not involving station blackout or loss of RCS integrity. EFWST refill is generally a reliable action, with its internal events human error probability in the mid E-04 range, due to the long time window available, low execution complexity, multiple diverse methods available for the refill, and general operator sensitivity to maintaining the decay heat removal key safety function.
The fire PRA, consistent with the internal events PRA, models spurious RAS upon loss of instrument air and subsequent depletion of backup accumulators in 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br />. The consequence of a spurious RAS is loss of suction to both HPSI and LPSI. The fire PRA does not credit AOP-17 to prevent/block a spurious RAS upon nitrogen depletion. The CDF, LERF, ACDF, and ALERF results reported in the LAR are therefore bounding, conservatively not crediting the benefit of such an action.
LIC-1 2-0083 Enclosure Page 79 of 164 While the fire PRA conservatively does not credit operator action to prevent the spurious RAS upon loss of instrument air, this action would only be relevant to fire sequences involving LOCAs or once-through-core-cooling. While such sequences do contribute an important part of the overall fire risk profile, they tend to be less risk significant than the more dominant fire sequences involving station blackout.
With respect to the following action described in LAR Section 4.2.1.2:
"Loss of instrument air, per AOP-17, (specifically, approximately 12-hour actions to prevent a recirculation actuation signal [RAS] from safety injection and refueling water storage tank low level signal [STLS])."
The main control room operator is provided with the capability to prevent the RAS.
Consequently, no field action is required for NFPA 805 safe and stable plant operation to prevent STLS (RAS). The statement immediately above will be removed from LAR Section 4.2.1.2. This change to the LAR Transition Report Section 4.2.1.2 will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
Safe Shutdown Analysis RAI 03:
Nuclear Safety Capability Analysis (NSCA), Safe Shutdown and Circuit Analysis, and Table B-3 Cable Routing for Fire PRA: The following are the cable routing assumptions (EA-10-0037) (R2008-007-003) for cables where no routing information appears to be available:
- a. Cables that transition from one numbered tray to another through an "unnumbered" vertical riser the cable path was assumed to include the riser.
- b. Short runs of cable from a numbered tray to an end device, where the fire area location for both the tray and the end device are known, and where both are located in the same fire area or two adjacent fire areas, but the connecting conduit/air drop could not be identified from the plant conduit and raceway layout drawings, the cable is assumed to be located in the same fire area or the two adjacent fire areas.
- c. Short runs of cable from one end device to another end device, where the fire area location for both are known and both are located in the same fire area or two adjacent areas the cable is assumed to be located in the fire area(s).
- d. Cables of short length (20 feet or less) where the fire area location for only one of the end devices are known, the fire area location for the other end device is assumed to be located in the same fire area (typically valve junction boxes connecting to valve operators, solenoids, and or limit switches).
- e. Cables from one end device in the Containment to another end device in the Containment are assumed to be contained entirely within the Containment.
LIC-1 2-0083 Enclosure Page 80 of 164
- f. Cables from one end device in the Intake Structure to another end device in the Intake Structure are assumed to be contained entirely within the Intake Structure.
- g. Cables connecting panels between the MCR (for cables that could not be traced) are assumed to route from the MCR to the Cable Spread Room (CSR) and then back to the MCR. They are assumed to be exclusively in the MCR and CSR.
Please indicate the number of cables/circuits that do not have routing information where these assumptions were applied.
In target selection, ZOI were established above each ignition source. Please indicate how targets without cable routing information were selected to be included in the ZOI for each ignition source.
Please indicate if this routing was factored into the field walkdowns for ZOl. In addition, indicate what specifically was done to account for this uncertainty in the analysis.
The analysis (EA-10-0037) (R2008-007-003) identifies that raceway and conduit layout drawings, architectural drawings, and fire area boundary drawings were used to locate equipment, raceways, and cables within the plant fire areas and fire zones.
Please describe any field verification done to determine the suitability of the above assumptions.
OPPD's Response to Safe Shutdown Analysis RAI 03:
- Please indicate the number of cables/circuits that do not have routing information where these assumptions were applied.
Assumption a: For cables that transition from one numbered tray to another through an "un-numbered" vertical riser the cable path was assumed to include the riser.
Assumption "a" is not applicable to any cables as the assumption has been verified. A review of FCS conduit and raceway layout drawings has identified the "un-numbered" vertical risers listed below. Cable fire area locations have been verified to appropriately include the fire areas associated with the vertical risers. The presence of a cable passing through vertical riser is established from the numbered horizontal tray sections and/or cables (conduits) that lead to and from the vertical riser as identified in the FACTS database, and on the FCS conduit and raceway layout drawings.
LIC-1 2-0083 Enclosure Page 81 of 164 VERTICAL RISER SHOWN ON:
1 11405-E-62, 11405-E-64, 11405-E-78 SH. 1, 11405-E-78 SH. 2, coord. 7a.5-J.6, in FA 06-3, FA 20-1, and FA 41 (two trays) 2 11405-E-62, 11405-E-64, 11405-E-78 SH. 1, 11405-E-78 SH. 2, coord. 7a-J.6, in FA 06-3, FA 20-1, and FA 41 (two trays) 3 11405-E-64, 11405-E-66, coord. 8a-P.1, in FA 20-1 and FA 20-7 (two trays) 4 11 405-E-64, 11405-E-66, coord. 7a-P.1, in FA 20-1 and FA 20-7 (two trays) 5 11405-E-72 SH. 1, 11405-E-73 SH. 1, 11405-E-78 SH. 1, 11405-E-78 SH. 2, D-4397, coord. 6d.7-C, in FA 32, FA 36A, and FA 41 (one tray) 6 11405-E-72 SH. 1, 11405-E-73 SH. 1, 11405-E-78 SH. 1, 11405-E-78 SH. 2, D-4397, coord. 6d.4-C, in FA 32, FA 36A, and FA 41 (one tray) 7 11 405-E-72 SH. 1, 11 405-E-73 SH. 1, coord. 4a.8-D, in FA 33, FA 32, and FA 36B (one tray) 8 11405-E-72 SH. 1, 11405-E-73 SH. 1, coord. 4a.4-D, in FA 33, FA 32, and FA 36B
_ (one tray) 9 11405-E-72 SH. 1, 11405-E-73 SH. 1, coord. 5.b-D.9, in FA 34A and FA 34B-1 (one tray) 10 11405-E-73 SH. 1, coord. 4a.6-D.9, in FA 34B-1 (one tray - up to but not into FA 43) 11 11 405-E-73 SH. 1, coord. 2b-D.7, in FA 34B-1 (one tray - up to but not into FA 43) 12 11405-E-73 SH. 1, 11405-E-78 SH. 1, 11405-E-78 SH. 2, D-4397, coord. 6d-C.2, in FA 36A and FA 41 (two trays) 13 11405-E-73 SH. 1, 11405-E-78 SH. 1, 11405-E-78 SH. 2, D-4397, coord. 6d-C.5, in FA 36A and FA 41 (two trays) 14 11405-E-73 SH. 1, 11405-E-78 SH. 1, 11405-E-78 SH. 2, coord. 8a.5-C.8, in FA 32 and FA 41 (one tray) 15 11405-E-73 SH. 1, 11405-E-78 SH. 1, 11405-E-78 SH. 2, coord. 8a.5-C.9, in FA 32 and FA 41 (one tray) 16 11405-E-73 SH. 1, 11405-E-78 SH. 1, 11405-E-78 SH. 2, coord. 7a.3-E, in FA 32 and FA 41 (four trays) 17 11 405-E-78 SH. 1, coord. 7a.4-C.3, in FA 41 (one tray - up to but not into FA 42) 18 11405-E-78 SH. 1, coord. 7a.4-C.7, in FA 41 (one tray - up to but not into FA 42) 19 11 405-E-78 SH. 1, 11 405-E-78 SH. 2, E-4113, coord. 6d-J1, in FA 28 and FA 43
_ (one tray) 20 11 405-E-78 SH. 1, 11 405-E-78 SH. 2, coord. 5b-D, in FA 36C (two trays - up to but not into FA 43) 21 11 405-E-78 SH. 2, coord. 7a.9-D, in FA 41 (one tray - up to but not into FA 42) 22 11405-E-78 SH. 2, coord. 8a.2-D.8, in FA 41 (one tray - up to but not into FA 42) 23 D-4397, coord. 6d.5-D.2, in FA 41 (three trays - up to but not into FA 42) 24 D-4397, coord. 6d.5-D.9, in FA 41 (two trays - up to but not into FA 42)
NOTE 11405-E-72 SH. 1, 11405-E-73 SH. 1, 11405-E-74, 11405-E-76, all vertical risers 1 adjacent to electrical penetrations outside of containment are uniquely identified (numbered) in FACTS routing with tray section numbers.
NOTE 11405-E-92 SH. 1, 11405-E-93 SH. 1, 11405-E-94, 11405-E-98, all vertical risers 2 adjacent to electrical penetrations inside of containment are uniquely identified (numbered) in FACTS routing with tray section numbers.
NOTE 11405-E-92 SH. 1, 11405-E-92 SH. 2, 11405-E-93 SH. 1, 11405-E-93 SH. 2, all 3 vertical risers not adjacent to electrical penetrations inside of containment are not uniquely identified in FACTS routing with tray section numbers; however, the containment is one fire area, and all risers connect to horizontal tray sections which are uniquely identified (numbered) in FACTS routing with tray section numbers.
LIC-1 2-0083 Enclosure Page 82 of 164 Assumption b: For short runs of cable from a numbered tray to an end device, where the fire area location for both the tray and the end device are known, and where both are located in the same fire area or two adjacent fire areas, but the connecting conduit/air drop could not be identified from the plant conduit and raceway layout drawings, the cable is assumed to be located in the same fire area or the two adjacent fire areas.
Assumption "b" is applicable to 26 cables (for longer cables, in context "short runs" applies to the portion of the cable route near the to/from destination).
Assumption c: For short runs of cable from one end device to another end device, where the fire area location for both are known and both are located in the same fire area or two adjacent areas the cable is assumed to be located in the fire area(s).
Assumption "c" is applicable to 356 cables (for longer cables, in context "short runs" applies to the portion of the cable route near the to/from destination).
- 36 cables with indicated length of 1 foot
- 262 cables with indicated length of 2 feet
- 4 cables with indicated length of greater than 2 feet, but less than 10 feet
- 45 cables with indicated length of greater than or equal to 10 feet, but less than or equal to 40 feet
- 9 cables with indicated length of greater than 40 feet, but less than 300 feet This assumption generally applies to valves and local junction boxes which are both shown on the conduit and raceway layout drawings. Although the valve is typically shown, the appurtenances typically mounted on the valve or immediately local to the valve (i.e., associated solenoid(s), limit switches, I/P devices, E/P devices, etc.) are not.
Assumption d: For cable of short length (20 feet or less) where the fire area location for only one of the end devices are known, the fire area location for the other end device is assumed to be located in the same fire area.
Assumption "d" is applicable to 38 cables.
- 9 cables with indicated length of 1 foot
- 24 cables with indicated length of 2 feet
- 2 cables with indicated length of greater than 2 feet, but less than 10 feet
- 3 cables with indicated length of greater than 10 feet, but less than 15 feet This assumption generally applies to local junction boxes which are shown on the conduit and raceway layout drawings, but where the connecting end devices (i.e.,
process controllers, process switches, etc.) are not.
Assumption e: Cables from one end device in the containment to another end device in the Containment are assumed to be contained entirely within the containment.
LIC-1 2-0083 Enclosure Page 83 of 164 Assumption "e" is applicable to 72 cables. This is not an assumption per se as different cable IDs are used for the same electrically continuous conductor contained within a cable outside containment and within a cable inside containment.
Assumption f: Cables from one end device in the intake structure to another end device in the intake structure are assumed to be contained entirely within the intake structure.
Assumption 'T" is applicable to 3 cables. This is not an assumption per se as the analysis cables which exit the intake structure all do so through the same manhole and duct bank, and are explicitly identified on the FCS conduit and raceway layout drawings.
Assumption g: Cables connecting panels between the main control room (MCR) (for cables that could not be traced) are assumed to route from the MCR to the cable spreading room (CSR) and then back to the MCR. They are assumed to be exclusively in the MCR and CSR.
Assumption "g" is applicable to 13 cables.
- Intarget selection, ZOI were established above each ignition source. Please indicate how targets without cable routing information were selected to be included in the ZOI for each ignition source.
Please refer to the response to PRA RAI 01 g. for discussion of cable assumptions and the fire PRA.
- Please indicate whether or not this routing was factored into the field walk downs for ZOl. In addition, also indicate what specifically was done to account for this uncertainty in the analysis.
Please refer to the response to PRA RAI 01 g. for discussion of cable assumptions and the fire PRA.
" The analysis (EA-10-0037) (R2008-007-003) identifies that raceway and conduit layout drawings, architectural drawings, and fire area boundary drawings were used to locate equipment, raceways, and cables within the plant fire areas and fire zones.
Please describe any field verification done to determine the suitability of the above assumptions.
The initial project activity to assign fire area locations to analysis cables was performed by FCS primarily utilizing the FCS conduit and raceway layout drawings and FACTS data. Plant walkdowns were performed by FCS engineering personnel for a small number of cables (no more than 20) to verify conduit locations where this information could not be determined from the conduit and raceway layout drawings and FACTS data.
Please refer to the response to PRA RAI 01 g. for discussion of cable assumptions and the fire PRA.
LIC-1 2-0083 Enclosure Page 84 of 164 Safe Shutdown Analysis RAI 04:
Table 4.3, Summary of Compliance Bases: LAR Table 4.3 and LAR Attachment C, lists "NFPA 805 Regulatory Basis" for each fire area listed. For the Containment (FA 30), the basis is listed as 4.2.3.2 which does not align with the separation description used in the Containment of Table B-3 and supporting documents. Section 4.2.3.2 of NFPA 805 states that: "One success path of required cables and equipment shall be located in a separate area having boundaries consisting of fire barriers with a minimum fire resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />." This does not appear to apply to the Containment.
- a. Please clarify this discrepancy in Table 4.3 and Attachment C regulatory compliance.
OPPD's Response to Safe Shutdown Analysis RAI 04 a.:
In response to this RAI, OPPD has reviewed the LAR Table 4.3 (column "NFPA 805 Regulatory Basis") and Table B-3 (field "Regulatory Basis") for each fire area. This review has resulted in a revision to LAR Table 4.3 and Table B-3 correcting the NFPA 805 Regulatory Basis to eliminate the discrepancy as noted by the NRC in Safe Shutdown Analysis RAI 04. For fire area 30 (Containment), the correct NFPA 805 regulatory basis is 4.2.3.4.a and 4.2.3.4.b. For fire area 47, the correct NFPA 805 regulatory basis is 4.2.3.3.a. The revisions to the LAR Transition Report Table 4.3 and Table B-3 are provided below and will be reflected in the NFPA 805 transition LAR supplement.
LAR Table 4.3 -
SUMMARY
OF NFPA 805 COMPLIANCE BASIS AND REQUIRED FIRE PROTECTION SYSTEMS AND FEATURES Fire Area 30 / NFPA 805 Regulatory Basis REVISED TEXT FROM: 4.2.3.2 REVISED TEXT TO: 4.2.3.4.a and 4.2.3.4.b Fire Area 47 / NFPA 805 Regulatory Basis REVISED TEXT FROM: 4.2.3.2 REVISED TEXT TO: 4.2.3.3.a LAR Attachment C, Table B-3, Fire Area Transition Fire Area 30 Regulatory Basis REVISED TEXT FROM [necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC]:
NFPA 805 Section 4.2.3.2 Deterministic Approach
LIC-1 2-0083 Enclosure Page 85 of 164 REVISED TEXT TO [necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC]:
NFPA 805 Sections 4.2.3.4.a and 4.2.3.4.b Deterministic Approach Fire Area 47 Regulatory Basis REVISED TEXT FROM [necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC]:
NFPA 805 Section 4.2.3.2 Deterministic Approach REVISED TEXT TO [necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC]:
NFPA 805 Section 4.2.3.3.a Deterministic Approach Furthermore, as a consequence of the OPPD and NRC breakout meetings held during the NFPA 805 LAR audit at the site in March 2012, and the NRC triennial fire protection audit in April 2012, OPPD has incorporated several clarifications and corrections associated with information in LAR Transition Report Section 4.2.3, Table B-3, and LAR Attachments G, K, and T pertaining to the NFPA 805 regulatory basis for:
- Fire Area 30, Containment
- Fire Area 31, Intake Structure
- Fire Area 34B-1, Upper Electrical Penetration Room
- Fire Area 36C, Pyrocrete Enclosure in Fire Area 36B The revision to the LAR Transition Report Section 4.2.3, Table B-3, and LAR Attachments G, K, and T pertaining to the clarifications and corrections for fire areas 30, 31, 34B-1, and 36C are provided in Attachment 2 (SSD RAI 04-2) to this Enclosure, and will be reflected in the NFPA 805 transition LAR supplement. [AR 48249] The basis for each change is also provided in Attachment 2 to this Enclosure.
- b. Please indicate whether or not there are any other fire areas that should be modified regarding this compliance column.
OPPD's Response to Safe Shutdown Analysis RAI 04 b.:
No other fire areas, other than fire areas 30 and 47, require modification regarding the compliance column.
LIC-12-0083 Enclosure Page 86 of 164 Safe Shutdown Analysis RAI 05:
Table B-3: LAR Table B-3, Fire Area 31 (Intake Structure), variance from deterministic requirement (VFDR)31-001, describes local manipulation of HCV-2805A and/or RW-144 and local pressure indication from PI-2805A-1/2 are required to manually rotate and backwash strainer AC-12A. AC-12A is required to function in order to support train 'A' raw water system operation, which in turn supports refill of the emergency feedwater storage tank (EFWST), to maintain reactor coolant system (RCS) core decay heat removal for long-term safe and stable post-fire safe shutdown. It also describes in VFDR 31-002 the same for the Train B strainer.
Please provide the following additional information:
- a. Indicate how often the strainers need to be backwashed.
- b. Indicate whether or not the need for backwashing the strainers changes seasonally or with different weather related events.
- c. Indicate how often the strainers need to be manually backwashed.
- d. Indicate whether or not the operators ever backwashed the strainers manually.
- e. Indicate whether or not the strainer has a bypass.
- f. Describe the compliance strategy for these VFDRs and if these actions will be included in the fire emergency procedure(s) even though the risk of their failure has been determined to be acceptably low.
- g. Indicate how these actions are reviewed for feasibility.
OPPD's Responses to Safe Shutdown Analysis RAI 05 a - q:
- a. The raw water (RW) strainers are automatically backwashed on a timer (-12 minutes) and also on differential pressure.
- b. The frequency of automatic backwash is typically greater during periods of high river flow in the spring and autumn.
- c. During normal operation, manual strainer operation is only performed briefly to free up a jammed strainer element. This is an infrequently performed operation that occurs during high river flows in the spring and fall. The need for and frequency of operation of the strainers in the event of a fire is very low. The likelihood of needing to backwash is very low given the performance of the new intake screens. The screens have eliminated the carryover of debris into the intake cells where the RW pumps take suction. This greatly reduces the debris going into the strainers and the need for backwash.
- d. See response to Item c. above.
- e. The RW strainers do not have a bypass.
LIC-12-0083 Enclosure Page 87 of 164
- f. The intake structure has transitioned to NFPA 805 as a performance based fire area.
The fire risk quantification performed for this fire area (compartment) has determined the delta CDF and delta LERF to be within the region III acceptance criteria of Regulatory Guide 1.174. Additionally, the fire risk evaluation performed for this fire area (EA 1-010) has determined that adequate defense-in-depth and safety margin is maintained without having to rely upon an NFPA 805 recovery action to manually backwash the RW strainer. Consequently, no actions will be included in the fire emergency procedure(s) to manually backwash the RW strainers for a fire event occurring at the intake structure.
- g. Recovery action feasibility is only assessed for credited NFPA 805 recovery actions.
Safe Shutdown Analysis RAI 06:
LAR Table B-3, Fire Area 34B-1 (Electrical Penetration Area Ground and Intermediate Levels), utilizes the fire risk evaluation performance-based approach in accordance with NFPA 805, Section 4.2.4.2, with 17 VFDRs and 5 recovery actions (RAs) which indicates both trains might be affected by fire in the compartment. However, the suppression effects description indicates, "Fire area 34B is not considered to contain redundant safe shutdown equipment or cables." Please clarify this discrepancy.
OPPD's Response to Safe Shutdown Analysis RAI 06:
This discrepancy arises from the methodology utilized by OPPD in the preparation of FCS engineering analysis EA-FC-99-023, "Fire Protection Suppression Effects Analysis." The EA was originally prepared in support of the 10 CFR 50 Appendix R licensing basis and was utilized as a direct input to the discussions for "Fire Suppression Effects on Nuclear Safety Performance Criteria" in LAR Table B-3. The EA identifies fire areas as being "not considered to contain redundant safe shutdown equipment or cables" where safe shutdown capability can be demonstrated with credit taken for operator manual actions (as evaluated by the separate 10 CFR 50 Appendix R safe shutdown analysis).
In response to this RAI, OPPD has reviewed the discussion of fire suppression effects on nuclear safety performance criteria for each fire area identified in LAR Table B-3. This review has resulted in revision to the LAR Table B-3 discussions for fire suppression effects to eliminate the discrepancy as noted by the NRC in Safe Shutdown Analysis RAI 06, and to improve consistency and enhance the level of technical and basis information regarding fire suppression effects. The revised LAR Table B-3 discussions continue to utilize a subset of information from EA-FC-99-023.
The revision to LAR Transition Report, Attachment C, Table B-3, which has been provided on the SharePoint Portal for the NRC technical reviewers to review, will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
LIC-1 2-0083 Enclosure Page 88 of 164 Safe Shutdown Analysis RAI 07:
LAR Table B-3, Attachment I, Power Block, Structure excluded from the power block:
Attachment I of the Transition Report states that "The service building (NFPA 805 fire area 45; the combination of legacy fire areas 45, 48, and 49 as identified for the service building in EA-FC-97-001) is excluded from the power block on the basis that it contains only the fuel oil transfer pump (and its associated power cable) for diesel driven AFW
[auxiliary feedwater] pump, FW-54."
Based on the use of FW-54 in the NFPA 805 analysis, it would appear that FW-54 and all required supporting SSCs are "required for plant operation" since these SSCs are credited as a means of meeting the nuclear safety performance criteria and should therefore be included within the structures considered within the "Power Block."
Please justify why this fire area (FA45) would not be considered part of the power block.
OPPD's Response to Safe Shutdown Analysis RAI 07:
The service building contains only one NSCA component (the fuel oil transfer pump for the non-safety related diesel driven AFW pump) and its associated power cable. This component supports operation of the non-safety related diesel driven AFW pump. This component and its associated cable are also included in the Fire PRA. The service building does not contain any other components and/or cables which are included in the NSCA, NPO, or the Fire PRA. The service building does not contain radiological sources within the scope of the radiological release requirements of NFPA 805.
The service building is located adjacent to the turbine building - fire area 46 and adjacent to the general outdoor yard areas of the plant - fire area YD. The non-safety related diesel driven AFW pump is not required for a deterministic fire event occurring in any of these fire areas alone or in combination. Both of the redundant safety related AFW pumps, one motor driven and one turbine driven, are available for deterministic compliance in these three fire areas. On this basis, FCS is excluding the service building from the power block so as to prevent having to meet NFPA 805 chapter 3 requirements for this low fire risk structure.
Safe Shutdown Analysis RAI 08:
Table S-3 Implementation item (REC-140) to perform a supporting analysis for diesel driven AFW pump, FW-54 states: "The analysis will document that FW-54 can supply adequate feedwater flow to the steam generators through the auxiliary feedwater flow path in support of the NFPA 805 safe and stable definition (i.e., achieve and maintain hot shutdown for a 24-hour coping time, with steam rejection through the spring relief mode of the main steam safety relief valves). The supporting analysis shall assume that the flow path from FW-54 to the main feedwater header remains un-isolated, and shall account for any flow diversion from FW-54 to the main feedwater header and/or the main feedwater pumps. The supporting analysis will be incorporated into a revised engineering analysis, or will be documented in a new engineering analysis.
Implementation item associated with the NFPA 805 area-by-area review, EA10-044."
LIC-1 2-0083 Enclosure Page 89 of 164
- a. Please describe the current licensing basis for FW-54 and indicate whether or not it differs from that being proposed under NFPA 805.
- b. Please describe the outcome or preliminary outcome of these results. In the event this is not feasible, describe any contingency established to accommodate feedwater for safe and stable operation and indicate how the FPRA analysis will be revised.
- c. If this analysis will support only a 24-hour safe and stable period, please describe the alternative means of achieving the nuclear safety performance goals and objectives.
OPPD's Response to Safe Shutdown Analysis RAI 08 a., b., and c.:
- a. Although there was no specific licensing basis identified for the diesel-driven AFW pump FW-54, this pump is described in a letter from OPPD (W. G. Gates) to NRC (Document Control Desk), "System Information for Auxiliary Feedwater Pump FW-54," dated October 20, 1992 (LIC-92-0312). From LIC-92-0312, the General Design Features for FW-54 were identified as:
"OPPD added a third auxiliary feedwater pump, Tag #FW-54, which is independent of the existing FCS power sources and of the water supply for the existl ng safety related auxiliary feedwater pumps FW-6 (electric motor driven) and FW-10 (steam turbine driven). FW-6 and FW-10 are supplied by the safety grade Emergency Feedwater Storage Tank in Room 81 of the Auxiliary Building. FW-54 is a centrifugal eight-stage unit driven by a diesel engine. It is manually started either locally or from the control room. It can be used for normal startup or shutdown operations, or it can be used as a backup to the safety grade pumps during a plant event."
The functional use intended for FW-54 under NFPA 805 remains consistent with the general design features identified in LIC-92-0312.
FW-54 is further discussed in the letter from OPPD (R. T. Ridenoure) to NRC (Documentation Control Desk), "Exemption Request from the Requirements of 10 CFR 50, Appendix R, Section III.G.2 for Fire Area 32 at the Fort Calhoun Station," dated November 8, 2002 (LIC-02-0118).
- b. The design intent for the diesel driven AFW pump (FW-54) is to provide the primary source of feedwater to the steam generators during plant startup, hot standby and cool down operations. The pump takes suction from the condensate storage tank and is rated to provide 300 gpm at 2600 feet of head.
By contrast, the turbine driven AFW pump (FW-1 0) and the motor driven AFW pump (FW-6) are each rated to provide 260 gpm at 2400 feet of head. FCS has performed a preliminary engineering assessment in support of this RAI response for the diesel driven AFW pump flow capability and has determined that the pump can meet the NFPA 805 safe and stable flow requirements inclusive of the potential flow diversion path to feedwater pump discharge
LIC-1 2-0083 Enclosure Page 90 of 164 piping. This preliminary engineering assessment will be turned into a formal calculation under the existing LAR implementation Item (REC-1 40).
- c. Existing LAR implementation REC-137 has committed FCS to perform a plant modification that will ensure the design requirement for a 24-hour makeup water supply to FW-54 in accordance with LAR Section 4.2.1.2 which describes the capability for achieving and maintaining safe and stable conditions. The design change as conceptually described in REC-137 is to provide an alternate source of suction for FW-54 from the main condenser. However, the design change as conceptually described in REC-137 may evolve during the conduct of the plant design change process as other potential and better alternatives are vetted. Deviations from the design change as conceptually described in REC-137 will be communicated to the NRC staff through the FCS licensing process.
The plant design change associated with REC-137 will also ensure provisions for the alignment of one or more makeup water supplies to FW-54 following the 24-hour safe and stable coping time for NFPA 805. These provisions will be implemented through plant procedures and/or facility changes to align makeup water for FW-54 from diverse sources potentially including: demineralized water, Blair City Water, raw water, fire water, and/or river water.
Safe Shutdown Analysis RAI 09:
The MCR abandonment analysis CN-RAM-10-014 (FC07824), Section 4.3.4, identifies revisions required to AOP-06 "Fire Emergency" for habitability criteria. It also states that these will be added to changes required for "significant loss of plant control"
- a. Please describe the change(s) that will be made to the procedure for significant loss of plant control.
OPPD's Response to Safe Shutdown Analysis RAI 09 a.:
AOP-06 will be revised to provide operators more flexibility in deciding if and when to abandon the control room. Fire scenarios where the environment remains habitable but plant control degrades are the subject of this RAI. For such scenarios, OPPD is considering a variety of options with the following objectives:
- Avoid premature abandonment due to prescriptive criteria that may not apply to the spectrum of fire severities and impacts that could occur.
- Provide guidance for determining when sufficient control has been lost that the alternate shutdown process must be implemented. The focus of this guidance will involve assessing, monitoring, and restoring key safety functions per EOP-01 (i.e., reactivity control, vital auxiliaries, RCS inventory control, RCS pressure control, core heat removal, RCS heat removal, and containment integrity).
LIC-1 2-0083 Enclosure Page 91 of 164
- b. Please indicate whether or not the MCR evacuation from loss of control is modeled in the FPRA. If not, describe why not.
OPPD's Response to Safe Shutdown Analysis RAI 09 b.:
The FCS fire PRA conservatively does not credit control room abandonment and recovery from the alternate shutdown panel for fire scenarios that cause a significant loss of plant control. Instead, the conditional core damage and large early release probabilities associated with the significant loss of control are assigned to the scenario, and this approach is conservative.
- c. Please clarify if fire scenarios initiated outside the MCR that may affect habitability in the MCR were considered (e.g.; heating, ventilation and air conditioning (HVAC) supply area(s)). If not, describe why not.
OPPD's Response to Safe Shutdown Analysis RAI 09 c.:
Fire scenarios occurring outside the control room that could potentially threaten control room habitability were considered in the fire PRA process. The main control room itself is enclosed within a fire-rated boundary, and therefore smoke and heat from fire outside the control room is not expected to substantially propagate into the control room. However, habitability could be affected by scenarios that cause loss of control room HVAC (either direct failure of HVAC or failure of its support systems).
Consistent with the internal events analysis, failure of the control room HVAC system was considered and qualitatively screened from the fire PRA. Operators enter AOP-13, Loss of Control Room Air Conditioning, upon either rising control room temperature, loss of air flow to the control room or computer room, or control room temperature exceeding 105 degrees Fahrenheit. AOP-13 directs operators to attempt re-establishing room cooling via any of the following:
- VA-46A or VA-46B (note that if both trains are inoperable, then Technical Specification 2.12.2 requires the reactor be placed in hot shutdown within six hours).
- Alternate alignment using portable fans and the turbine building exhaust fans, if available, drawing air from the auxiliary building, through the control room, and out to the turbine building.
- Alternate alignment using portable fans and the auxiliary building fans to draw air from the turbine building, through the control room, and out the auxiliary building.
Fire-induced loss of control room HVAC was qualitatively screened out for several reasons. First, there is a low frequency of fires with the potential to damage both VA-46A and VA-46B or their support systems. For these scenarios, at least one of
LIC-1 2-0083 Enclosure Page 92 of 164 the redundant and diverse HVAC alternatives is expected to remain available.
Secondly, the likelihood of operators failing to align either a normal or alternate form of HVAC is expected to be low. Finally, the operators would shut the plant down before high temperatures began to cause malfunctions of control room equipment, and equipment required for maintenance of shutdown conditions are not sensitive to elevated control room temperature.
- d. Please identify any deviations from the guidance in NUREG/CR-6850, MCR evacuation following both loss of control room function and loss of habitability.
OPPD's Response to Safe Shutdown Analysis RAI 09 d.:
The FCS fire PRA modeling treatment following control room abandonment does not deviate from the guidance of NUREG/CR-6850.
Safe Shutdown Analysis RAI 10:
AOP-17, "Loss of Instrument Air," has a 4-hour action time to isolate flow through condensate makeup control valve, LCV-1 190, to prevent condensate storage tank (CST) drain down into the hotwell for NFPA 805. The NRC staff is of the understanding that the isolation action is only necessary when the CST is used and the diesel driven AFW pump is credited (page 21 of the AOP). Please confirm that this is correct, and if not, provide further discussion related to when this action is required.
OPPD's Response to Safe Shutdown Analysis RAI 10:
FCS Calculation FC06748, Revision 0, "LCV-1 190 Backup Gas Source," provides the basis for the 4-hour action time associated with the AOP-17 action to isolate flow through LCV-1 190 upon depletion of backup nitrogen supply pressure for the valve operator. This action is necessary to prevent condensate storage tank (CST) draindown to the hotwell.
OPPD confirms that, with respect to NFPA 805 compliance, this action is only necessary for fire events occurring in fire areas where the diesel driven AFW pump (FW-54) is credited in LAR Table B-3 for NFPA 805 compliance. These are fire areas 28, 32, 34B-1, 34C, 36A, and 36B.
LCV-1190 is located on the turbine building mezzanine (fire area 46), as are manual valves FW-269 and FW-270, which are directed to be closed in AOP-17 to isolate CST flow through LCV-1 190 and into the hotwell. There is direct access to the turbine building from the main control room.
LIC-1 2-0083 Enclosure Page 93 of 164 Safe Shutdown Analysis RAI 11:
Table B-3, Table C-94, Fire Area 06-08, Heat Exchanger and Pump Area is one of many examples of flooding. Flooding to adjacent areas is possible and has been analyzed.
The adjacent areas affected by flood water include: corridor 4, the open stairwell next to fire area 06-3 and possibly room 23 which is located at the bottom of the stairwell.
Since no safe shutdown equipment is in any of these areas, flooding to adjacent areas is not a concern. With regard to the flooding analysis used to address the suppression effects of this fire area.
- a. Please indicate whether or not the flooding analysis credits the operation of sump pumps in the lower elevation areas.
- b. If yes, please indicate whether or not the control and power cables of the sump pumps have been included in the NSCA. Provide justification if not included.
OPPD's Response to Safe Shutdown Analysis RAI 11 a. and b.:
OPPD does not credit the operation of sump pumps or transfer pumps in the basement, ground, intermediate, or operating level areas.
Safe Shutdown Analysis RAI 12:
Attachment S, Table S-2, Committed Modifications Item REC-111 of the LAR indicates that High Energy Arcing Fault (HEAF) barriers will be installed around/near the 4 KV switchgear and bus ducts in the 4 KV Switchgear Rooms. These barriers are intended to reduce the local damage associated with a potential HEAF, and subsequently reduce the risk calculated for Fire Areas 36A and 36B.
- a. Please provide the criteria being used to design and build these barriers.
OPPD's Response to Safe Shutdown Analysis RAI 12 a.:
The intent is for the High Energy Arcing Fault (HEAF) barriers to be designed such that they limit damage to the faulted device and any components electrically dependent on that device. For example, with the barriers installed, a HEAF on the bus duct connecting Diesel Generator #1 output to the 1A3 4KV bus would fail the diesel generator power supply to 1A3, but the HEAF barriers would prevent damage to cable trays, conduits, and other potential targets surrounding the bus duct. These barriers are intended to apply to HEAF events on bus duct, switchgear, and load centers in FCS fire compartments FC36A and FC36B.
NUREG/CR-6850 Supplement 1 Section 7.2.1.5 provides guidance on determining the ZOI for bus duct HEAF events, and this guidance will be used to design the HEAF shields in FCS fire compartments FC36A and FC36B. This guidance will be applied to both bus duct HEAF and cabinet HEAF scenarios. In particular, the third bullet on page 7-10 provides examples of configurations expected to prevent damage from the HEAF, including:
LIC-1 2-0083 Enclosure Page 94 of 164
" "The solid metal side panels of a cabinet will prevent ignition of the combustible/flammable materials inside the cabinet."
" "For cabinets with a solid steel top where all cable or conduit penetrations are sealed . . . molten material deposited on top of the cabinet will not burn through the panel top."
- "For cabinet side panels or doors that include louvered ventilation openings where the louvers point downwards to the outside of the panel, molten material deposited on the surface of the panels will not penetrate into the cabinet."
" "Cables in conduit will not be ignited by molten materials deposited on the outer conduit surface if the open ends of the conduit are located outside the zone of influence."
" "Cable in trays that are equipped with unventilated steel covers will not be ignited by molten metals falling from above."
" "Cables in trays that are equipped with aluminum covers of any kind, or with ventilated steel covers, will be ignited by molten metals falling from above."
S'"The first solid surface encountered by the material ejected from the bus duct will truncate the zone of influence along that line of travel. (Examples include: where the zone of influence intersects the floor, a sealed cabinet top, or a cable tray with a solid metal cover, it does not extend through that surface to other targets or flammable material beyond)."
The above examples suggest that a steel metal shield, ranging between a sheet metal tray cover and heavy gauge plate steel used for electrical cabinets, would be sufficient to effectively truncate the HEAF ZOI and failure of cables or components behind the shield.
- b. Please indicate to what standard these barriers will be installed.
OPPD's Response to Safe Shutdown Analysis RAI 12 b.:
There is no known design standard for installation of the proposed HEAF barriers.
- c. If there is no standard to address the installation of HEAF barriers, please indicate how they will be tested to ensure suitable design and operational criteria are met.
OPPD's Response to Safe Shutdown Analysis RAI 12 c.:
OPPD plans to use the cited guidance in NUREG/CR-6850, Supplement 1 to design the HEAF shields. Testing is an integral part of the design modification process.
LIC-12-0083 Enclosure Page 95 of 164 Safe Shutdown Analysis RAI 13:
Please indicate whether or not there are any locations in the plant where a fire could cause the mechanical failure of flexible hoses/expansion joints such that the fire could result in either piping system failure or potential significant flooding concerns (i.e.,
condenser water box inlet and outlet expansion joints, diesel generator cooling water flexible joints, condensate pump suction flexible joints, etc.). If there are such locations, please provide a list of them and an assessment of the input to the NSCA for NFPA 805 compliance.
OPPD's ResDonse to Safe Shutdown Analysis RAI 13:
The plant database was searched and the table below was compiled of non-metallic (i.e.,
rubber, neoprene, etc.) bellows which could potentially lose pressure boundary integrity due to fire exposure.
Condenser A Expansion Joint Rutber Tutrine Bldg Water CW Condenser B Expansion Joint 46 Rubter Turbine Bldg Water CW Condenser FW- IA Expansion Joint 46 Rutter Tiurine Bldg Water CW Condenser FW- IA Expansion Joint 46 Rubber Turbine Bldg Water Cw Condenser Fw- I B Expansion Joint 46 Rubber Turtine Bldg Water CW Condenser Fw-IB Expansion Joint 46 Rubber Turbine Bldg Water CW Circulating Pump CW-IA Discharge Piping Expansion Joint 31 Rubbr Intake Water CW Circulating Pump CW-IB Discharge Piping Expansion Joint 31 Rubber Intake Water cw Circulating Pump CW-IC Discharge Piping Expansion Joint 31 Rubber Intake Water CW Intake Bldg Sump Pump VD-2A Discharge Expansion Joint 31 Rubbter Intake Water cw Intake Bldg Sump Pump VD-2B Discharge Expansion Joint 31 Rubber Intake Water CW Flex Connector On DW-3A 48 or 49 Not required Water Plant Sump Pit Water DW Flex Connector On DW-3B 48 or 49 Not required Water Plant Sump Pit Water DW Cond. Pump A Exp. Joint 46 Rubber, 600 Turbine Bldg Water FW-CD Cond. Pump B Exp. Joint 46 Rubber, 600 Turbine Bldg Water FW-CD Cond. Pump C Exp. Joint 46 Rubber, 600 Turbine Bldg Water FW-CD Steam Dump And Bypass To Condenser Fw-la Expansion Joint 46 Not required Turbine Bldg Steam Sr Steam Dump And Bypass To Condenser Fw- lb Expansion Joint 46 Not required' Turbine Bldg Steam Sr Condenser Drain Expansion Joint 46 Not required Turbine Bldg Water VD Condenser Drain Expansion Joint 46 Not required Turbine Bldg Water VD Condenser Drain Expansion Joint 46 Not required Turbine Bldg Water VD Condenser Drain Expansion Joint 46 Not required Turbine Bldg Water VD As described in the assessments below, failure of any of the bellows listed above would not impact the NFPA 805 compliance as described in the LAR Attachment C and FCS Engineering Evaluation EA10-036.
LIC-1 2-0083 Enclosure Page 96 of 164 A failure of the CW pump discharge bellows in the intake structure would be self-limiting as the CW pump motors would trip once the water reached the level of the CW pump motor power cable termination boxes. The CW pumps are located at elevation 976' and the CW pump motor power cable termination boxes are located at elevation 990'. The safe shutdown credited raw water pump motor power cable termination boxes in the area are located at elevation 996'-6", and the safe shutdown credited raw water valves in the area are located above elevation 993'-6" (Reference drawing 11 405-E-299, sheet 1).
The intake building sumps are located in the same area as the CW pumps and there is no credited equipment in the area that would be affected if a fire were to damage the bellows.
The water plant sump pumps are located on the 1007'-6" elevation of the service building and would not affect any safe shutdown equipment located outside of the service building. There is no safe shutdown equipment located within the service building that is credited for NFPA 805 compliance for a fire event occurring within the service building.
The rubber bellows in the Feedwater, Condensate, and CW systems located in the turbine building would not affect any safe shutdown equipment located outside of the turbine building.
There is no safe shutdown equipment located within the turbine building that is credited for NFPA 805 compliance for a fire event occurring within the turbine building.
LIC-1 2-0083 Enclosure Page 97 of 164 Radioactive Release RAI 01:
For liquid releases, Column 7 of Attachment E indicates the Fire Areas where floor drains are routed to the monitored Radioactive Waste Disposal System (RWDS). The liquids are collected in spent regenerate tanks via the floor drain header and sump pumps (pumped from lower elevations). For each area where credit for these engineering controls is taken, please provide clarification that the system is capable of handling and containing the estimated amount of water to be generated.
OPPD's Response to Radioactive Release RAI 01:
Initial fire suppression and firefighting will generate less than 30,000 gallons of wastewater per EA 99-023, "Fire Protection Suppression Effects Analysis." This analysis assumes the fire suppression system operates for 30 minutes and manual firefighting is completed within 30 minutes.
Waste water gravity drains via floor drains and/or from sumps to two Spent Regenerant Tanks (WD-13A/B), with one tank aligned having a capacity of 5,530 gallons, and then pumped to any of three Waste Holdup Tanks (WD-4A/B/C) each having a capacity of 45,800 gallons.
Should the inflow of water exceed the capacity of the Spent Regenerant Tanks, these tanks vent into the room where the Spent Regenerant Tanks reside (Room 23). Room 23 is a watertight room in the lowest level of the Auxiliary Building (971 ft level). Room 23 has sufficient volume (> 50,000 gallons) to hold the waste water prior to processing.
In the absence of Containment isolation, the Containment sump pumps (WD-3 A/B),
automatically deliver the gravity drains from the Containment structure to the Spent Regenerant Tanks. Should the Containment sump pumps be unable to maintain level, the sump will over flow into the lower level of Containment. The lower level of Containment has sufficient volume to hold the waste water (> 50,000 gallons) prior to processing through the waste disposal system.
Radioactive Release RAI 02:
For Fire Area 30, Containment, Refueling Pool, Column 7 of Attachment E states that during non-power operations, floor drains as described in Radioactive Release RAI 01 (above) remain present. It also states that "in the unlikely event that liquid effluents do escape containment, provisions are in place to contain liquid effluents and divert contaminated runoff away from the river." For this area, please provide clarification that the steps taken to contain liquid effluents comply with effluent requirements.
OPPD's Response to Radioactive Release RAI 02:
The Fire Brigade is trained with regard to precautions to be undertaken for fires involving radioactive substances in locations where fire suppression activity could result in leakage from buildings (i.e., hose stream application strategies, minimizing water quantity discharged, containment methods, etc.). The Fire Brigade uses standard industry methods, such as creating berms and using sandbags/tarps to cover drains, in order to prevent contaminated
LIC-1 2-0083 Enclosure Page 98 of 164 water runoff into the river. FCS procedure PE-RR-AE-1001, "Flood Barrier and Sandbag Staging and Installation," discusses equipment to be used for flood control. This procedure will be revised to allow this equipment to also be used by the Fire Brigade for contaminated fire water containment purposes as part of implementation. FCS procedures, including the Off-Site Dose Calculation Manual (ODCM) and the pre-fire plan in SO-G-28, will be established for RP to monitor contaminated liquid effluents prior to safe removal from the station's Protected Area.
Radioactive Release RAI 03:
The Original Steam Generator Storage Facility (OSGSF) (page E-15 of Attachment E) has been identified to be of no consequence and does not require a specific radiological calculation.
Calculation FC07865 is referenced as the basis for this conclusion. Please provide justification for screening out this area.
OPPD's Response to Radioactive Release RAI 03:
The OSGSF was not evaluated for a specific fire or radiological consequences as a result of a fire inside the facility. Two calculations that were performed for the OSGSF provide justification for not requiring a specific fire release calculation. Calculation FC07083, "Doses Due to Component Cover Plate Damage," was performed to evaluate the consequences of the OSGSF collapsing. The conclusions from that calculation indicated that the resultant dose from a radiological release from that scenario (collapsed facility) was 8.7-mRem at the EAB, 1.08-mRem at the LPZ, and 20.1 -mRem at the Control Room. The assessment was based on an assumed release of 1% of the total radionuclides in the components and of that, 1% would be of particulate size large enough to be airborne. As such, the assessment conducted for engineering change (EC) 33099 in FC07083 for a collapsed building included an effective release reduction of 0.0001. However, a factor of 10 increase was used with the atmospheric dispersion factor.
With a fire event, the potential for airborne release is expected to be higher slightly due to volatility of fire combustion products and particulate release. The only items, which could be potentially combustible inside the OSGSF, are the plastic materials covering the contaminated components. Therefore the justification for a release of 1% would still be valid, and contaminants fixed inside the components would not be combustible (metal components). The plastic covering could however result in a release of those radionuclides that were on the surface of the components, and if 10% of all the plastic covering burned inside the facility resulted in a particulate release, the total effective amount of fixed radionuclides present in the original components stored in the facility would be a factor of 0.001. This release of radionuclides would not be filtered. As such, using a scaling factor for this assessment on a fire impact would result in less than 87-mRem at the EAB, 10.8-mRem at the LPZ, and 201-mRem for the Control Room.
LIC-1 2-0083 Enclosure Page 99 of 164 Firefighting activities inside the OSGSF with water was "screened out" because the dose from an OSGSF fire event was determined to be bounded by a previous evaluation performed for flooding. Calculation FC07084, "Original Steam Generator Storage Facility Flooding Concentrations," evaluated a 1029-ft flood elevation or 7-ft flood height inside the facility.
FC07084 documented a radionuclide concentration that was less than the 10 CFR 20, Appendix B, Table 2, Column 2 concentration limits. Table 5 of FC07084 indicated that because of flooding, the sum of fractions was calculated to be 0.604 that would correspond to a TEDE of 30.2 mRem. This calculation was based upon the highest surface concentration and assumed that half of the components surface areas were submersed in floodwater. This calculation is considered bounding in regards to any firefighting activities that would be performed in the OSGSF as the water would not exceed 7 ft flood level, nor would there be an immersion of the components that would equal half of the component surface areas. From a firefighting event, only the outside surfaces would be washed down as noted above, and the only combustible component could be the plastic coverings.
Based on these arguments, the OSGSF does not require a specific radiological calculation and the existing calculations documented above are bounding from a dose consequence for potential fire events at the OSGSF.
Radioactive Release RAI 04:
For liquid releases from fires involving sea-land containers, two cases are discussed in Column 7 of Attachment E. If the sea-land containers are located within the "screened in" protected areas identified in the table, the engineering controls discussed for their respective areas can safely handle liquid effluents. However, for sea-land containers located outside of the "screened in" areas, the table states that provisions are in place to "commence communication between the Fire Brigade and Radiation Protection and to contain liquid effluents for the purpose of diverting contaminated liquid flow away from the river." Please describe the provisions that are in place for the purpose of controlling liquid effluent releases for this situation or provide a bounding analysis, quantitative analysis, qualitative analysis, or other analysis that demonstrates that the limitations for instantaneous release of liquid radioactive effluents specified in the Technical Specifications (TS) are met (FAQ 09-0056.)
OPPD's Response to Radioactive Release RAI 04:
As described in the training materials and pre-fire plans, the Fire Brigade is also trained with regard to precautions to be undertaken for fires involving radioactive substances in exterior locations (i.e., hose stream application strategies, minimizing water quantity discharged, containment methods, etc.). The Fire Brigade uses standard industry methods, such as creating berms and using sandbags/tarps to cover drains, in order to prevent contaminated water runoff into the river. FCS procedure PE-RR-AE-1001, "Flood Barrier and Sandbag Staging and Installation," discusses equipment to be used for flood control. This procedure will be revised to allow this equipment to also be used by the Fire Brigade for contaminated fire water containment purposes as part of LAR implementation. FCS procedures, including the Off-Site Dose Calculation Manual (ODCM) and the pre-fire plan in SO-G-28, will be established for RP to monitor contaminated liquid effluents prior to safe removal from the station's Protected Area.
LIC-1 2-0083 Enclosure Page 100 of 164 Radioactive Release RAI 05:
A similar explanation as discussed in Radioactive Release RAI 04 (above) is offered for sea-land containers or other radiological sources located in the temporary Maintenance Shop Lower Expansion (Lower Mezz.) and the Old Warehouse. The table also states provisions are in place to "commence communication between the Fire Brigade and Radiation Protection and to contain liquid effluents within the area and away from the river." Please describe the provisions that are in place for the purpose of controlling liquid effluent releases for this situation or provide a bounding analysis, quantitative analysis, qualitative analysis, or other analysis that demonstrates that the limitations for instantaneous release of liquid radioactive effluents specified in the TS are met.
OPPD's Response to Radioactive Release RAI 05:
As described in the training materials and pre-fire plans, the Fire Brigade is also trained with regard to precautions to be undertaken for fires involving radioactive substances in exterior locations (i.e., hose stream application strategies, minimizing water quantity discharged, containment methods, etc.). The Fire Brigade uses standard industry methods, such as creating berms and using sandbags/tarps to cover drains, in order to prevent contaminated water runoff into the river. FCS procedure, PE-RR-AE-1001, "Flood Barrier and Sandbag Staging and Installation," discusses equipment to be used for flood control. This procedure will be revised to allow this equipment to also be used by the Fire Brigade for contaminated fire water containment purposes as part of LAR implementation. FOS procedures, including the ODCM and the pre-fire plan in SO-G-28, will be established for RP to monitor contaminated liquid effluents prior to safe removal from the station's Protected Area.
Radioactive Release RAI 06:
Column 8 of Attachment E indicates that "if normal ventilation is not available, smoke will be removed using manual ventilation to the outside or to an area where normal ventilation will remove the smoke." It also states that prior to any release, the radiological hazards associated with releasing contaminated smoke from the building will be considered and direct communication between the Fire Brigade and Radiation Protection will be initiated. Please clarify the method used for "manual ventilation" and provide methods for directing smoke to areas where "normal ventilation will remove the smoke" and clarify the specific actions/methods needed to minimize and/or monitor the release of the contaminated gaseous effluent.
OPPD's Response to Radioactive Release RAI 06:
The use of permanently installed engineering controls (i.e., fixed drainage and HVAC systems) is intended to support Fire Brigade and Radiation Protection (RP) containment, monitoring, and removal operations. Both the Fire Brigade and RP are trained in responding to and assessing emergency situations with regard to the containment and removal of contaminated gaseous effluents. Pre-fire plans and training materials will describe the presence and potential use of monitored HVAC systems, if the systems are deemed operational and capable of supporting manual removal efforts. Per the references stated in the Radioactive Release
LIC-1 2-0083 Enclosure Page 101 of 164 Review, EA10-043, the Fire Brigade is trained with regard to precautions to be undertaken for fires involving radioactive substances (i.e., application of a "fog spray" directed above the release path so smaller water droplets are entrained into the gaseous release). The Fire Brigade is also trained to use HVAC systems if deemed operational and capable of handling contaminated smoke. The Fire Brigade communicates with RP regarding the availability and potential use of installed engineering controls. When deemed safe by RP to exhaust contaminated smoke and if the HVAC system cannot be used, the Fire Brigade is instructed to exhaust contaminated smoke to the outside, or to an area having an operable HVAC system, applying standard industry techniques such as the use of portable smoke ejectors and flexible ductwork. FCS procedure, SO-G-28, will be revised to minimize and/or monitor the release of the contaminated gaseous effluent as part of LAR implementation.
Radioactive Release RAI 07:
For radioactive sources in the Maintenance Shop Lower Expansion, page E-16 of Attachment E indicates that any fire involving radioactive sources in this area would never exceed the dose consequences from the combustion of the sea-land container and therefore the 10 CFR dose limits would never be exceeded from burning radioactive sources in this area. Please provide a qualitative response relating the amount of the activity contained in the radioactive sources relative to that of the amount assumed in the sea-land container.
OPPD's Response to Radioactive Release RAI 07:
Sources of radioactive material are not stored in the Maintenance Shop; however, periodically components that have low levels contamination are machined in the Maintenance Shop.
These evolutions generally occur during refueling outages. For this work, a small enclosure is built around the machine (satellite RCA). As a matter of good health physics practices, radioactive components are decontaminated to less than 20,000 dpm/1 00-cm 2 before being taken to the Machine Shop for maintenance.
The only potentially combustible materials that would be contaminated are small amounts of oil used in the machining, protective clothing, and wood and plastic used in the construction of the enclosure.
Simple calculations show the dose would not exceed 51 -mRem total effective dose equivalent (TEDE) if a satellite RCA in the Machine Shop would catch fire. These calculations conservatively assume the contamination level in the enclosure prior to a fire is 100,000-dpm/cm2 . These calculations assume a level of contamination 5 times higher than is normally acceptable. The effluent releases from a fire in the Machine Shop are less than 11% of the Technical Specification limits. The radioactivity in the Machine Shop is a small fraction (<
0.2%) of the radioactivity in a sea-land container.
LIC-1 2-0083 Enclosure Page 102 of 164 Probabilistic Risk Assessment RAI 01:
The FPRA peer review findings and observations (F&Os) are provided in Attachment V of the LAR. Please clarify the dispositions to the following F&Os:
- a. F&O PP-B2-01: Enclosure 1 to the LAR supplement dated December 22, 2011, indicates for Supporting Requirement (SR) PP-B3 that manual suppression capability was used to help define compartment boundaries.
Manual suppression is not allowed as criteria to define a compartment in NUREG/CR-6850. Therefore, please clarify this definition of compartment boundaries and modify your model if necessary, or identify this as a deviation from NUREG/CR-5860 and justify the deviation. In addition, please indicate if spatial separation is used to define any compartment boundaries and, if used, describe the criteria used to identify and accept the spatial separation boundaries.
In addition, please define and justify the criteria to establish the water curtain as a boundary between FC 6-3 and FC 20-1. Clarify if water curtains were used as a boundary for any other fire compartments and, if so, define and justify the criteria used.
OPPD's Response to Probabilistic Risk Assessment RAI 01 a.:
The FCS fire PRA does not credit manual suppression as a partitioning feature.
While manual suppression is mentioned in the context of justifying boundary adequacy of a stairwell between FC20-1 and FC20-7, the plant partitioning report will be revised to clarify that manual suppression is not credited as a partitioning feature, and instead the lack of credible fire scenarios, including consideration of intervening combustibles, is the basis for boundary adequacy regarding this stairwell. This basis was re-confirmed by plant walkdown the week of April 2nd, 2012.
The FCS fire PRA credits spatial separation as a partitioning feature for a stairwell and open hatches between FC06-3, FC20-1, and FC20-7. Boundary adequacy was confirmed in these cases by plant walkdown and engineering judgment that no credible fire scenarios were physically capable, including consideration of intervening combustibles, of causing target damage in the adjacent compartment.
A water curtain is installed at the stairwell between FC06-3 and FC20-1. While the basis for boundary adequacy, as discussed above, is the lack of fire scenarios capable of causing target damage across the compartment boundary, the water curtain provides further assurance of boundary adequacy.
Water curtains are also installed to backup several fire doors between the Turbine Building (FC46) and the following fire compartments: FC32, FC36A, FC37, FC41, FC42, and FC43. While the rated fire doors provide the basis for boundary adequacy, the installed water curtains provide further assurance of boundary adequacy.
LIC-1 2-0083 Enclosure Page 103 of 164
- b. F&Os PRM-A3-01 and FQ-Al-01: LAR Table V1 shows SR PRM-A3 as "met."
However, the "Summary" of the F&O in LAR Table V2 states that SR PRM-A3 is "NOT MET." Please clarify this discrepancy. In addition, describe the extent to which hot short probabilities are applied in the FPRA model. Please identify the components to which these hot short probabilities are applied and the basis for determining which components/cables were selected for application of circuit failure probabilities.
OPPD's Response to Probabilistic Risk Assessment RAI 01 b.:
Regarding PRM-A3, the discrepancy between LAR Tables V-1 and V-2 was carried over from the fire PRA peer review report. Specifically, the peer review report summary Table 4-17 shows PRM-A3 as met, while the F&O PRM-A3-01 text states the supporting requirement is not met because the peer reviewer believed that conditional wire-to-wire short probabilities were applied only to the pressurizer PORVs, whereas analysis of a broader set of risk significant components was expected by the reviewer. Note that while PRM-A3 is a broad supporting requirement related to construction of the plant response model, the peer reviewer focused the content of this F&O on circuit failure likelihood analysis. OPPD contends that PRM-A3 is met, and the discussion of circuit failure likelihood analysis requested by this RAI is provided below.
Circuit failure likelihood analysis was performed to apply hot short probabilities on the following components, and incorporated into the fire PRA model:
Component Description HCV-1041A STEAM GENERATOR RC-2A; MS ISOLATION VALVE HCV-1042A STEAM GENERATOR RC-2B ; MS ISOLATION VALVE HCV-1107A STEAM GENERATOR RC-2A; AUXILIARY FEEDWATER INLET VALVE HCV-1 107B STEAM GENERATOR RC-2A AUXILIARY FEEDWATER INLET VALVE HCV-1 108A STEAM GENERATOR RC-2B ; AUXILIARY FEEDWATER INLET VALVE HCV-1 108B STEAM GENERATOR RC-2B AUXILIARY FEEDWATER INLET VALVE HCV-1384 MAIN AND AUXILIARY FEEDWATER; CROSSCONNECT VALVE HCV-1385 STEAM GENERATOR RC-2B INLET ISOLATION VALVE HCV-1 386 STEAM GENERATOR RC-2A INLET ISOLATION VALVE HCV-150 PRESSURIZER RC-4; RELIEF ISOLATION VALVE HCV-151 PRESSURIZER RC-4; RELIEF ISOLATION VALVE HCV-204 LETDOWN HEAT EXCHANGER CH-7; INLET VALVE HCV-2987 HPSI ALTERNATE HEADER ISOLATION VALVE HCV-2987 HPSI ALTERNATE HEADER ISOLATION VALVE HCV-306 HPSI HEADER ISOLATION VALVE HCV-307 HPSI HEADER ISOLATION VALVE HCV-383-3 CONTAINMENT SUMP; RECIRC ISOLATION VALVE HCV-383-4 CONTAINMENT SUMP; RECIRC ISOLATION VALVE HCV-385 SIRW TANK SI-5 RECIRCULATION VALVE HCV-386 SIRW TANK SI-5 RECIRCULATION VALVE
LIC-1 2-0083 Enclosure Page 104 of 164 HCV-438A RCP RC-3A-D LUBE OIL & SEAL CLRS; CCW INLET INBOARD ISOLATION HCV-438B RCP RC-3A-D LUBE OIL & SEAL CLRS; CCW INLET OUTBOARD ISO VALVE HCV-438C RCP RC-3A-D LUBE OIL & SEAL CLRS; CCW OUTLET INBOARD ISO VALVE HCV-438D RCP RC-3A-D LUBE OIL & SEAL CLRS; CCW OUTLET OUTBOARD ISOL LCV-383-1 SIRWT SI-5 OUTLET HEADER LEVEL CONTROL VALVE LCV-383-2 SIRWT SI-5 OUTLET HEADER LEVEL CONTROL VALVE PCV-1 02-1 PZR POWER OPERATED RELIEF VALVE PCV-102-2 PRESSURIZER; POWER OPERATED RELIEF VALVE PCV-910 MS HEADER TO CONDENSER BYPASS CONTROL VALVE TCV-202 REACTOR COOLANT SYSTEM LOOP 2A; LETDOWN TEMPERATURE TCV-909-1 MS HEADER TO CONDENSERS; DUMP VALVE TCV-909-2 MS HEADER TO CONDENSERS; DUMP VALVE TCV-909-3 MS HEADER TO CONDENSERS; DUMP VALVE TCV-909-4 MS HEADER TO CONDENSERS; DUMP VALVE YCV-1045 AUX FEEDWATER PUMP FW-10; INLET VALVE YCV-1045A MAIN STEAM LINE "A"TO ; AUX FEEDWATER PUMP FW-10 ; SUPPLY YCV-1045B MAIN STEAM LOOP "B" ; AUX FEEDWATER PUMP FW-10 ; SUPPLY VALVE The above components were identified based on risk significance, as determined by reviewing fire PRA cutsets throughout model development, as well as experience with the internal events PRA. The identified components are generally valves whose spurious closure would degrade frontline mitigating systems (e.g.,
auxiliary feedwater, high pressure safety injection, etc.) or threaten RCS integrity (e.g., spurious opening of pressurizer PORVs).
- c. F&O FSS-A2-01: Based on discussions during the audit, the ZOI for various ignition sources is defined by a column above the ignition source. However, targets beyond those identified using this columnar ZOI could be damaged if a fire propagates to trays above the ignition source due to the 35 degree (")
spread of fire in a cable tray stack (see Appendix R of NUREG/CR-6850).
iL Please confirm whether cable trays exist in the region defined by the vertical column and the 35° angle, yet lie outside of the columnar ZOL Provide an assessment of the importance of the cable trays lying within this region.
OPPD's Response to Probabilistic Risk Assessment RAI 01 c.i.:
A plant walkdown was performed April 2-6, 2012, to assess whether fire PRA targets exist outside the modeled ZOI but within the 35 degree spread of fire upward through a cable tray stack. This walkdown included all areas in which the zone of influence fire modeling approach was implemented. The walkdown did not identify any cases in which this effect required revising (adding to) the existing source-target data set.
LIC-1 2-0083 Enclosure Page 105 of 164 Examination of a typical example bolsters the sensibility of the walkdown conclusion. Given a cable tray stack of four trays, each separated by one foot, the fire would be expected to propagate horizontally (4)(Tan 35) = 2.8 feet. Starting at the edge of the ignition source, 2.8 feet horizontally is generally bounded by the modeled zones of influence.
ii. Please describe how the evaluation includes the possible increase in HRR caused by the spread of a fire from the ignition source to other combustibles.
OPPD's Response to Probabilistic Risk Assessment RAI 01 c.ii.:
The FCS Fire PRA supporting the LAR models fire propagation from the ignition source to overhead cable trays. The fire growth profile of the ignition source and ignited cable tray configuration are summed to obtain the overall fire scenario growth profile, which is used to determine if and when hot gas layer formation may occur. The fire PRA does not model propagation to secondary combustibles other than cable trays, and this is addressed in OPPD's response to Fire Modeling RAI 01 b.
- d. F&O FSS-A4-01: In response to this F&O, a subset of plant components (the most risk significant) were examined and 153 cases were corrected where the target set was insufficient. It is not clear what shortcoming in the original approach produced these omissions. Please explain this oversight and its root cause. Also, please provide the extent of the sample process in context of the full FPRA, including describing what percentage of the PRA components were sampled. If further examination of components is warranted, describe what needs to be done and the results of the additional examination.
OPPD's Response to Probabilistic Risk Assessment RAI 01 d.:
The fire PRA has approximately 15,000 unique scenario-cable combinations. A scenario-cable combination is defined as a particular fire scenario that is expected to fail a particular cable. The sampling study conducted to assess the accuracy of the fire PRA source-target data sampled 4,973 scenario-cable combinations, or approximately 33% of the total fire PRA data set. Sampled cables included VFDR cables (for relevance to the NFPA 805 application) as well as a sampling of cables whose associated components had high fire risk significance as indicated by early fire PRA quantifications.
The sampling study indicated 153 errors of the 4,973 sampled scenario-cable combinations, or approximately 3%. These errors were corrected in the final fire PRA model. An example of an error is that a particular conduit traverses through a scenario ZOI one-of-influence but was not identified as a target for that scenario.
The cause of these errors was associated with challenges reading the cable and raceway layout drawings, which tend to be high density, cluttered, and with hand-annotated raceway labels. In addition, the FCS electrical raceways are not field labeled.
LIC-1 2-0083 Enclosure Page 106 of 164 This study and resulting improvements to the fire PRA data provided confidence that the source-target data is adequate for the purposes of the fire PRA and its application to NFPA 805. No further sampling or validation was determined necessary.
- e. F&O FSS-D8-01: A determination of no outlier behavior is required to meet capability category (CC)-Il of the corresponding SR. Please evaluate plant history and determine if the Halon system has revealed outlier behavior relative to system availability. Also, if outlier behavior relative to system unavailability exists include this behavior in your PRA or show how the fire watch and backup suppression is as effective and reliable as the Halon system.
OPPD's Response to Probabilistic Risk Assessment RAI 01 e.:
The Halon system is subject to routine surveillance testing to ensure high system availability. Plant condition reports covering the time period between 1995 and early April 2012 were reviewed to determine if the system has experienced outlier behavior with respect to unavailability. This review did not identify any repeated patterns of system unavailability, and therefore the generic system failure data presented by NUREG/CR-6850 is considered applicable to the FCS. OPPD now considers Supporting Requirement FSS-D7 to be met at CC-Il.
Furthermore, if the Halon system is declared non-functional a continuous fire watch with backup suppression is established. The continuous fire watch with backup suppression is judged to be of similar reliability and effectiveness as the Halon system. The fire watch would be able to quickly detect the fire, notify the control room, and implement initial suppression activities if appropriate.
- f. F&O FSS-E1-01: Please discuss the difference in CDF, LERF, ACDF, and ALERF of using the draft of FAQ 08-50 rather than the final version as documented in NUREG/CR-6850 Supplement 1.
OPPD's Response to Probabilistic Risk Assessment RAI 01 f.:
A sensitivity study was performed to assess the impact of using the draft version of FAQ 08-50 rather than the final version in NUREG/CR-6850 Supplement 1.
The FCS fire PRA primarily credits manual suppression to prevent hot gas layer formation. The probability of manual non-suppression is calculated as e`At using an average Avalue of 0.078, whereas NUREG/CR-6850 Supplement 1 cites an average Avalue of 0.067. Using a reduced lambda value has the effect of increasing the non-suppression probability.
A sensitivity study was performed by varying the Avalue between 0.067 and 0.078.
The revised Avalue increased the CDF and LERF slightly for several scenarios, as summarized in the following table:
LIC-1 2-0083 Enclosure Page 107 of 164 A = 0.078 A = 0.067 Scenario CDF(lyr) LERF (lyr) CDF(/yr) LERF (/yr)
FC10-IS1-011100 2.41 E-07 1.63E-08 2.43E-07 1.64E-08 FC10-1S2-OiI1OO 4.02E-08 2.71 E-09 4.05E-08 2.73E-09 FC10-1S3-OiI1OO 1.22E-07 8.14E-09 1.23E-07 8.20E-09 FC31-IS1-OillOO 9.1OE-09 7.80E-11 9.89E-09 8.46E-11 FC31-1S2-Oi1lOO 5.19E-09 4.45E- 11 5.64E-09 4.82E-11 FC31-1S3-OiI1OO 2.59E-09 2.22E-1 1 2.82E-09 2.41 E-1 1 FC31-1S4-OillOO 1.30E-09 1.12E-1 1 1.42E-09 1.21 E-1 1 FC31-1S5-OiI1OO 3.89E-08 3.16E-10 3.94E-08 3.20E-10 FC31-1S6-OiI1OO 3.89E-08 3.16E-10 3.94E-08 3.20E-10 FC31-1S7-0ill00 3.89E-08 3.16E-10 3.94E-08 3.20E-10 FC32-1S7-OiI100 3.20E-08 2.55E-09 3.22E-08 2.57E-09 FC41-PW-TFWC-4 3.67E- 10 2.65E-1 1 3.76E-10 2.72E-1 1 FC41 -PW-TRAN-4 1.79E-09 1.29E- 10 1.83E-09 1.32E- 10 TOTAL 5.72E-07 3.10E-08 5.79E-07 3.12E-08 The net CDF and LERF increases for these scenarios are 6.64E-09 /yr and 2.29E-10 /yr, respectively. The following table extends this sensitivity study for the overall plant risk.
Base Fire PRA* Sensitivity Study**
(A= 0.078Case) (A = 0.067 Case)
Net ACDF for NFPA 805 Transition (/yr) 5.72E-06 5.73E-06 Net ALERF for NFPA 805 Transition (Iyr) 6.67E-07 6.67E-07 Total CDF (internal, flood, fire) (/yr) 6.01 E-05 6.01 E-05 Total LERF (internal, flood, fire) (/yr) 4.82E-06 4.82E-06
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, for the range of Avalues assessed by this sensitivity study.
- g. F&O FSS-E4-01: The LAR describes several cable routing assumptions that were made. Please explain how these assumed routings were treated on a fire scenario basis. For example, a cable known to be in two different trays in the same physical analysis unit (PAU) or in adjacent PAUs is assumed to traverse between the two trays. Please explain how this cable is treated in the PRA model for the different ignition sources present in the PAU since the path of the cable between the trays is unknown.
LIC-12-0083 Enclosure Page 108 of 164 OPPD's Response to Probabilistic Risk Assessment RAI 01 g.:
Response to Safe Shutdown (SSD) RAI 03 identifies assumptions made while determining the fire compartments through which each fire PRA cable is routed.
Assumptions made during this process carry directly into the fire PRA quantification of scenarios where all cables in the compartment are conservatively assumed to fail. These assumptions are well-founded as explained in response to Safe Shutdown RAI 03 and are therefore judged to be a negligible source of uncertainty for the fire PRA, especially for compartments in which all ignition sources are conservatively assumed to fail all targets in the compartment.
The underlying reason for making these assumptions is that the complete cable route is sometimes not shown on the cable and raceway layout drawings. The target identification process for fire scenarios using the zone-of-influence approach relies heavily on the cable and raceway layout drawings. Therefore, the absence of cables from the drawings creates some potential for cables to be erroneously excluded from the target sets of scenarios physically capable of failing those cables. The following paragraphs discuss the potential fire PRA impacts of the uncertainty associated with each cable routing assumption.
Assumption "a" of SSD RAI 03 is no longer applicable to any cables as the assumption has been verified per response to SSD RAI 03.
Assumption "b" of SSD RAI 03 is applied to runs of cable from a numbered tray to an end device, where that short run is not explicitly shown on the cable and raceway layout drawings. These short runs are either connecting conduits or airdrops. The fire PRA impact of these short runs not being shown on the cable and raceway layout drawings is judged to be minimal because:
- The final raceway in a cable route sequence is typically directly above or in the immediate vicinity of its end device. In such cases, the raceway itself would be identified as a target (provided it is within the scenario ZOI),
and the cable would therefore be failed during fire PRA quantification regardless of a portion of the final route not being identified on the drawings.
" The length of airdrops is typically of similar scale to, and often smaller than, the modeled ZOls. It is therefore unlikely that an airdrop cable would pass through a ZOI without the originating raceway also being within the ZOI. This generalization was substantiated via plant walkdown (April 2-6, 2012) of all fire scenarios in which the ZOI approach was applied. It is also supported by a sampling review of the cable and raceway layout drawings for fire areas in which this assumption was applied.
LIC-1 2-0083 Enclosure Page 109 of 164 Assumption "c" of SSD RAI 03 is applied when short runs of cable, not shown on the cable and raceway layout drawings, are routed from one end device to another end device, where both end devices are either in the same fire area or two adjacent fire areas. A hypothetical example is a cable that runs from a junction box inside containment to a valve inside containment. The NSCA would apply Assumption "c" and model this cable as being entirely routed inside containment.
This assumption is appropriate and reasonable for the fire PRA modeling of scenarios where all cables within the compartment are assumed to fail. For compartments in which the zone-of-influence approach is implemented, the absence of these conduits from the cable and raceway layout drawings could cause individual cable targets to be erroneously excluded from the target sets of fire scenarios physically capable of damaging those cables.
OPPD judges this issue to not significantly affect the results and conclusions of the fire PRA and NFPA 805 transition, primarily because the assumed routing applies to relatively short runs of cable, it applies to a limited number of components, and additional bases provided in the following table. Note that this table identifies component failures associated with cable routing that applied Assumption "c" and which the indicated cable length was greater than or equal to 10 feet (note that assumed routing often applies to only a portion of the cable length).
LIC-1 2-0083 Enclosure Page 110 of 164 Fire Component Discussion of Potential Risk Compartment Cable(s) Component Description Significance Assumption "c" applies to relatively short runs of cable.
In this area, there is minimal fire impact on AFW, and fire is not expected to induce a LOCA that would necessitate CCW (to support once through cooling and recirculation).
While fire-induced loss of CCW would cause a loss of RCP seal cooling, the ability to trip EA4220A Component Cooling the RCPs is expected to EA4220B Water Heat Exchanger remain available. Also, fire in EA4220C AC-1A; Raw Water Inlet this area will not spuriously FC06-3 EA4220D HCV-2880A Valve open a pressurizer PORV.
EA4220A Component Cooling EA4220B Water Heat Exchanger EA4220C AC-1 A; Raw Water FC06-3 EA4220D HCV-2880B Outlet Valve See above for FC06-3 CCW EB4222A Component Cooling EB4222B Water Heat Exchanger EB4222C AC-1 B; Raw Water Inlet FC06-3 EB4222D HCV-2881A Valve See above for FC06-3 CCW EB4222A Component Cooling EB4222B Water Heat Exchanger EB4222C AC-1 B; Raw Water FC06-3 EB4222D HCV-2881 B Outlet Valve See above for FC06-3 CCW Assumption "c" applies to relatively short runs of cable.
While the shutdown cooling heat exchanger supports containment sump recirculation, there is a relatively low frequency of fire in this area that could significantly impact AFW or Shutdown Cooling Heat RCS integrity (i.e., fire in this EB4257C Exchanger AC-4A; CCW area does not present a strong FC06-3 EB4257D HCV-480 Inlet Valve demand for recirculation)
Shutdown Cooling Heat Exchanger AC-4B; CCW FC06-3 EA4256C HCV-481 Inlet Valve See above for FC06-3 SDC Shutdown Cooling Heat Exchanger AC-4A; CCW FC06-3 EA4256C HCV-484 Outlet Valve See above for FC06-3 SDC Shutdown Cooling Heat EB4257C Exchanger AC-4B; CCW FC06-3 EB4257D HCV-485 Outlet Valve See above for FC06-3 SDC
LIC-1 2-0083 Enclosure Page 111 of 164 Fire Component Discussion of Potential Risk Compartment Cable(s) Component Description Significance Assumption "c" applies to relatively short runs of cable.
Fire-induced failure of this valve affects containment isolation capability. The only ignition sources in the vicinity of these cables are VA-24A and VA-24B (Coordinate A-5 of 11405-E-66), which are fans that are fully enclosed within their associated ductwork and not expected to damage the cables of concern.
Note that failure of this valve also prevents capability to isolate an ISLOCA (%13Q) on EB6111A the RCP Seal Cooler CCW EA61 11 B Containment Instrument lines. %13Q requires a passive EA61 11C Air Supply Outboard pipe failure and is therefore not FC20-7 EA61 11D PCV-1 849B Pressure Control Valve a credible fire-induced event.
Assumption "c" applies to relatively short runs of cable.
Fire-induced failure of this valve affects containment isolation capability. There are no fixed or transient fire ignition sources postulated in the vicinity of these cables (Coordinate A-4 of 11405-E-93 Sheet 1).
Note that failure of this valve also prevents capability to isolate an ISLOCA (%13Q) on EA6117B the RCP Seal Cooler CCW EA6117C Containment Instrument lines. %13Q requires a passive EA6117D Air Supply Inboard pipe failure and is therefore not FC30 EA6117E PCV-1 849A Pressure Control Valve a credible fire-induced event.
LIC-12-0083 Enclosure Page 112 of 164 Fire Component Discussion of Potential Risk Compartment Cable(s) Component Description Significance Assumption "c" applies to relatively short runs of cable.
The relevant cable terminates at YCV-1045, the steam supply valve for FW-10. Other than FW-10 itself, the only ignition source in the vicinity of this cable is the motor driven auxiliary feed water pump FW-
- 6. The fire PRA models FW-6 fires as damaging only the pump itself, as the only credible fire source is its motor (9 8th percentile heat release rate of 69 kW) and the limited quantity of lube oil (0.5 gallons). This cable will be walked down to verify its Auxiliary Feed water routing could not be affected FC32 EB12191G FW-10 Pump; Turbine-Driven by FW-6 fire scenarios.
The fire PRA models compartment conservatively assuming that all FC33 ignition Component Cooling sources fail all PRA targets in EC4223A Water Heat Exchanger the compartment. Failure of EC4223B AC-1C; Raw Water Inlet this valve is included in the FC33 EC4223C HCV-2882A Valve quantification of FC33.
Component Cooling EC4223A Water Heat Exchanger EC4223B AC-1 C; Raw Water FC33 EC4223C HCV-2882B Outlet Valve See Above for FC33 ED4225A Component Cooling ED4225B Water Heat Exchanger ED4225C AC-1 D; Raw Water Inlet FC33 ED4225D HCV-2883A Valve See Above for FC33 Component Cooling ED4225A Water Heat Exchanger ED4225B AC-1 D; Raw Water FC33 ED4225C HCV-2883B Outlet Valve See Above for FC33 All modeled fire scenarios in FC35B postulate failure of DG-
- 2. Therefore, modeling the Diesel Generator DG-2; exact routing of the relevant Room Fresh Air Supply cables would not increase the FC35B ED13404A YCV-871A Damper reported fire CDFs and LERFs.
Diesel Generator DG-2; See above for FC35B Room Fresh Air Supply FC35B ED13405A YCV-871B Damper Diesel Generator DG-2; See above for FC35B Room Fresh Air Supply FC35B ED9590A YCV-871 C Damper Diesel Generator DG-2; See above for FC35B Room Fresh Air Supply FC35B ED9591 A YCV-871 D Damper
LIC-1 2-0083 Enclosure Page 113 of 164 Fire Component Discussion of Potential Risk Compartment Cable(s) Component Description Significance Assumption "c" applies to relatively short runs of cable.
In addition, a plant modification is proposed for the relevant cables subject to Assumption C (7700A and 7700B) by R2008-004-002. This modification will eliminate the only risk-relevant component failure (DG-2 due to load shed 7700A failures) associated with these FC36A 7700B 4A3/1 B4A-7 480V Feed to MCC-4A3 cables in FC36A.
Assumption "c" applies to relatively short runs of cable.
Plant modifications cited in the R2008-004-002 evaluation of FW-8B and FW-8C in FC36A will eliminate the only risk-relevant component failure (DG-2 due to load shed Condenser Evacuation failures) associated with FW-FC36A 11160 FW-8B Pump 8B and FW-8C.
Condenser Evacuation See above for FC36A Cond FC36A 11160 FW-8C Pump Evac Pump Assumption "c" applies to relatively short runs of cable.
The fire PRA assumes failure of the relevant cable (B1641 B) causes a loss of 161 kV offsite power. Loss of 161 kV is already assumed for most modeled FC36B scenarios near bus 1A4 and in FC32.
Although modeling the exact routing of the relevant cables is therefore not expected to appreciably increase the reported fire CDFs and LERFs, the cables will be walked down to verify their routing could not Breaker Unit for 1A24 - be affected by fire scenarios 345 KV Standby Feed to not already modeled to fail 161 FC36B, FC32 B1641B 1A4-18 Bus 1A4 kV offsite power
LIC-1 2-0083 Enclosure Page 114 of 164 Fire Component Discussion of Potential Risk Compartment Cable(s) Component Description Significance Assumption "c" applies to relatively short runs of cable.
The fire PRA assumes failure of the relevant cable (B1655A) causes a loss of 161 kV offsite power. Loss of 161 kV is already assumed for most modeled FC36B scenarios near bus 1A4 and in FC32.
Although modeling the exact routing of the relevant cables is therefore not expected to appreciably increase the reported fire CDFs and LERFs, the cables will be walked down to verify their routing could not Breaker Unit for 1A44 - be affected by fire scenarios 161KV Normal Feed to not already modeled to fail 161 FC36B, FC32 B1655A 1A4-20 Bus 1A4 kV offsite power Assumption "c" applies to relatively short runs of cable.
The applicable cables are 7700A and 7700B, and the only associated risk-relevant component failure (DG-2 due to load shed failures). Since DG-2 is already assumed failed for most damaging fire scenarios, modeling the exact routing of the relevant cables is not expected to appreciably 7700A increase the reported fire FC36B 7700B 4A3/1 B4A-7 480V Feed to MCC-4A3 CDFs and LERFs.
Assumption "c" applies to relatively short runs of cable.
PT-1 18 is one of four cues for operators to diagnose and mitigate a spuriously opened pressurizer PORV. The fire HRA conservatively assumes all fire scenarios in FC34A fail PT-1 18, and therefore modeling the exact routing of Pressurizer RC-4 the relevant cables would not Narrow Range Pressure increase the reported fire FC34A B1236 PT-118 Transmitter CDFs and LERFs.
LIC-1 2-0083 Enclosure Page 115 of 164 Fire Component Discussion of Potential Risk Compartment Cable(s) Component Description Significance The PRA model credits the MSIVs to isolate main steam during mitigation of an SGTR, which is not a credible fire-induced event. The PRA also credits the MSIVs to remain open, providing makeup to the condenser, and a suction source for MFW. However, MFW is conservatively not credited by the fire PRA.
Finally, the MSIVs are conservatively not modeled to isolate transient induced excess steam flow via failure to close the steam dump and bypass valves. In conclusion, fire impact to MSIV cables in which Assumption "c" was applied does not affect the EB4928A Steam Generator RC-2A reported fire CDF and LERF FC43 EB4928N HCV-1041A MSIV results.
EB4930A Steam Generator RC-2B FC43 EB4930N HCV-1 042A MSIV See above for FC43 MSIV Assumption "c" applies to relatively short runs of cable.
Failure of one or more of these valves in the open position could cause excess steam flow to the condenser, which is conservatively modeled as a main steam line break.
5022C However, there is minimal 5022F TCV-909- potential fire impact on 5022J 1/2/3/4- Condenser Steam Dump mitigating systems in the FC46 5022M PCV-910 and Bypass Valves turbine building.
In summary, cables not shown on the cable and raceway layout drawings and subject to Assumption "c" are judged to not significantly affect the results and conclusions of the fire PRA and NFPA 805 transition.
Assumption "d"of SSD RAI 03 is applied to short runs of cable when the fire area location of only one end device is known. In such cases, the fire area location for the second end device is assumed to be located in the same fire area as the known end device. Per response to RAI SSD 03, this assumption applies only to a couple of cables where the indicated cable length is greater than 10 feet.
Furthermore, this assumption is specific to the location of an end device, which is of less importance to the fire PRA than the actual cable routing. Cases requiring use of Assumption "d" are therefore judged to be a negligible source of uncertainty for the fire PRA.
LIC-12-0083 Enclosure Page 116 of 164 Assumption "e" of SSD RAI 03 is that cables from one end device in the containment to another end device Jin the containment are assumed to be contained entirely within the containment. The components to which this assumption is applied are not explicitly modeled by the fire PRA, and therefore the underlying issue requiring use of Assumption "e" (cable routing not documented on the drawings) has no impact on the fire PRA results supporting the NFPA 805 transition.
Assumption "f"of SSD RAI 03 is that cables from one end device in the intake structure to another end device in the intake structure are assumed to be contained entirely within the intake structure. This assumption is applied to three cables, each with an indicated length of only one foot. Furthermore, the plant modification proposed for the relevant cables (A6527C, B6528C, and D6529C) in R2008-004-002 will eliminate the risk-relevant impact of these cable failures (potential degradation of offsite power, bus 1A4, and DG-2 due to electrical faults).
Assumption "g" of SSD RAI 03 is that cables connecting panels within the main control room (for cables not shown on the cable and raceway layout drawings) are assumed to route from the control room to the cable spreading room and back to the control room. The underlying lack of routing data for these cables does not impact the fire PRA, which developed control room source-target data by reviewing the cables terminating in each panel, reviewing the instrumentation and controls on the panel, and by reviewing the general design function of the panel.
- h. F&O FSS-H8-01: The multi-compartment analysis (MCA) assumes 30 minutes is available for manual suppression credit prior to the fire propagating into the neighboring compartment.
- i. Please summarize how fire suppression is generally included in your evaluation.
ii. Please discuss the basis for the 30-minute assumption, including the rating of barriers defining fire compartments.
iii. Please provide justification for the use of the manual suppression failure probability of 0.074 using the methods described in FAQ 08-OOSO or otherwise justify the estimate. Describe the results of the MCA for areas where spatial separation is credited.
OPPD's Response to Probabilistic Risk Assessment RAI 01 h.i., ii., and iii.:
The multi-compartment analysis has been upgraded in response to this RAI. In order to incorporate the upgraded analysis into the requested sensitivity studies, this RAI response will be submitted on August 25, 2012, as discussed with the NRC technical reviewers during the June 26, 2012, teleconference. [AR 525083
LIC-12-0083 Enclosure Page 117 of 164
- i. F&O IGN-B5-01: The generic fire ignition frequencies used in the FPRA are taken from NUREG/CR-6850, Supplement 1. Section 10 of NUREG/CR-6850, Supplement 1, states that a sensitivity analysis should be performed when using these fire ignition frequencies rather than the fire ignition frequencies provided in Table 6-1 of NUREG/CR-6850. Please provide a sensitivity analysis of the impact on using the Supplement 1 frequencies instead of the Table 6-1 frequencies on CDF, LERF, ACDF, and ALERF for all of those bins that are characterized by an alpha that is less than or equal to one.
OPPD's Response to Probabilistic Risk Assessment RAI 01 i.:
NUREG/CR-6850 Supplement 1 requires a sensitivity study be performed for bins where the ignition frequency distribution has an alpha value of less than or equal to one, and the following table identifies these bins.
NUREG/CR-6850 NUREG/CR-6850 Bin Description Supplement 1 Alpha Beta (mean, per rx-yr)
(mean, per rx-yr) 1 Batteries 3.26E-04 0.5 1534 7.5E-04 4 Main Control Board 8.24E-04 1 1212.9 2.5E-03 9* Air Compressors 4.65E-03 0.5 1075.3 2.4E-03 Cable fires caused by 11 welding and cutting, 9.43E-04 1 1060.5 2.OE-03 Plant Wide 13 Dryers 4.20E-04 0.5 1189.9 2.6E-03 Electrical Cabinets 15.1 EaF non-HEAF 2.36E-02 0.453 19.16 4.5E-02 Reactor Protection 22** System Motor 9.33E-04 0.92 985.87 1.6E-03 Generator Sets Cable fires caused by 31 welding and cutting, 4.50E-04 0.5 1110.2 1.6E-03 Turbine Building I I I
- Note that NUREG/CR-6850 Supplement 1 states that a sensitivity analysis need not be performed for Bin 9 Air Compressor, as the reported alpha value appears to be in error.
- FCS does not have reactor protection system motor generator sets.
A sensitivity study was performed in which the NUREG/CR-6850 Supplement 1 fire frequencies with an alpha less than or equal to one were replaced with the fire frequencies provided in NUREG/CR-6850 Table 6-1. The following table summarizes the results of this sensitivity study for fire compartments containing VFDRs.
LIC-1 2-0083 Enclosure Page 118 of 164 Base Case Sensitivity Sensitivity FC Description ase Case ALERF Case ACDF Case ACDF (Iyr) (/yr) (/yr) ALERF (/yr)
Personnel Corridor 20-1 Area (Rooms 26, 4.84E-08 2.68E-09 8.03E-09 4.46E-09 31, 58)
VA-46A and VA-20.7 46B condenser N/A N/A N/A N/A area on the ROOF auxiliary building (Note 1) (Note 1) (Note 1) (Note 1) roof Room 71 N/A N/A N/A N/A 28 (Note 1) (Note 1) (Note 1) (Note 1) 31 Intake 7.81E-08 6.31E-10 7.81 E-08 6.31 El0 32 Compressor Area 5.69E-08 3.68E-09 5.69E-08 3.69E-09 (Note 2) (Note 2) (Note 2) (Note 2)
Electrical ElcrclN/A N/A N/A N/A 34A Penetration Area N/A N/A N/A N/A Basement (Note 3) (Note 3) (Note 3) (Note 3)
Electrical 34B-1 Penetrations Area 4.34E-08 2.02E-10 7.33E-10 3.69E-10
- Ground and Intermediate Levels 36A East Switchgear 1.47E-07 9.22E-10 1.71 E-07 1.10E-09 Area 36B West Switchgear 4.02E-07 1.91 E-08 6.30E-07 3.55E-08 Area 41 Cable Spreading 3.44E-07 1.81 E-07 8.64E-08 8.64E-09 Room (Note 2) (Note 2) (Notes 2 & 4) (Notes 2 & 4)
Control Room 4.59E-06 4.59E-07 7.30E-06 7.30E-07 Complex (Note 5) (Note 5) (Note 5) (Note 5)
EFWST Area 5.32E-08 4.49E-10 5.34E-08 4.57E-10 (Room 81) (Note 2) (Note 2) (Note 2) (Note 2) 5.72E-06 6.67E-07 8.46E-06 7.85E-07 Totals Note 1: VFDR failures determined to be not fire risk relevant. Therefore ACDF and ALERF calculations are not applicable.
Note 2: A bounding approach was used to characterize the ACDF and ALERF. The reported "ACDF" and "ALERF" represent the total fire risk for the compartment, reflective of the as-built plant design. Since eliminating the VFDR failures would result in a net decrease in the as-built fire risk, the ACDF and ALERF cannot be higher than the total as-built CDF and LERF, which are within the Regulatory Guide 1.174 acceptance criteria.
Note 3: It was determined that no fire scenario could physically induce the VFDR failure of concern. Therefore, the ACDF and ALERF are not applicable.
LIC-1 2-0083 Enclosure Page 119 of 164 Note 4: The reported ACDF and ALERF for the cable spreading room sensitivity case include credit for the alternative shutdown process per AOP-06. This credit was not included in the values reported in the LAR. CCDP and CLERP values of 0.1 and 0.01 were assumed for alternate shutdown, consistent with the main control room analysis. These values may be modified as a result of analysis performed in response to PRA RAI 01 j.
Note 5: A bounding approach was used to characterize the ACDF and ALERF for the main control room. The reported "ACDF" and "ALERF" represent the total fire risk associated with main control room abandonment and use of the alternate shutdown process. Also, note that the main control room CDF and LERF reported in the LAR erroneously excluded the plant availability factor, and this resulted in slightly conservative risk estimates. The main control room CDF and LERF reported here for the sensitivity case include the plant availability factor.
The following table extends this sensitivity study for the overall plant risk.
Base Fire PRA* Sensitivity Study Net ACDF for NFPA 805 Transition (/yr) 5.72E-06 8.46E-06 Net ALERF for NFPA 805 Transition (/yr) 6.67E-07 7.85E-07 Total CDF (internal, flood, fire) (/yr) 6.01 E-05 6.86E-05 Total LERF (internal, flood, fire) (/yr) 4.82E-06 5.60E-06
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, for this sensitivity study.
- j. F&O HRA-A1-01: Calculations FC07826 and FC07824 state that the abandonment criteria currently in AOP-06 will be revised to reflect the assumption in the fire PRA that operators will evacuate only when it becomes uninhabitable due to heat and/or smoke or a significant loss of control has occurred. This procedure change is implementation item REC-096 in Table S-2 of the LAR. Please address the following with regard to how MCR abandonment was modeled in the fire PRA:
iL Please state if any fires outside of the MCR cause MRC abandonment due to a significant loss of control. If so, please state how many.
OPPD's Response to Probabilistic Risk Assessment RAI 01 i.i.:
A severe fire in the cable spreading room (FC41) could cause control room abandonment due to loss of control per AOP-06. There are no other plant locations, aside from the control room itself, where fire could require control room abandonment.
LIC-1 2-0083 Enclosure Page 120 of 164 ii. While the FPRA models MCR abandonment due to uninhabitability, loss of control does not appear to have been modeled. Please provide justification for not modeling MCR abandonment due to loss of control.
OPPD's Response to Probabilistic Risk Assessment RAI 01 i.ii.:
The FCS fire PRA conservatively does not credit control room abandonment and recovery from the alternate shutdown panel for fire scenarios that cause a significant loss of plant control. Instead, the conditional core damage and large early release probabilities associated with the significant loss of control are assigned to the scenario, and this approach is conservative.
iii. Assumption 1 of Calculation FC07824 states that a conditional core damage probability (CCDP) of 0.1 is assumed for scenarios that cause MCR abandonment and use the alternate shutdown panel. A conditional large early release probability (CLERP) of 0.01 is also assumed. Please confirm that these conditional probabilities include all failures of the operators and of mitigating equipment. Please provide justification that the CCDP of 0.1 and CLERP of 0.01 are appropriate in light of the multitude of recovery actions needed to implement plant shutdown via the alternate shutdown panel.
OPPD's Response to Probabilistic Risk Assessment RAI 01 i.iii.:
The FCS fire PRA supporting the LAR assumes, for each core damage and large early release, a single overall probability of failure to prevent these end-states using the alternate shutdown process upon fire-induced control room abandonment. This overall probability is intended to be conservative and include the contribution of both human and hardware failures. Regarding hardware failures, note that the FCS fire protection license basis requires that a minimum set of safe shutdown equipment, controls, and instrumentation remain available to mitigate control room fire scenarios that require abandonment.
NUREG-1921 (March 2012 Draft submitted for ACRS Review) Section 5.1.3 provides the following guidance with respect to using a single overall failure probability for control room abandonment scenarios:
NUREG/CR-6850 suggests that the use of a single overall failure probability value to represent the failure of reaching safe shutdown using alternative means can be used if the probability value is evaluated conservatively and a proper basis is provided. It notes that this approach was used in several IPEEE submittals and that, in many cases, 0.1 was used as a point value estimate for the probability.
LIC-1 2-0083 Enclosure Page 121 of 164 Before crediting this approach, the analyst must have applied the criteria discussed in Section 4.3 for assessing the feasibility of the operator action(s) associated with that HFE. Additionally, Section 4.8 provides qualitative analysis considerations for modeling MCR abandonment.
Section 4.3 of NUREG-1921 provides guidance for assessing the feasibility of operator actions. OPPD analysis EA10-041 includes a systematic assessment of the alternate shutdown process and concludes that it is feasible. Consistent with NUREG-1921, this assessment includes consideration of the following attributes: timing, staffing, indication, procedures and training, action location accessibility, communication, lighting, and availability of equipment and tools.
Section 4.8 of NUREG-1921 provides guidance for assessing if and when operators will abandon the main control room. This guidance focuses on loss of habitability, which is quantitatively modeled by the FCS fire PRA, and significant loss of plant control, which is discussed in response to PRA RAI 01j.ii.
In conclusion, OPPD considers its use of a single overall failure probability for control room abandonment scenarios to be generally conservative and consistent with the guidance in NUREG-1921.
iv. Section 4.3.3 of Calculation FC07824 states that control room abandonment due to main control board (MCB) fires is not postulated (i.e., NSP = 0.0) and a qualitative justification is given for this assumption. Please provide a quantitative justification for not postulating MCR abandonment due to MCB fires.
OPPD's Response to Probabilistic Risk Assessment RAI 01 i.iv.:
The mean fire frequency for the Main Control Board (MCB) modeled by the FCS fire PRA is 8.24E-04 /yr. The MCB is protected by a smoke-actuated automatic halon system, which has a 5.OE-02 non-suppression probability per NUREG/CR-6850 Appendix P. The manual non-suppression probability is approximated using Figure L-1 of NUREG/CR-6850, which provides manual non-suppression probabilities specific to the MCB as a function of distance of fire spread. Conservatively assuming unqualified cable (which is more conservative than qualified cable in this context) and assuming that fire must spread at least 0.5 meters on the MCB to threaten abandonment either due to habitability or lost control yields a manual non-suppression probability of 5.OE-03 (per Figure L-1 of NUREG/CR-6850). Finally, overall probabilities for failing to prevent core damage and large early release from the alternate shutdown panel of 1.OE-01 and 1.OE-02 are assumed. Refer to RAI PRA-01 j.iii for discussion of these values. This yields the following CDF and LERF estimates for abandonment caused by MCB fires:
LIC-1 2-0083 Enclosure Page 122 of 164 S OCDF = (8.24E-04)(5.OE-02)( 5.OE-03)(1.OE-01) = 2.06E-08 /yr
- LERF = (8.24E-04)(5.0E-02)( 5.0E-03)(1.0E-02) = 2.06E-09 /yr The above conservative estimates of CDF and LERF are not considered "significant" in the sense that they are neither within the top 95% of total fire CDF and LERF, nor contribute greater than 1% of total CDF and LERF reported by the LAR.
- k. SR HRA-A4: There is no F&O against this supporting requirement and the peer review team assessed this SR to meet only CC-I. SR HRA-A4 requires a talk-through with plant operations and training personnel to confirm interpretation of relevant procedures. The justification for not performing a talk-through is that the human failure events (HFEs) modeled in the FPRA are already modeled in the Internal Events (IEPRA) and have already received significant operator, training, and PRA input. Appendix G of the LAR indicates, however, that there are new operator actions that have been added to the FPRA. Also, operator performance and reliability for performing actions corresponding to HFEs in general is potentially impacted when performed in response to a fire. In light of this, a talk-through with plant operations and training personnel is necessary to meet the requirements of SR HRA-A4. Please conduct this talk-through and provide a summary of the results.
OPPD's Response to Probabilistic Risk Assessment RAI 01 k.:
An interview with plant operations and training personnel were conducted on April 4, 2012. Participants included an operations shift manager, the operations training supervisor, a fire PRA analyst, and a senior PRA analyst. The interview initially focused on the generalities of fire response, including:
" What indications alert the operating crew to the occurrence of fire?
- What initial actions are taken and which procedure(s) are implemented?
- How and when is the fire brigade dispatched, and what is its composition?
- How are EOP and fire response procedure implementation coordinated?
- How does the operating crew communicate and coordinate with the fire brigade?
Next, the interviewees were asked to identify the factors they consider would most challenge their successful completion of operator actions during a fire event as compared to mitigation of internal initiating events. The interviewees identified various types of distractions, communication with offsite facilities and fire departments, potential for unintended or spurious alarms, timing challenges, and offsite NRC interface.
Finally, each human failure event and its fire PRA treatment were systematically discussed. This discussion focused on how fire might challenge completion of each action, including:
LIC-1 2-0083 Enclosure Page 123 of 164
- Sequence of events and timing,
" Procedure usages,
- Diagnosis (cues) and execution,
- Potential recovery
" Additional stressors, as compared to mitigation of internal initiating events
" Action location and travel paths (for ex-control room actions)
" Fire PRA treatment The operator interviews confirmed the appropriateness of the current fire HRA assumptions and process. The discussions provided a variety of insights for additional refinement and consideration. A summary of the major insights is provided below:
The fire PRA treatment of diagnosis is generally conservative. The interviewees explained that they would use the fire response procedure to understand which plant instrumentation can be relied upon and would not be significantly affected by instrumentation failures that the procedure identifies to be within the fire-affected compartment. Also, the interviewees identified a variety of diverse cues conservatively not credited by the current fire PRA, for example using cold leg temperature as a surrogate or supplement for steam generator level.
- Fire increases the time required to diagnose the need for, and execution of, operator actions. The primary delays are associated with implementation of the fire brigade and use of the fire response procedures in concert with the EOPs. This impact is at least partially offset by the time delay generally expected between fire detection and onset of an initiating event.
- Operators generally felt that stress levels during a fire event would likely be similar to those during mitigation of an internal initiating event. In this regard, the fire PRA treatment of operator actions is conservative in that it increased the expected stress by one level when calculating human error probabilities for actions minimally impacted by the fire.
- Operators agreed with the fire PRA treatment of actions that are either executed within the fire location or whose travel paths are impacted by fire effects. The fire PRA generally fails these actions, with the exception of long-term actions that could be completed after the fire was extinguished (e.g.,
EFWST refill within eight hours).
SR HRA-B3: There is no F&O against this SR and the peer review team assessed this SR to meet only CC-1. SR HRA-B3 requires the completion of HFE definitions. Table V3 of the LAR states that the HFE definitions are generally performed and defined in the FCS HRA Calculator. Please provide further justification that this SR meets CC-Il by addressing how the FPRA meets each of the requirements of the SR for CC-Il, or provide a justification that CC-I is acceptable for this application.
LIC-1 2-0083 Enclosure Page 124 of 164 OPPD's Response to Probabilistic Risk Assessment RAI 01 I.:
According to the fire PRA peer review report, the CC-I assessment of HRA-B3 references F&O HRA-A1-01, which identifies that the fire PRA does not reflect the as operated plant, since it relies on a future procedure modification to be performed during the 10 CFR 50.48(c) license implementation period. It is acceptable and expected at LAR review stage of the NFPA 805 application that risk insights would not yet be incorporated into the procedures.
Regarding the actual requirements of HRA-B3, the FCS fire PRA human failure event definitions consider accident sequence-specific timing of cues and time window for successful completion, availability of cues, and the specific high-level tasks (e.g., train level). This level of human failure event definition is consistent with CC-Il of HRA-B3.
Probabilistic Risk Assessment RAI 02:
Calculation FC07821 describes the fire ignition frequency development methodology and results. The second footnote to Table 4-2 provides the severity factors that were applied for main feedwater pump oil fires and states that these are in accordance with FAQ 08-44. However, these severity factors are not consistent with this FAQ, as described in Section 9 of NUREG/CR-6850, Supplement 1. Please provide an assessment of the impact on CDF, LERF, ACDF, and ALERF of applying the NUREG/CR-6850, Supplement 1 severity factors.
OPPD's Response to Probabilistic Risk Assessment RAI 02:
The FCS fire ignition frequency calculation applies the following main feedwater pump oil fire severity factors in accordance with the version of FAQ 08-044 used to develop the fire PRA:
- 95.5% of the oil fire frequency damages only the ignition source itself
& 4.4% of the oil fire frequency involves 10% of the oil inventory
- 0.09% of the oil fire frequency involves 98% the oil inventory (note that the FCS calculation actually applied 100% of oil inventory to this severity factor)
Supplement 1 to NUREG/CR-6850 (page 9-5) specifies the following main feedwater pump oil severity factors:
9 96.6% of the oil fire frequency damages only the ignition source itself
- 3.06% of the oil fire frequency involves 10% of the oil inventory
- 0.34% of the oil fire frequency involves 100% the oil inventory The FCS main feedwater pumps are located in the Turbine Building, FC46. There are no VFDRs associated with the Turbine Building, and therefore the applied severity factors do not affect the ACDF or ALERF calculations supporting the NFPA 805 fire risk evaluations.
LIC-1 2-0083 Enclosure Page 125 of 164 The total plant CDF and LERF are re-calculated here using the NUREG/CR-6850 Supplement 1 severity factors. The current and re-calculated main feedwater pump oil fire CDF and LERF are provided in the following table.
Current Re-R- Current Re-R- CurrentRe Re-Scenario scrio Bin Current Calculated Current Calculated Current Calculated Description B A (yr) A (/yr) CDF (Iyr) CDF (/yr) LERF (Iyr) LERF (/yr)
FW-4C (Small Oil) 32 1.51 E-03 1.52E-03 1.03E-07 1.03E-07 5.39E-09 5.43E-09 FW-4C (10% Oil) 32 6.93E-05 4.82E-05 4.71 E-09 3.28E-09 2.47E-10 1.72E-10 FW-4C (100% Oil) 32 1.42E-06 5.36E-06 1.45E-10 5.47E-10 8.82E-12 3.33E-11 FW-4B (Small Oil) 32 1.51 E-03 1.52E-03 1.03E-07 1.03E-07 5.39E-09 5.43E-09 FW-4B (10% Oil) 32 6.93E-05 4.82E-05 4.71 E-09 3.28E-09 2.47E-10 1.72E-10 FW-4B (100% Oil) 32 1.42E-06 5.36E-06 1.45E-10 5.47E-10 8.82E-12 3.33E-11 FW-4A (Small Oil) 32 1.51 E-03 1.52E-03 1.03E-07 1.03E-07 5.39E-09 5.43E-09 FW-4A (10% Oil) 32 6.93E-05 4.82E-05 4.71 E-09 3.28E-09 2.47E-10 1.72E-10 FW-4A (100% Oil) 32 1.42E-06 5.36E-06 1.45 E-10 5.47E-1 0 8.82E-12 3.33E-1 1
_________________ -~ 4 I. +
Sum 3.23E-07 3.22E-07 1.69E-08 1.69E-08 Note that the above calculations conservatively apply the entire Bin 32 frequency to the oil fire cases. Separating out the motor fire split fraction (modeled with the same impact as the small oil fire due to lack of PRA targets near the main feedwater, pumps) .would result in an 11%
(Reference NUREG/CR-6850 Table 6-1) reduction in the above oil fire frequencies. The above calculations include the oil fire severity factors, generic Bin 32 frequency from NUREG/CR-6850, Supplement 1 (5.44E-03 /yr), and the FCS plant availability factor (86.9%).
The CCDP and CLERP for each small oil and 10% oil case are 6.80E-05 and 3.57E-06, respectively, and the CCDP and CLERP for each 100% oil case are 1.02E-04 and 6.21 E-06, respectively.
The following table extends this sensitivity stud for the overall plant risk.
Base Fire PRA* Sensitivity Study Net ACDF for NFPA 805 Transition (/yr) 5.72E-06 5.72E-06 Net ALERF for NFPA 805 Transition (/yr) 6.67E-07 6.67E-07 Total CDF (internal, flood, fire) (/yr) 6.01 E-05 6.01 E-05 Total LERF (internal, flood, fire) (/yr) 4.82E-06 4.82E-06
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174 when the current main feedwater pump oil fire severity factors are replaced by the severity factors in NUREG/CR-6850, Supplement 1, Chapter 9.
LIC-12-0083 Enclosure Page 126 of 164 Probabilistic Risk Assessment RAI 03:
Section V of the LAR (FPRA Quality), Table V1, "Peer Review Team Assessment of Capability Categories for all SRs in ASME/ANS [American Society for Mechanical Engineers/American Nuclear Society] RA-Sa-2009 Part 4," indicates that SRs FSS-C8, PRM-B13, PRM-B15, HRA-A2, and HRA-B2, are not applicable to the FPRA. Please provide justification' for this conclusion for each of these SRs.
OPPD's Response to Probabilistic Risk Assessment RAI 03:
The non-applicability of each supporting requirement specified by this RAI is discussed in each of the subsequent bullet points.
FSS-C8 requires, for each electrical raceway fire barrier credited by the fire PRA, establishing a technical basis for the fire resistance rating and confirming that the barrier will not be subject to mechanical damage or direct flame impingement from a high hazard ignition source. At the time of the peer review, this supporting requirement was erroneously judged by the self-assessment and by the peer review team to be not applicable with the erroneous basis that the fire PRA did not credit raceway fire barriers.
The following table identifies each area where raceway fire barriers are credited by the fire PRA and provides an assessment for each case against FSS-C8.
Fire ERFBS Description FSS-C8 Assessment Summary Compartment FC31A Pyrocrete enclosures The fire PRA does not directly credit this ERFBS protecting cables supporting because it is not routed through any intake structure RW pumps, RW pump fire scenario zones of influence, with the exception of discharge valves, and RW the hot gas layer scenarios, in which case all RW is crosstie valves as identified in assumed to fail, regardless of the ERFBS, due to the R2008-004-002 Revision 3. RW pumps and valves physically located in this compartment.
FC34A Fire wrap protecting 3A1,3B1, The fire PRA does not directly credit this ERFBS and 3C1 power feeder cables. because it is not routed through any FC34A fire scenario zones of influence.
FC36C Pyrocrete enclosure in West Consistent with the NFPA 805 Nuclear Safety Switchgear room protecting Capability Assessment (NSCA), the fire PRA credits various cables. this pyrocrete enclosure as a fire compartment boundary.
The fire resistive effectiveness of this fire barrier against switchgear room fire scenarios is assessed.
The enclosure may be subject to mechanical damage due to high energy arcing faults in the switchgear room, and this will be considered in the design of the proposed high energy arcing fault shields. The intent of the proposed shields is to limit damage to the faulted bus (and any components dependent on the faulted bus). So, the proposed shields will ensure that cables within the FC36C pyrocrete barrier are not compromised.
LIC-1 2-0083 Enclosure Page 127 of 164 FC36A Pyrocrete enclosure protecting The fire resistive effectiveness of this fire barrier various Train B cables in against switchgear room fire scenarios is assessed.
vertical Tray Sections 10S and 22S, and horizontal conduit The enclosure may be subject to mechanical damage bank from Tray Section 57S in due to high energy arcing faults in the switchgear FC36B to Al-1 09B in FC36A. room, and this will be considered in the design of the proposed high energy arcing fault shields. The intent of the proposed shields is to limit damage to the faulted bus (and any components dependent on the faulted bus). So, the proposed shields will ensure that cables within the subject pyrocrete barriers are not compromised.
PRM-B13 provides data analysis requirements for probability input values that either require reanalysis given the fire context or that were not included in the internal events PRA. The peer review determined this supporting requirement is not applicable because the FCS fire PRA did not require alternation of existing basic event probabilities and the probability of new fire-specific basic events were either set to zero or logical TRUE during scenario quantification.
The fire PRA did not require altering existing basic event probabilities. If a particular basic event could be induced by a fire scenario, it was set TRUE during the CCDP and CLERP quantifications of that scenario. If a particular basic event could not be induced by a fire scenario, its base probability from the internal events PRA model was used during quantification of that scenario such that random failures contribute to the fire sequences.
The exception to the above is application of conditional hot short probabilities. For example, if a particular fire scenario could impact PORV control cables and potentially cause spurious operation, the conditional probability of the basic event representing PORV spurious operation may be set to some probability (instead of TRUE) in accordance with a Circuit Failures (CF) requirements of the ASME/ANS RA-Sa-2009 standard.
A number of new basic events were added to the fire PRA model. These new basic events represent fire risk-relevant failures not already included in internal events PRA.
The base probabilities of these new basic events were set to zero in the fault tree. If a particular basic event could be induced by a fire scenario, it was set TRUE during the CCDP and CLERP quantifications of that scenario.
- PRM-B15 provides modeling requirements for new accident progressions beyond the onset of core damage to determine fire-induced LERF. The peer review assessed this supporting requirement as not applicable because the FCS fire PRA did not identify any new accident progressions. The existing internal events PRA accident sequence modeling was sufficient for the fire PRA.
" HRA-A2 requires review and identification of fire risk-relevant operator actions specified by the fire response procedures but not already modeled by the internal events PRA.
LIC-1 2-0083 Enclosure Page 128 of 164 The peer review assessed this supporting requirement as not applicable with the logic that the FCS fire PRA did not credit mitigating actions from the fire response procedures; only operator actions modeled by the internal events PRA were carried into the fire PRA. This assessment is substantially correct, with the exception that the control room abandonment alternate shutdown process, which is a fire response procedure action, credited by the fire PRA but not by the internal events PRA. The fire PRA models the alternate shutdown process with a single overall failure probability, as described in response to PRA RAI 01 j.iii. OPPD considers HRA-A2 to be met.
HRA-B2 requires incorporation into the plant response model any new fire risk relevant human failure events corresponding to the operator actions identified per HRA-A2. The peer review assessed this supporting requirement as not applicable with the logic that the FCS fire PRA did not credit mitigating actions from the fire response procedures; only operator actions modeled by the internal events PRA were carried into the fire PRA. This assessment is substantially correct, with the exception that the control room abandonment alternate shutdown process, which is a fire response procedure action credited by the fire PRA but not by the internal events PRA. The fire PRA models the alternate shutdown process with a single overall failure probability, as described in response to PRA RAI 01 j.iii. OPPD considers HRA-B2 to be met.
Probabilistic Risk Assessment RAI 04:
Attachment A of Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," Revision 2, provides the NRC staff clarifications and qualifications on each of the FPRA SRs in the ASME/ANS RA-Sa-2009 PRA Standard. Please clarify how the clarifications and qualifications were addressed during the Peer Review of the FPRA. If the peer review did not address the RG 1.200, Rev.2 clarifications and qualifications, develop and provide that assessment.
OPPD's Response to Probabilistic Risk Assessment RAI 04:
The following table provides an assessment of the FCS fire PRA against each of the Regulatory Guide 1.200 Revision 2 clarifications and qualifications to Part 4 of ASME/ANS-RA-Sa-2009. Note that the peer review team did not document any assessments against these clarifications and qualifications.
Index FCS Assessment Table 4-2.2-1 HLR-ES-A The peer review assessment remains valid in the context of this clarification, which is a grammatical clarification.
Table 4-2.2-2(a) HLR-ES-A The peer review assessment remains valid in the context of this clarification, which does not impose any new or modified requirements.
ES-Al The peer review assessment remains valid in the context of this clarification, which does not impose any new or modified requirements.
ES-B1 The peer review assessment remains valid in the context of this clarification.
The FCS fire PRA did not limit component selection to the internal events PRA. Several systematic reviews documented in CN-RAM-09-024 were
LIC-1 2-0083 Enclosure Page 129 of 164 Index FCS Assessment performed specifically to identify potentially fire risk significant component failures beyond those modeled by the internal events PRA.
The FCS fire PRA is consistent with the NRC clarification to NOTE-ES-B1 -7 since both functional and spurious failure modes are considered for components assumed to be failed for all fire scenarios (e.g., main feedwater).
ES-B4 The peer review assessment remains valid in the context of this clarification, which does not impose any new or modified requirements.
ES-Cl The peer review assessment remains valid in the context of this clarification, which is intended to reduce ambiguity in the definition of the term "significant" by referencing the Part 1 definitions.
CS-A10 The peer review assessment remains valid in the context of this clarification, which eliminates the CC-I and CCII distinction between fire areas and physical analysis units. Note that the peer review team assessed the FCS fire PRA at CC-Ill for Supporting Requirement.
FSS-A2 The peer review assessment remains valid in the context of this clarification. The FCS fire PRA specifies for each target set the cable, component, and basic event failures, including spurious operation.
FSS-A4 The peer review assessment remains valid in the context of this clarification, which replaces the ambiguous term "one or more" with the less ambiguous term "sufficient".
FSS-A5 The peer review assessment remains valid in the context of this clarification, which clarifies that the number of individual scenarios defined and level of detail should be commensurate with fire risk significance of the physical analysis unit.
FSS-C2 The peer review assessment remains valid in the context of this clarification, which is intended to reduce ambiguity in the definition of the term "significant" by referencing the Part 1 definitions, similar to the ES-C1 clarification.
FSS-D3 The peer review assessment remains valid in the context of this clarification, which better articulates the differences between CC-I, CC-Il, and CC-Ill.
FSS-F1 The peer review assessment remains valid in the context of this clarification, which replaces the ambiguous term "one or more" with the less ambiguous term "sufficient".
IGN-A1 The peer review assessment remains valid in the context of this clarification, which provides clarification on the use of non-nuclear fire event data. The FCS fire PRA did not require use of non-nuclear fire event data.
QNS-C1 The peer review assessment remains valid in the context of this clarification, which requires that the quantitative screening criteria be based on total fire risk, as opposed to total internal events risk as currently written in this Supporting Requirement. The FCS fire PRA conservatively did not implement quantitative screening in order to maximize application to the NFPA 805 transition.
CF-Al The peer review assessment remains valid in the context of this clarification, which is intended to reduce ambiguity in the definition of the term "significant" by referencing the Part 1 definitions.
LIC-1 2-0083 Enclosure Page 130 of 164 Index FCS Assessment HRA-D1 [Note (1)] The peer review assessment remains valid in the context of this clarification, which corrects an editorial issue with the Supporting Requirement indexing.
Table 4-2.12-1 HLR-FQ-E The peer review assessment remains valid in the context of this clarification, which is intended to reduce ambiguity in the definition of the term "significant" by referencing the Part 1 definitions.
FQ-E1 The peer review assessment remains valid in the context of this clarification, which is intended to reduce ambiguity in the definition of the term "significant" by referencing the Part 1 definitions.
FQ-F1 The peer review assessment remains valid in the context of this clarification, which is intended to reduce ambiguity in the definition of the term "significant" by referencing the Part 1 definitions.
4-3.2 The peer review assessment remains valid in the context of this clarification, which clarifies that fire HRA expertise is required for the peer review. The FCS fire PRA peer review team had the appropriate expertise in fire HRA.
References The peer review assessment remains valid in the context of this clarification, which clarifies that the NRC neither approves nor disapproves the information provided in each document referenced by the standard.
Appendix 4-A The peer review assessment remains valid in the context of this clarification, which clarifies that the NRC neither approves nor disapproves the information provided in Appendix 4-A.
Probabilistic Risk Assessment RAI 05:
According to Calculation FC07819, pages 35-36, those actions included in the Fire Response Procedures that could be adverse to plant risk are not modeled. Please provide justification for the exclusion of these actions from the FPRA. The justification should summarize the type and number of actions that could be adverse to plant risk and why the actions are not expected to be taken. In addition, please explain if these particular actions have been discussed with operators who might potentially perform such actions and whether such actions were considered during the human reliability analysis (HRA).
OPPD's Response to Probabilistic Risk Assessment RAI 05:
The FCS fire response procedure AOP-06, Fire Emergency, does not direct operators to implement actions that could degrade plant safety. The purpose of this procedure is to assist operators in safely shutting down the plant in the event of fire. AOP-06 is written with a symptomatic philosophy such that equipment will only be shut down if the fire is threatening that equipment or if required by the fire brigade to extinguish the fire. The fire PRA model is consistent with this procedure in that any fire scenario threatening particular equipment is quantified with that equipment failed.
The exception to this symptomatic philosophy is fire in the main control room for which AOP-06 could direct abandonment under certain conditions and implementation of alternate shutdown (potentially prior to significant component failures or loss of habitability). This was recognized
LIC-1 2-0083 Enclosure Page 131 of 164 during the NFPA 805 transition, and as a result AOP-06 will be revised as part of LAR implementation to not require abandonment unless either inhabitability or significant loss of plant control has occurred or is expected.
Probabilistic Risk Assessment RAI 06:
Four deviations from NUREG/CR-6850 are identified in one-sentence descriptions in Section 4.5.1.2 of the LAR. A sensitivity study of these four deviations that calculate the aggregate CDF and LERF when these deviations are replaced by methods aligned with NUREG/CR-6850 is provided in Attachment W of the LAR. Please describe each of these deviations from NUREG/CR 6850 in greater detail. Please identify what values were changed, and which fire areas were impacted for each deviation, and include a description of the analyses leading to the PRA results. In addition, please discuss the intentions regarding the treatment of each of these deviations in the FPRA post-transition.
OPPD's Response to Probabilistic Risk Assessment RAI 06:
As identified in Section 4.5.1.2 of the NFPA 805 LAR and its supplement dated December 19, 2011 (LIC-1 1-0130), the FCS FPRA applies several methods outside of NUREG/CR-6850 that have not been formally "approved" by the NRC Frequently Asked Question (FAQ) process or the peer review process EPRI has set up for new FPRA methods. Sensitivity studies have been performed to demonstrate that FCS meets the NFPA 805 acceptance criteria, even if the alternate methods are replaced by methods within the guidance of NUREG/CR-6850 and its supplemental documents.
Post-transition, OPPD plans to remove these alternate methods from the NFPA 805 fire PRA, replacing them with methods within the guidance of NUREG/CR-6850 and its supplemental documents. These alternate methods were applied early in the FCS fire PRA development and are no longer necessary given the more recent advances in fire PRA methods.
Each alternate method is described in more detail in the following sections.
Pump Fire Frequency Apportioninq The current fire PRA apportions Bin 21 pump frequency based on pump runtime. The intent of this approach was to assign a higher frequency to normally running pumps and a lower frequency to standby pumps. NUREG/CR-6850 apportions pump fire frequency based on pump count and does not explicitly consider pump runtime.
This approach is implemented in the ignition frequency calculation and impacts the fire frequency for all areas that contain pump ignition sources. Each pump was assigned a pump runtime fraction, rounded to the nearest 10%. A backstop of 10% was used, even for purely standby pumps. The Bin 21 frequency was then apportioned to each pump based on its runtime fraction. The total Bin 21 frequency was conserved.
LIC-1 2-0083 Enclosure Page 132 of 164 A sensitivity study was performed to assess the CDF, LERF, ACDF, and ALERF impacts of using this method. In conclusion, the NFPA 805 acceptance criteria are still met when this method is removed from the fire PRA and replaced with the NUREG/CR-6850 method.
Application of this alternate method had a trivial impact on the fire PRA results, as the pumps for which this approach originally had an appreciable impact were ultimately supplemented with additional fire modeling and other refinements within the guidance of NUREG/CR-6850.
Diesel Generator Fire Treatment The current fire PRA assumes that some fraction of diesel generator fires occur during a surveillance or other attended activity and do not cause failures beyond the diesel generator itself. Specifically, it was assumed that only 3% of the diesel generator fire frequency would be unsuppressed manually and result in damage beyond the diesel generator. This was based on a review of the fire events contributing to the diesel generator fire frequency, which indicated that the vast majority of reported diesel generator fires occurred during attended surveillance or maintenance. The 3% is implemented for fire compartments FC35A and FC35B as a 0.03 severity factor.
This concept is no longer considered to be valid due to ongoing industry and NRC efforts to refine the EPRI Fire Events Database, which indicate that the reported fires occurring during surveillance testing could also have occurred during a plant transient involving automatic start of the diesel generator(s), in which case prompt suppression would likely not be available.
A sensitivity study was performed to assess the CDF, LERF, ACDF, and ALERF impacts of using this method. In conclusion, the NFPA 805 acceptance criteria are still met when this method is removed (i.e., the 0.03 factor is set to 1.0) from the fire PRA and replaced with a ZOI approach in which the specific targets affected by each ignition source are examined.
Use of Draft Version of FAQ 08-50 The current fire PRA implements a draft version of FAQ 08-50 Manual Non-Suppression Probability. OPPD response to PRA RAI 01f discusses how this draft version of FAQ 08-50 was implemented and includes a sensitivity study comparison against use of the approved final version of FAQ 08-50 documented in NUREG/CR-6850 Supplement 1. In conclusion, the NFPA 805 acceptance criteria are still met when the final version of FAQ 08-50 is implemented in the Fort Calhoun Station fire PRA.
Electrical Cabinet Generic Severity Factor The current FCS FPRA assumes that 10% of electrical cabinet fires will damage targets external to the cabinet of origin. This approach implicitly and generically credits manual suppression, and therefore no additional credit for manual suppression is taken. This approach is based on a review of the electrical cabinet fire events used to calculate the cabinet fire frequency, which indicated very little evidence of fires damaging targets outside the cabinet of origin. The 10% is applied to Bin 15.1 electrical cabinets as a 0.10 severity factor.
Current industry efforts exploring generic conditional probabilities of target damage external to the cabinet of origin will likely supersede the approach implemented by the FCS fire PRA.
LIC-12-0083 Enclosure Page 133 of 164 A sensitivity study was performed to assess the CDF, LERF, ACDF, and ALERF impact of using this method. In conclusion, the NFPA 805 acceptance criteria are still met when this method is removed (i.e., the 0.10 factors are set to 1.0) from the fire PRA.
Probabilistic Risk Assessment RAI 07:
Deviations from NUREG/CR 6850 beyond the four identified in Section 4.5.1.2 of LAR were noted. Regarding these deviations, please address the following:
- a. The first of these deviations is crediting a hot work procedural non-compliance factor of 0.01 for Compressor Area Room 19 (FC32), Switchgear Room FC36A, and Switchgear Room FC36B. This deviation is the subject of Item #3 in the supplement to the LAR dated December 19, 2011. For this deviation, it appears that the non-compliance factor should be removed since this factor is implicitly included in the hot work fire frequency. Please indicate any other physical analysis units/fire areas in which this factor is applied. For all cases, indicate the effect of removing this factor on CDF, LERF, ACDF, and ALERF.
- b. The second deviation is crediting a continuous fire watch on storage of combustibles through a non-suppression factor of 0.01. In the case of the roof of FC32, this fire watch credit is applied for all combustible loading; in the case of the Cable Spreading Room (FC41), this fire watch credit is applied for combustibles above 5 lbs. The use of the non-suppression factor must be justified, keeping in mind that the continuous fire watch generally provides detection, but not necessarily suppression. The scenario for the roof should take into account the full range of combustibles that are expected to be located on the roof, as well as the possible locations of those combustibles on the roof. Should the fire watch perform a suppression function, it should be confirmed that one of the fire watch's purposes is to extinguish the fire, that an extinguisher will be readily available, and that the fire watch will have undergone adequate training in the use of extinguishers.
Also, the time over which the combustibles are to be placed in each location must be described. Please provide the effect of the second deviation on CDF.
LERF, ACDF, and ALERF. For the justification of non-suppression, address the detection and suppression timing, the combustible storage limit, and any other key assumptions. If this type of non-suppression credit is used elsewhere in the fire PRA, identify and discuss along similar lines, ensuring that the physical analysis unit/fire area in which it is used is identified.
- c. Please identify and describe any other deviation from NUREG/CR-6850 not already identified and described in this RAI or in Section 4.5.1.2 of the LAR (the subject of another RAI). Please provide a sensitivity analysis of each of these other deviations from NUREG/CR-6850. Also, provide the composite impact on CDF, LERF, ACDF, and ALERF from the sensitivity studies applied upon removal of all deviations identified in this RAI and the four deviations identified in the LAR.
LIC-1 2-0083 Enclosure Page 134 of 164 OPPD's Response to Probabilistic Risk Assessment RAI 07:
The fire PRA has been systematically reviewed for the NUREG/CR-6850 deviations. OPPD needs some additional time to perform and document the sensitivity studies demonstrating that the NFPA 805 acceptance criteria are still met without the use of NUREG/CR-6850 deviations beyond those identified in the LAR. Therefore, as discussed with the NRC technical reviewers during the June 26, 2012, teleconference and as agreed upon via a July 3, 2012, email from the NRC NFPA 805 Transition Project Manager, OPPD will provide the responses to PRA RAI 07 items a., b., and c. on September 28, 2012. [AR 52508]
Probabilistic Risk Assessment RAI 08:
Calculation EA 10-039 identified approximately 50 valves and instruments where failure likelihood was evaluated using option 1 (Table values 10-1 thru 10-5) of NUREG/CR-6850 Chapter 10. It was recently stated at the industry fire forum that the Phenomena Identification and Ranking Table Panel being conducted for the circuit failure tests from the DESIREE-FIRE and CAROL-FIRE tests may be eliminating the credit for Control Power Transformer (CPT) (about a factor 2 reduction) currently allowed by Tables 10-1 and 10-3 of NUREG/CR-6850, Vol. 2, as being invalid when estimating alternating current (AC) circuit failure probabilities. Please provide a sensitivity analysis that removes this CPT credit from the PRA and provide new results that show the impact of this potential change on CDF, LERF, ACDF, and ALERF.
OPPD's Response to Probabilistic Risk Assessment RAI 08:
EA10-039 documents circuit failure likelihood analysis on a number of components. Analysis of a subset of these components credits the probability reduction afforded by the presence of a control power transformer per NUREG/CR-6850 Task 10 Option 1.
A bounding sensitivity study was performed by re-quantifying the fire PRA model without any credit for conditional wire-to-wire short probabilities. That is, all basic events for which a conditional probability had been developed in EA1 0-039, regardless of whether the component control circuit includes a control power transformer, were set TRUE by this model re-quantification. The sensitivity study results are summarized in the following table.
Base Fire PRA* Sensitivity Study**
(crediting short (not crediting short probabilities) probabilities)
Net ACDF for NFPA 805 Transition (/yr) 5.72E-06 5.96E-06 Net ALERF for NFPA 805 Transition (/yr) 6.67E-07 9.32E-07 Total CDF (internal, flood, fire) (/yr) 6.01 E-05 6.03E-05 Total LERF (internal, flood, fire) (/yr) 4.82E-06 5.09E-06
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099
LIC-1 2-0083 Enclosure Page 135 of 164
The relative insensitivity of the overall fire PRA results to application of conditional wire-to-wire short probabilities is due to the large fire risk contribution of switchgear room fires, which cause broad de-energization (non-spurious) failures in the electrical distribution system. Main control room fire scenarios also contribute a significant fraction of total fire risk, and the benefit of conditional short probabilities is generally minimal, especially for scenarios that cause abandonment.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, even if conditional wire-to-wire short probabilities included in the base fire PRA are not credited. By extension, this bounding assessment demonstrates that the NFPA 805 acceptance criteria can still be met even without the probability reduction afforded by the presence of a control power transformer.
Probabilistic Risk Assessment RAI 09:
Please describe the methodology that was used to evaluate defense-in-depth. The description should include what was evaluated, how the evaluation was performed, and what, if any, actions or changes to the plant or procedures were taken to maintain the philosophy of defense-in-depth.
OPPD's Response to Probabilistic Risk Assessment RAI 09:
The methodology that was used to evaluate defense-in-depth is documented in procedure EPM-DP-RSD-001, "Fire Risk Evaluation", Revision 1, dated August 2010. The methodology is as follows:
Defense-in-Depth Approach Consistency with the defense-in-depth philosophy is maintained if the following acceptance guidelines, or their equivalent, are met:
" A reasonable balance is preserved among 10 CFR 50.48(c) defense-in-depth elements.,
- Over-reliance and increased length of time or risk in performing programmatic activities to compensate for weaknesses in plant design is avoided.
- Pre-fire nuclear safety system redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences of challenges to the system and uncertainties (e.g., no risk outliers). (This should not be construed to mean that more than one safe shutdown/NSCA train must be maintained free of fire damage.)
" Independence of defense-in-depth elements is not degraded.
0 Defenses against human errors are preserved.
LIC-1 2-0083 Enclosure Page 136 of 164
- The intent of the General Design Criteria in Appendix A to 10 CFR Part 50 is maintained.
A review of the impact of the VFDRs on defense-in-depth shall be performed, regardless of the risk evaluation method used.
The review of defense-in-depth is typically qualitative and should address each of the elements with respect to the proposed change.
Evaluate the fire area for the impact of the VFDRs on fire protection defense-in-depth. Fire protection defense-in-depth is achieved when an adequate balance of each of the following elements is provided:
" preventing fires from starting;
" rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting fire damage; and,
- 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.
In general, the defense-in-depth requirement is satisfied if the proposed change does not result in a substantial imbalance among these elements. See Table PRA-RAI-09-1 below which contains additional defense-in-depth guidance.
Fire protection features and systems relied upon to ensure defense-in-depth should be clearly identified in the assessment (e.g., detection, suppression system).
For PWRs, a confirmation should be performed and positively stated that the remedy for a VFDR does not result in sole reliance on feed and bleed for the fire area under review.
Verify that defense-in-depth is maintained by assessing and documenting that the balance is preserved among prevention of core damage, prevention of containment failure, and mitigation of consequences. Regulatory Guide 1.174 provides guidance on maintaining the philosophy of nuclear safety defense-in-depth that is acceptable for NFPA 805 fire risk evaluations.
Each fire area shall be evaluated for the need to incorporate defense-in-depth enhancements to provide assurance that plant performance goals can be achieved and maintained.
Documentation of these defense-in-depth enhancements can be on a fire area basis and/or tied directly to a VFDR disposition, as appropriate.
Provide the results of the defense-in-depth review in a tabular format, such as that shown in the example in Table PRA-RAI-09-2. Defense-in-depth attributes shall be evaluated for applicability to NFPA 805, Section 4.2.3 or 4.2.4 (Ch. 3, as required).
LIC-1 2-0083 Enclosure Page 137 of 164
- If a fire protection feature or specific defense-in-depth attribute is required based on established deterministic criteria, licensing action or engineering equivalence evaluation then it should be designated required per section 4.2.3.
- If the fire PRA utilizes any of the fire protection features or a recovery action is deemed appropriate to improve the risk profile then these attributes or features are required per section 4.2.4.
It is essential that defense-in-depth attributes credited in the fire PRA model be reflected in the defense-in-depth evaluation. Of particular concern are detection, suppression and passive fire protection features. Where these attributes may not be credited in the deterministic evaluation, PRA may require a fire protection feature to improve the risk profile.
Documentation of the required defense-in-depth features are those fire protection features (programmatic controls, physical upgrades, recovery actions, etc.) that are intended to go above and beyond the existing requirement(s) with the purpose of bolstering derived weaknesses within the defense-in-depth elements to maintain an overall balance. These features or enhancements warrant inclusion in the monitoring program. Care should be taken to only include those fire protection features, enhancements, or recovery actions that are necessary to ensure that defense-in-depth is maintained.
Approach to Document Required Fire Protection Systems and Features The methodology that was used to document required fire protection systems and features, for defense-in-depth, is documented in section 5.8.2 of procedure EPM-DP-FP-003, "Fire Area Review", Revision 1, dated March 2011. The methodology is as follows:
"If the fire protection system or feature is required to demonstrate the acceptability of risk or defense-in-depth, as determined in step 5.6 [of procedure EPM-DP-FP-003], then it is required by Chapter 4 and is then subject to the applicable requirements of NFPA 805 Chapter 3.
Note: The defense-in-depth review performed as part of step 5.6 determines which systems and features require enhancement in order to demonstrate adequate balance of the three echelons of defense-in-depth. Only those fire protection systems and features that require enhancement are the ones considered "required" in the context of this review."
LIC-1 2-0083 Enclosure Page 138 of 164 Table PRA-RAI-09 Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Echelon 1: Prevent fires from starting
" Combustible Control Combustible and hot work controls are fundamental elements of
" Hot Work Control defense-in- depth and as such are always in place. The issue to be considered during the fire risk evaluations is whether this element needs to be strengthened to offset a weakness in another echelon thereby providing a reasonable balance. Considerations include:
- Creating a new transient free areas
- Modifying an existing transient free area The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine if additional controls should be added.
Review the remaining elements of defense-in-depth to ensure an over-reliance is not placed on programmatic activities for weaknesses on plant design.
Echelon 2: Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage
" Detection system Automatic suppression and detection may or may not exist in the fire
" Automatic fire suppression area of concern. The issue to be considered during the fire risk evaluation is whether installed suppression and or detection is f
Portable fire extinguishers provided required for defense-in-depth or whether suppression/detection for the area needs to be strengthened to offset a weakness in another echelon
" Hose stations and hydrants provided thereby providing a reasonable balance. Considerations include:
for the area a If a fire area contains both suppression and detection and
" Pre-Fire Plan fire fighting activities would be challenging, both detection and suppression may be required 0 If a fire area contains both suppression and detection and fire fighting activities would not be challenging, require detection and manual fire fighting (consider enhancing the pre-plans) 0 If a fire area contains detection and a recovery action is required, the detection system may be required.
0 If a fire area contains neither suppression nor detection and a recovery action is required, consider adding detection or suppression.
The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine the types of fires and reliance on suppression probability should be evaluated in the area to best determine options for this element of defense-in-depth.
LIC-12-0083 Enclosure Page 139 of 164
-,Table PRA-RAI-09 Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Echelon 3: Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed
" Walls, floors ceilings and structural If fires occur and they are not rapidly detected and promptly elements are rated or have been extinguished, then third echelon of defense-in-depth would be relied evaluated as adequate for the upon. The issue to be considered during the fire risk evaluation is hazard, whether existing separation is adequate or whether additional
" Penetrations in the fire area barrier measures (e.g., supplemental barriers, fire rated cable, or recovery are rated or have been evaluated as actions) are required offset a weakness in another echelon thereby adequate for the hazard. providing a reasonable balance. Considerations include:
" Supplemental barriers (e.g., ERFBS, N If the variance is never affected in the same fire scenario, cable tray covers, combustible liquid internal fire area separation may be adequate and no dikes/drains, etc.) additional reliance on recovery actions is necessary.
" Fire rated cable a If the variance is affected in the same fire scenario, internal
- Reactor coolant pump oil collection fire area separation recovery action may may not be adequate and reliance on a be necessary.
system (as applicable)
" Guidance provided to operations 0 If the consequence associated with the variances is high personancel provdetaig theoeqra s regardless of whether it is in the same scenario, a recovery personnel detailing the required action and / or reliance on supplemental barriers should be success path(s) including recovery considered.
safety actions to achieve nuclear performance criteria. . There are known modeling differences between a fire PRA and nuclear safety capability assessment due to different success criteria, end states, etc. Although a variance may be associated with a function that is not considered a significant contribution to core damage frequency, the variance may be considered important enough to the NSCA to retain as a recovery action.1 The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine the fires evaluated and the consequence in the area to best determine options for this element of defense-in-depth.
1 An example would be components in the NSCA associated with maintaining natural circulation at a pressurized water reactor that are not modeled explicitly in the fire PRA since they are not part of a core damage sequence.
For risk-informed, performance-based areas at FCS, any and all detection and suppression systems were credited for defense-in-depth (if not credited for another reason).
For risk-informed, performance-based areas at FCS, even if the fire scenarios showed acceptably low risk, the recovery actions were intentionally credited for defense-in-depth (if not credited for another reason) to accomplish the following objectives, as applicable based on the VFDRs:
" Actions to align AFW (FA 32, FA 34B-1, FA 36B)
- Action to maintain long-term DC electrical power for the available steam generator level indication `(FA 34B-1)
LIC-12-0083 Enclosure Page 140 of 164 The alternate shutdown recovery actions were maintained for the main control room (FA 42 for risk reduction) and the cable spreading room (FA 41 for DID).
For discussion of defense-in-depth actions, see the LAR Table G-1 and Transition Report Section 4.2.1.3.
Table PRA-RAI-09-2: Example Defense-in-Depth Impact Review for Fire Area Method of Providing DID T Assessment Basis/Justification Echelon 1: Prevent fires from starting Combustible Control is implemented in accordance with procedure X, "Control of Adequate *This element is adequate based on no perceived Combustible Materials". weakness of, or over-reliance on, another echelon of defense-in-depth.
in Hot Work Control is implemented accordance with procedure X, "Welding, Adequate Burning, and Grinding Activities" I I Echelon 2: Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage Fire detection system Adequate Automatic fire suppression Adequate
- Although fire fighting activities are not expected to be challenging, the detection and suppression Portable fire extinguishers Adequate systems are credited in the performance-based Hose stations and hydrants located in the analysis, and therefore, have also been credited area(s) Adequate for defense-in-depth Pre-Fire Plan Adequate Echelon 3: Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed Walls, floors ceilings and structural elements are rated or have been evaluated as Adequate adequate for the hazard.
Openings in the fire area barrier are rated or have been evaluated as adequate for the Adequate hazard.
Supplemental barriers (e.g., ERFBS, cable Adequate tray covers, etc.) Recovery actions credited to achieve Nuclear Not Required Safet Performance Criteria Fire rated cable A success path, including operational guidance if needed, remains free of fire Adequate damage.
Recovery action required in operational procedure to achieve nuclear safety Required performance criteria
LIC-12-0083 Enclosure Page 141 of 164 Probabilistic Risk Assessment RAI 10:
Please describe the methodology used to evaluate safety margins.
OPPD's Response to Probabilistic Risk Assessment RAI 10:
The methodology that was used to evaluate safety margins is described in procedure EPM-DP-RSD-001, "Fire Risk Evaluation", Revision 1, dated August 2010. The methodology is summarized below.
A review of the impact of the VFDRs on safetymargin shall be performed. An acceptable set of guidelines for making that assessment is summarized below. Other equivalent acceptance guidelines may also be used.
" Codes and standards or their alternatives accepted for use by the NRC are met, and
" Safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provide sufficient margin to account for analysis and data uncertainty.
The requirements related to safety margins for the fire risk evaluation are grouped into four categories. These categories are:
Fire Modeling Review the quantitative margin between the parameters describing the MEFS and the LFS per section 6.2.4 [of procedure EPM-DP-RSD-001], "Initial Evaluation - Fire Modeling."
Plant System Performance Review the methods, input parameters, and acceptance criteria used in the fire PRA analyses against that used for the plant design basis events and verify that the safety margin inherent in the analyses for the plant design basis events have been preserved in the analysis for the fire events.
The following questions should be addressed when evaluating the plant system performance:
" Were input parameters for plant performance analyses altered from those used for plant design basis events such that the safety margin was lessened? (Ex: pump performance curves, heat transfer coefficients, etc.)
" Would the NRC find the codes and standards used to determine plant system performance acceptable?
PRA Logic Model Evaluate the quantification for fire related CDF/LERF based on the plant PRA model.
Evaluate the plant PRA model modifications required for the fire PRA including altered basic event failure probabilities, added basic events, and changes to the logic structure.
LIC-1 2-0083 Enclosure Page 142 of 164
" Evaluate the changes against the methods and criteria for the overall internal events PRA model development for consistency, or confirmation of bounding treatment, to confirm that the safety margin is preserved.
" Confirm that the quantified model is sufficient to treat the fire induced core damage sequences. If the analysis was performed using the plant PRA model with no modifications other than normalizing the initiating event frequency to 1.0, and setting other non-credited events to 'TRUE' or 1.0, then the safety margin is preserved and no further assessment for safety margin is necessary for this category.
" Assess the fire modeling margin by crediting the NUREG/CR-6850 methods used.
Miscellaneous Address any other analyses that may have been performed that have not been addressed by the prior categories.
Example of a Typical Safety Margin Review as Contained in a Fire Risk Evaluation for a Fire Area with One or More VFDRs In accordance with NEI 04-02, 5.3.5.3 guidance, the maintenance of adequate safety margin is assessed by the consideration categories of analyses utilized by this fire risk evaluation.
Safety margins are considered to be maintained if:
- Codes and standards or their alternatives accepted for use by the NRC are met, and
- Safety analyses acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provides sufficient margin to account for analysis and data uncertainty.
The following summarizes the bases for ensuring the maintenance of safety margins:
- The risk-informed, performance based processes utilized are based upon NFPA 805, 2001 edition, endorsed by the NRC in 10 CFR 50.48(c).
- The fire risk evaluation process is in accordance with NEI 04-02, Revision 2, which is endorsed by the NRC in Regulatory Guide 1.205, Revision 1.
- The fire PRA is developed in accordance with NUREG/CR-6850.
" The VFDRs are assessed using the FPRA model documented in References 4 and 6-13 of FC07883. The FCS FPRA has received a formal industry peer review against the section IV requirements of ASME/ANS RA-Sa-2009 and in accordance with the peer review guidelines of NEI 07-12. This peer review was conducted September 2 7 th through October 1 st, 2010 by a diverse group of industry experts, collectively representing all skill sets required to critically review a FPRA. The review covered all aspects of the FCS FPRA model and the administrative processes used to maintain and update the model. The review generated specific, recommendations for model,
LIC-12-0083 Enclosure Page 143 of 164 documentation, and process improvements, and these recommendations are documented in the form of facts and observations (F&Os) in the peer review report. All F&Os have been addressed as documented in LTR-RAM-II-10-046. The peer review report documents that the FCS FPRA model meets the requirements of capability category II or III for all requirements, with the exception of those identified and dispositioned in FC07883.
" The "combined analysis approach" is used during transition (NEI 04-02, Section 5.3.4.3); therefore, MEFS/LFS is not analyzed separately from the Fire PRA results.
" Peer reviews, gap assessments, and self-assessments were performed of the FCS PRA as discussed in the response to Probabilistic Risk Assessment RAI 14.
- Fire protection systems and features determined to be required by NFPA 805 Chapter 4 have been confirmed to meet the requirements of NFPA 805 Chapter 3 and their associated referenced codes and listings, or provided with acceptable alternatives using processes accepted for use by the NRC (i.e., FAQ 06-0008, FAQ 06-0004, FAQ 07-0033).
- Fire modeling performed in support of the change evaluations is performed using conservative methods and input parameters that are based upon NUREG/CR-6850.
These conservative methods and input parameters are documented in FC07883, Fire Risk Assessment of FCS Variances from Deterministic Requirements of NFPA 805.
Probabilistic Risk Assessment RAI 11:
While Attachment W of the LAR provides the ACDF and ALERF for the VFDRs for each of the fire areas, the LAR does not describe either generically or specifically how ACDF and ALERF were calculated. Please describe the method(s) used to determine the changes in risk reported in the Tables in Appendix W. Include in the description discussion of specific PRA model additions or modifications needed to determine the changes.
OPPD's Response to Probabilistic Risk Assessment RAI 11:
The LAR Table B-3 identifies all Variances from the Deterministic Requirements (VFDRs) to be assessed with the Fire Risk Evaluation (FRE) process. An overview of the FRE process is provided in Table 4.5.2.2 and supplemented by the following paragraphs.
First, each VFDR in a given fire compartment was assessed for risk relevance. Since the Nuclear Safety Capability Assessment (NSCA) requirements of NFPA 805 Chapter 4 are often different from the purposes and modeling approaches of a fire PRA, some VFDRs may not be applicable to or able to be assessed by the fire PRA. For example, the NSCA might consider functional failure of the pressurizer heaters. However, since the fire PRA does not model functional failure of the pressurizer heaters, which are not modeled to mitigate fire-induced initiating events, the risk impact of the VFDR is assumed negligible. All VFDR failures determined relevant to the fire PRA were identified for risk assessment. If the compartment did not contain any risk-relevant VFDRs, then this was documented and no further evaluation was required.
LIC-1 2-0083 Enclosure Page 144 of 164 A fire PRA modeling strategy for each risk-relevant VFDR was then developed. This generally consisted of identifying basic events that represent the component failures associated with each VFDR. If a risk-relevant VFDR could not readily be modeled with existing basic events, then new logic would be added to the fire PRA model. The fire PRA database was then queried to identify fire scenarios within the compartment that can induce VFDR failures. If no fire scenarios were identified to be capable of inducing VFDR failures within the compartment, then this was documented and no further evaluation was required.
The fire-induced CDF and LERF for the as-built plant were then quantified for each identified scenario. Each scenario was quantified by setting all basic events that can be credibly induced by the scenario to TRUE (or to a conditional probability where supported by circuit failure likelihood analysis), includinq those basic events that model risk-relevant VFDR failures and can be credibly induced by the scenario.
The fire-induced CDF and LERF associated with the compliant case were then quantified for each identified scenario. The compliant case represents a hypothetical plant configuration in which some physical modification (e.g., re-routing cables, installing cable tray fire barriers, etc.)
were to be implemented to eliminate the VFDR. This quantification was performed by setting all basic events that can be credibly induced by the scenario to TRUE (or to a conditional probability where supported by circuit failure likelihood analysis), excludinq those basic events that model risk-relevant VFDR failures. The probabilities of basic events associated with the VFDRs were left at their base values. The ACDF and ALERF between the as-built and compliant cases were then calculated.
This calculation was performed at the fire compartment level. That is, the as-built and compliant fire risk each included the contribution of all fire scenarios within the compartment of interest that could credibly induce risk-relevant VFDR failures. The cutsets for the as-built and compliant cases were reviewed to identify the primary factors contributing to the ACDF and ALERF for each VFDR-relevant scenario. Finally, the ACDF and ALERF were compared against the acceptance criteria of Regulatory Guide 1.174.
As an alternative to the detailed approach described above, a bounding approach was implemented for some compartments when the total fire risk for the compartment was within the acceptance criteria of Regulatory Guide 1.174. In this approach, it is acknowledged that performing physical plant modifications to prevent the VFDR failures from occurring would result in some un-quantified reduction in compartment CDF and LERF. The ACDF and ALERF associated with these modifications would be therefore, by definition, within the acceptance criteria of Regulatory Guide 1.174. This bounding approach was implemented for some fire compartments involving a broad spectrum of possible fire impacts. This simplification was made because it can be difficult to resolve (i.e., "see") individual failures within such compartments, since the relatively severe damage states overwhelm or mask individual failures associated with VFDRs. Despite being a simplification, this bounding approach ensured that the ACDF and ALERF were within the acceptance criteria.
LIC-1 2-0083 Enclosure Page 145 of 164 Probabilistic Risk Assessment RAI 12:
Please describe any PRA methods used for the plant modifications identified in the LAR Attachment S, and for other plant modifications that were not included in the PRA at the time of the FPRA peer review. Include a discussion of any basic events or models added to the PRA to address these modifications.
OPPD's Response to Probabilistic Risk Assessment RAI 12:
All fire PRA methods and plant modifications supporting the LAR were in place at the time of the fire PRA peer review September 27 through October 1, 2010. New fire PRA methods and plant modifications have not been added to the fire PRA model supporting the LAR since the peer review.
The following bulleted paragraphs discuss how each type of plant or procedure modification is reflected in the fire PRA model.
- Install electrical isolation for various breaker trip control circuits to ensure the breakers remain functional to trip on demand for automatic load shed, overcurrent, and manual control from the control room. These modifications help assure offsite power and diesel generator availability during fire scenarios that affect the relevant breaker trip control circuits. The fire PRA models the benefit of these modifications by not assuming failure of offsite power and/or the diesel generator(s) as a consequence of fire impacting the relevant breaker trip control circuits.
- Install interlock to prevent 4 KV bus 1A3 failure via spurious alignment to both offsite power and the diesel generator. This modification also protects offsite power to redundant 4 KV bus 1A4 for certain scenarios. The fire PRA models the benefit of this modification by not assuming failure of 4 KV bus 1A3 for fire scenarios that may, with the current plant design, spuriously align both offsite power and the diesel generator to 1A3 simultaneously.
- Provide additional electrical isolation between the station batteries and their respective direct current distribution panels. For a battery room fire, this modification ensures availability of the distribution panel to be powered from its battery charger. The fire PRA models the benefit of this modification by not assuming de-energization of the distribution panel for its respective batter room fire scenario.
" Install shielding in the switchgear rooms to limit the extent of fire damage to failure of the high energy arc faulted bus and any downstream component powered by the faulted bus. The fire PRA models the benefit of this modification by not assuming consequential failure of components beyond the faulted bus itself.
LIC-1 2-0083 Enclosure Page 146 of 164 Modify the administrative control procedures to disallow hotwork in FC32 while the plant is at-power, require a continuous fire watch anytime transient combustibles are stored on the roof of Room 18, and require a continuous fire watch anytime the five pound transient combustible limit is exceeded in the cable spreading room. The current fire PRA models the benefit of these administrative controls via a 0.01 reduction factor (representing a combination of failure to comply with the procedure and failure of the continuous fire watch to suppress an incipient fire). This approach is recognized as a deviation from NUREG/CR-6850 and will be addressed via sensitivity studies in response to PRA RAI 07. (OPPD will provide the responses to PRA RAI 07 on September 28, 2012. [AR 52508])
- Revise AOP-06 to direct operators to abandon the control room when either a fire has caused inhabitability or a significant loss of plant control. The current procedure directs abandonment upon receipt of two valid smoke alarms or halon discharge (either inside the main control board or in the cable spreading room). The fire PRA models the benefit of this procedure revision by not assuming potential pre-mature abandonment based on receipt of two smoke alarms or halon discharge. The fire PRA models abandonment upon loss of habitability. The fire PRA currently does not explicitly model abandonment upon significant loss of control, and this is generally conservative because "recovery" (from the alternate shutdown panel) of the high CCDP and CLERP caused by the significant loss of control is not credited.
Probabilistic Risk Assessment RAI 13:
There are several implementation items in LAR Attachment S that have not been completed but which have been credited directly or indirectly in the change-in-risk estimates provided in Attachment W. When an implementation item has been included in the PRA but not yet implemented, the models and Values used in the PRA are necessarily estimates based on current plans. The as-built facility after implementation is completed may be different than the plans. Please add an implementation item that, upon completion of all PRA credited modifications (including procedural modifications),
verify the change-in-risk estimate reported in the LAR. This implementation item should include your plan of action should the as-built change-in-risk exceed the estimates reported in the LAR.
OPPD's Response to Probabilistic Risk Assessment RAI 13:
As the planned plant modifications, including procedure revisions, are planned and implemented, OPPD will verify that their risk reducing benefit remains consistent with, or bounded by, the estimates provided in the LAR. As part of implementation, OPPD will submit to the NRC, a summary of the issue and its proposed resolution should any nonconservative inconsistencies with the credit taken for plant modifications in the LAR be discovered. The updated LAR Attachment S, including this implementation item will be reflected in the NFPA 805 transition LAR supplement. [AR 48249]
LIC-1 2-0083 Enclosure Page 147 of 164 Probabilistic Risk Assessment RAI 14:
It is not clear that the quality of the IEPRA against the requirements of RG 1.200, Rev. 2 is established. Based on the LAR and clarifications in the LAR supplement dated December 22, 2011, the most current full peer review was performed in 1999 well prior to when the ASME PRA standards were first issued. The only full gap assessment against the PRA standards was performed in 2003 against a draft version of RA-Sa-2003. There were a number of changes in the internal events SRs between the draft RA-Sa-2003 and the RA-Sb-2005, and a few minor changes between RA-Sb-2005 and RA-Sa-2009, the current version of the standard (as endorsed by RG 1.200). A handful of focused scope reviews and self-assessments were performed after the 2003 gap assessment. Notable was a 2006 focused scope self-assessment in which 41 SRs from a ASME RA-Sb-2005 draft standard applying to Mitigating Systems Performance Indicators (MSPI) were assessed and for which it is stated that "the remaining SRs were also reviewed but to a lesser degree of scrutiny." The LAR supplement provides a road map of the different PRA reviews performed since the 1999 peer review and attempts to make the case that review against the RA-Sa-2009 version of the PRA standards and RG 1.200, Rev. 1 is established. However, the differences between all the SRs in the draft ASME RA-Sa-2003 standard used in the 2003 gap assessment and the RA-Sa-2009 version have not been identified and dispositioned as being addressed in the intervening assessments. Please explain how the quality of the IEPRA meets the requirements of RG 1.200 given the changes between the ASME standard which your PRA was peer reviewed against and the current version of the standard as endorsed by RG 1.200, Rev. 2.
OPPD's Response to Probabilistic Risk Assessment RAI 14:
Based on a review of the IEPRA peer review / self-assessment history it was confirmed that the most current official full scope peer review was performed in 1999, where Fort Calhoun Station (FCS) was a pilot for the Westinghouse and Combustion Engineering Owners Group (CEOG) peer review process, which was equivalent to Nuclear Energy Institute (NEI) 00-02.
In February 2006, an independent gap assessment was performed under the auspices of RG 1.200, Appendix B. Although the purpose of this independent gap assessment was to review the 41 Supporting Requirements (SRs) of a draft of PRA Standard, American Society of Mechanical Engineers (ASME) RA-Sb-2005 that apply to Mitigating System Performance Index (MSPI), a full review of the internal events SRs was also completed in this gap assessment. There is no indication that any of the SRs were reviewed to "a lesser degree of scrutiny." The Enclosure 1 of the LAR supplement dated December 22, 2011, (LIC-1 1-0136),
has been revised and this statement has been removed.
The February 2006 review was performed generally in accordance with the guidance in Appendix B of RG 1.200, considering regulatory interpretations of the ASME Standard as noted in Appendix A of RG 1.200. The differences between this review and the Peer Review requirements documented in Section 6 of ASME RA-Sb-2005 (the version of the ASME PRA standard available at the time) are minimal and are judged not to impact the technical adequacy of the review. Documentation of the review team details (e.g., reviewer resume,
LIC-12-0083 Enclosure Page 148 of 164 experience / knowledge base etc...) was not included and a single reviewer was utilized to review multiple PRA elements. However, this review followed the process documented in the original version of NEI 05-04 and was performed by an independent party who had substantial peer review experience and was highly experienced and knowledgeable in all PRA areas being reviewed. All relevant documentation of the PRA changes was transmitted prior to the review and adequate time was allotted for the review of each change.
While the primary focus of this review was to evaluate MSPI, the final evaluation contained a thorough review of all individual SRs from ASME RA-Sb-2005, a detailed review and the assigned capability category are documented for each SR. The February 2006 review generated a total of 16 B Level Facts and Observations (F&Os) and 9 C Level F&Os. Based on the quality of the review performed in February 2006, a review of the changes between the draft ASME RA Sa 2003, standard and the ASME RA-Sb-2005 standard is not necessary.
In December 2008, a focused scope peer review was performed specifically on the Internal Flooding SRs of AMSE RA-Sb-2005. This peer review was triggered by an upgrade of the FCS internal flooding PRA model.
In order to demonstrate compliance with RG 1.200, Rev. 2, the Enclosure 1 of the LAR supplement dated December 22, 2011, (LIC-1 1-0136), has been revised and now includes the following:
a) The resolution of all A or B Level F&Os generated throughout the history of the FCS PRA model.
b) The existing ASME RA-Sb-2005 SR grades were mapped to the appropriate ASME/American Nuclear Society (ANS) RA-Sa-2009 SRs and any SR wording changes were reviewed to verify that the current grade was appropriate for the revised SR.
c) The clarifications and qualifications documented in RG 1.200, Rev. 2 were reviewed against the mapped ASME/ANS RA-Sa-2009 SR grades to verify that the current grade met the expectations of RG 1.200, Rev. 2.
d) A self-assessment was performed for all SRs with a grade less than CC-II; all SRs that had not been previously assigned a grade (either new ASME/ANS RA-Sa-2009 SRs or gaps in prior peer reviews) and all SRs where gaps were identified in items b) and c). This assessment reviewed all SRs and provided the basis for each grade (either a reference to the review that assigned the grade, or the disposition provided as part of the self-assessment). In addition, this assessment also considered the impact of wording changes between ASME RA-Sb-2005 and ASME/ANS RA-Sa-2009 as well the clarifications and qualifications in Appendix A of RG 1.200, Rev. 2.
e) All open / unresolved F&Os and all Internal Events and Internal Flooding SRs with a grade less than CC-Il are dispositioned with respect to the FPRA and the NFPA 805 application.
The results of the activities listed above indicate that the quality of the Revision 11 IEPRA meets the requirements of RG 1.200 for the NFPA 805 risk informed application.
LIC-1 2-0083 Enclosure Page 149 of 164 Probabilistic Risk Assessment RAI 15:
Please clarify the following dispositions to IEPRA peer review findings identified in Attachment U of the LAR, Table 6-2 of LTR-RAM-11-1 0-046, Revision 1, and from the 2007 PRA Self-Assessment (SA-07-48) that appear to have the potential to noticeably impact the fire PRA results but which do not seem to be fully resolved.
- a. The disposition to SR IE-A8 on page 10 of the 2007 PRA Self-Assessment Report (SA-07-48) acknowledges that no interviews were specifically performed to meet this SR. (This SR is also identified in Attachment U, Table U-2 of the LAR). Please summarize efforts that contribute to ensuring comprehensive identification of initiating events, describe how sensitive this process is to possible incompleteness in the internal initiating event models, and, based on these summaries, explain why interviews are not needed to provide confidence that the IlEs [internal events] have been appropriately characterized.
OPPD's Response to PRA RAI 15 a.:
ASME/ANS RA-Sa-2009, SR IE-A8, Capability Category II, states: "INTERVIEW plant personnel (e.g., operations, maintenance, engineering, safety analysis) to determine if potential initiating events have been overlooked." Early in the development of the PRA, while the list of initiating events was still in draft form, two engineers with extensive FCS plant engineering experience joined the PRA group.
One of the individuals held an active Senior Reactor Operator license, and his previous positions included Reactor Engineer, operations and engineering training supervisor, and emergency operating procedure coordinator. The other individual had extensive experience in plant engineering, maintenance, and engineering programs such as in-service testing and in-service inspection. Both of these individuals continue to support the FCS PRA.
During preparation of the Individual Plant Examination (IPE) and Individual Plant Examination of External Events (IPEEE) PRA models, results and insights were presented regularly to the PRA Oversight Committee, a cross-disciplinary committee representing areas such as operations, maintenance, system engineering, and design engineering. This interaction provided ongoing validation of the PRA model and results. Deliberations of the PRA Oversight Committee were documented in minutes, and opportunities for improving the PRA model were implemented using the PRA configuration control process.
The completeness of the initiating events list continued to be challenged after the initial models were complete. Over time, additional members were added to the PRA staff, each of whom had years of FCS plant experience. FCS positions previously held by the staff included Shift Technical Advisor, plant engineering, chemistry, and radiological engineering. A member of the PRA staff added in 2006 had previously been a Shift Manager and Work Week Manager at FCS. FCS continues to maintain an on-site PRA group that is intimately familiar with the plant design and operation.
LIC-1 2-0083 Enclosure Page 150 of 164 The PRA group continues to work closely with representatives from operations, work week management, and system engineering with respect to Maintenance Rule risk assessments and Significance Determination Process issues. This relationship provides ongoing challenge and validation of the PRA model and assumptions.
FCS and industry operating experience, as well as plant modifications, are routinely reviewed by the PRA staff. This information is compared with the internal events PRA, to determine if changes to the initiating events list are warranted. As additional assurance that potential initiating events had not been overlooked, formal interviews were conducted with licensed operators, Shift Technical Advisors, and system engineers in March of 2012. No new insights were gained regarding initiating events for the internal events PRA; consequentially no new internal events initiators were identified.
These interviews were performed and documented consistent with the expectations of ASME/ANS RA-Sa-2009.
To ensure that no new initiating events are overlooked in the future, the process of conducting interviews with plant personnel has been added to the initiating event frequency data update process.
- b. LAR Attachment U, Table U-2 indicates that SR SY-A4 is met at only CC-1.
Please describe what efforts were made during the systems analysis to correctly reflect the as-built, as-operated plant. Include in this explanation:
identification of walkdowns or interviews performed in support of the PRA, the scope and extent of those efforts, and who performed them (e.g., PRA staff, engineering, plant operations, etc.).
OPPD's Response to PRA RAI 15 b.:
ASME/ANS RA-Sa-2009, SR SY-A4, Capability Category Il/111, states:
"PERFORM plant walkdowns and interviews with knowledgeable plant personnel (e.g., engineering, plant operations, etc.) to confirm that the systems analysis correctly reflects the as-built, as-operated plant."
The FCS IPE submittal includes the following statements:
" "The IPE approach included system, walkdown, procedure, and drawing reviews as well as discussions with Operations, Engineering, and other plant personnel."
" "Systems notebooks were prepared after plant walkdowns and review of drawings, system descriptions ... "
LIC-1 2-0083 Enclosure Page 151 of 164
- "For the development of the computer [PRA] model, system notebooks, and various analyses performed in the IPE, plant information was utilized to confirm that the evaluation represented as-built, as-operated conditions.
This was done by using the following: extensive walkdowns ... and information from knowledgeable plant staff."
0 "The behavior of containment was determined by various walkdowns, training ... "
Since there were no formal walkdown or interview documentation requirements at the time, documentation of those walkdowns and interviews is limited. They were performed by OPPD PRA staff and consultants who worked on the FCS PRA during its early phases.
Early in the development of the PRA, while the system models were still in draft form, two engineers with extensive FCS plant engineering experience joined the PRA group. One of the individuals held an active Senior Reactor Operator license, and his previous positions included Reactor Engineer, operations and engineering training supervisor, and emergency operating procedure coordinator. The experience of these individuals had a major impact upon development of the PRA system models, and they were responsible for many changes and improvements.
Both of these individuals continue to support the FCS PRA.
During preparation of the IPE and IPEEE PRA model, results and insights were presented regularly to the PRA Oversight Committee, a cross-disciplinary committee representing areas such as operations, maintenance, system engineering, and design engineering. This interaction provided ongoing validation of the PRA model and results.
The quality of the system models continued to be improved after the initial models were complete. Over time, additional members were added to the PRA staff, each of whom had years of FCS plant experience. FCS positions previously held by the staff included Shift Technical Advisor, plant engineering, chemistry, and radiological engineering. A member of the PRA staff added in 2006 had previously been a Shift Manager and Work Week Manager at FCS. FCS continues to maintain an on-site PRA group that is intimately familiar with the plant design and operation.
The system engineering qualification card for each system includes a sign-off for a PRA engineer, indicating that the system engineer understands the PRA insights and perspective related to that system. The PRA engineer signs the card after a one-on-one interview with the system engineer. In preparation for the interview, the system engineer reviews the applicable section of the PRA Summary Notebook. This process provides effective exchange of information between the PRA group and the system engineering department with respect to the as-built, as-operated plant.
LIC-1 2-0083 Enclosure Page 152 of 164 In support of the NFPA 805 fire RAI response effort, a structured walkdown of all PRA systems was performed. The purpose of this walkdown was to confirm that plant systems and their associated dependencies were properly modeled and that the PRA system notebooks describing the system models were accurate.
Walkdowns were limited to accessible equipment. Samples of accessible equipment located in the Intake Structure, Yard Area, Auxiliary Building, Turbine Building, and Containment Building were reviewed against the PRA.
Documentation of the walkdown methodology and observations made during the walkdowns is being assembled into a system walkdown notebook supporting the PRA model. Documentation will also include photographs of the equipment and plant areas taken during the walkdowns. No discrepancies between the PRA and the current plant design, plant procedures, and plant performance data were identified.
The systems analysis procedure has also been revised to include direction to perform walkdowns or interviews for plant changes that could cause PRA model or system notebook changes.
In summary, the above actions provide assurance that the systems analysis portion of the FCS PRA correctly reflects the as-built, as-operated plant.
C. F&O SY-01-GA identified in LAR Attachment U, Table U-1 (originating from the 2006 gap self-assessment) finds that component boundaries used in the PRA models were not matched to component boundary definitions of the component failure data. Please explain what impact inconsistent boundaries definitions might have on the FPRA. Please clarify whether a review of component boundary definitions has been completed and, if so, provide a sensitivity analysis to determine the impact to the FPRA CDF and LERF, and ACDF and ALERF. Also, please justify that any inconsistent boundaries (e.g.,
placement of isolation breakers) does not mask identification of dependencies or insights important to the FPRA.
OPPD's Response to PRA RAI 15 c.:
The Revision 11 PRA does not include updated component boundary definition failure modes. The component boundary definitions have been updated in the Revision 12 PRA, and will therefore be included in the final FPRA that will be implemented as part of the NFPA 805 application.
In order to implement the updated component boundary definitions model, a consistency check was performed for the components included in the PRA model.
The check was performed to determine whether the component boundaries for the basic events included in the PRA model were consistent with the component boundaries used in the data analysis. The consistency check also revealed that certain system sub-components that were included within the component boundary were also explicitly modeled. This resulted in double-counting of the affected component failure mode. To address this concern, sub-components that were
LIC-12-0083 Enclosure Page 153 of 164 included in the component boundary definition and explicitly modeled in the PRA were removed. For example, time delay relay for a pump was removed from the PRA model because this sub-component is part of the pump control circuit and is included in the boundary for a pump.
A sensitivity study was performed that included updated fault tree modeling for those PRA components that either had (a) subcomponents explicitly modeled that were included in the component boundary definition (i.e., double counting sub-component failures) or (b) sub-components failures that impacted other components and have been shown to not be explicitly modeled (i.e., the model was missing sub-component failures). The results of this sensitivity study are shown below in Table 1.
Table 1: Results of FPRA Sensitivity Study for PRA RAI 15c:
Incorporation of Corrected Component Boundary Definitions into the FPRA Base Fire Sensitivity Delta PRA* Study VFDR ACDF for NFPA 805 Transition (/yr) 5.72E-06 8.46E-06** 5.95E-07 VFDR ALERF for NFPA 805 Transition (/yr) 6.67E-07 5.82E-07** -8.49E-08 Total CDF (Internal Events, Internal Flood 6.01 E-05 6.09E-05 7.95E-07
& Fire) (/yr) 6.01E-05 6.09E-05 7.95E-07 Total LERF (internal', flood, fire) (/yr) 4.82E-06 4.76E-06 -6.49E-08
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
Sensitivity study case for Variance from Deterministic Requirement (VFDR) ACDF and ALERF is conservatively assessed by adding the change in the FPRA CDF and LERF, between the base and sensitivity case, to the VFDR ACDF and ALERF.
This sensitivity study shows that the revised component boundary definitions increased the total CDF by 7.95E-07/yr and decreased the total Large Early Release Frequency (LERF) by 6.49E-08/yr. The reason LERF is reduced for this sensitivity study is primarily due to the correction of double-counting certain failure modes for the containment isolation valves controlling instrument air to the containment loop. The resulting reduction to the containment isolation failure probability offsets the slight increase in CDF shown for this sensitivity study. In reviewing the cutsets for this sensitivity study it was noted that the changes did not result in any major change or new significant sequences, thus it is concluded that this issue is not masking any important insights.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, when the revised component boundary definitions are applied to the fire PRA supporting the LAR.
- d. F&O QU-02-GA identified in LAR Attachment U, Table U-1 (originating from the Internal Events GAP Review) found that state-of-knowledge correlations between event probabilities were not performed for the IEPRA. This SR remains unresolved. Please explain how state-of-knowledge correlations
LIC-1 2-0083 Enclosure Page 154 of 164 were addressed for the FPRA, both those in the internal events models' events, and those added by the fire events. If state-of-knowledge correlations were not addressed, please provide an estimate of the impact of not considering state-of-knowledge correlations including impact to the FPRA CDF and LERF, and ACDF and ALERF.
State-of-Knowledge Correlation (SOKC) uncertainty exists when two or more events from the same data set (i.e., use the same type code) exist in the same cutset. For the case where the cutset consists of two basic events (i.e., A and B),
the quantification software product (CAFTA) quantifies the cutset value as the product of the point estimate value (or mean value) of event A and the point estimate value (or mean value) of event B. In reality, A and B are not independent as they result from sampling the same distribution for correlated data and the sample means of A and B cannot be directly multiplied to estimate the expected value of the combined event. The correct representation of expected value of the combined event is the product of the mean values for events A and B, plus the variance of the distribution. Not including this explicitly in a PRA model results in an underestimation of CDF and LERF cutsets that have more than one event in the same cutset with the same type code. FCS has not explicitly included this in their base IEPRA or in their FPRA.
To estimate the impact of SOKC uncertainty on the FCS PRA model, combinations of events with the same type code were identified, and a factor was applied to them that accounts for the impact of SOKC uncertainty. This factor effectively increases the contribution of the variance of the distribution specific to that combination (i.e., a factor was developed for each type code that represents the addition of the distribution's variance). Note that this sensitivity study was not necessary for the IEPRA as the reported risk metrics are mean values calculated using the UNCERT program, which inherently includes the impact of SOKC uncertainty. The results of this sensitivity study are shown below in Table 2.
Table 2: Results of FPRA Sensitivity Study for PRA RAI 15d:
Impact of SOKC Uncertainty on the FPRA Base Fire Sensitivity Delta PRA* Study VFDR ACDF for NFPA 805 Transition (/yr) 5.72E-06 6.84E-06** 1.12E-06 VFDR ALERF for NFPA 805 Transition (/yr) 6.67E-07 7.04E-07** 3.71 E-08 Total CDF (Internal Events, Internal Flood 6.01 E-05 6.12E-05 1.12E-06
& Fire) (/yr) _.01E-05 _.12E-05 1.12E-06 Total LERF (internal, flood, fire) (/yr) 4.82E-06 4.86E-06 3.71 E-08
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
LIC-12-0083 Enclosure Page 155 of 164 The result of this sensitivity study showed that the incorporation of SOKC uncertainty into the FPRA increased total CDF and LERF by 1.12E-06/yr and 3.71 E-08/yr, respectively, as compared to the base fire PRA supporting the LAR.
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, when the SOKC uncertainty is applied to the fire PRA supporting the LAR.
- e. LAR, Table V1, indicates that SRs LE-C9, LE-C11, and LE-C13 are not met.
The 2007 PRA Self-Assessment points out that Level 2 modeling is conservative because no credit was taken in the Containment Event Tree (CET) for system operation or human actions under adverse environments or after containment failure. The 2007 self-assessment also states that no credit was taken for "scrubbing." Please clarify what credit was taken in the CET supporting the fire PRA. Justify that the lack of modeling detail in the CET does not mask important insights or dependencies for the FPRA.
OPPD's Response to PRA RAI 15 e.:
The FCS LERF methodology is based on a Pressurized Water Reactor Owners Group (PWROG) enhancement to NUREG/CR-6595, to include increased realism in the LERF calculation with respect to the treatment of induced steam generator tube ruptures, hydrogen burn and direct containment heating. This model represented a simplification of the full Level 2 model used for the FCS IPE with no significant loss of risk insights.
Each of the three SRs indicated (LE-C9, LE-C1 1 and LE-C13) will be addressed individually with respect to the FPRA.
LE-C9:
With respect to Supporting Requirement LE-C9, credit was taken for operation of Containment Sprays (CS) and Containment Fan Coolers (CFCs) during a severe accident. Equipment survivability assessments performed for the CFCs determined that the CFC units were likely to survive the early temperature and pressure environments associated with a severe accident. The CFCs and CSs are not credited for LERF as they are judged to be ineffective for mitigating any rapid early containment failure challenges (e.g., hydrogen burns, Direct Containment Heating (DCH)) and as such CS and CFC operation following a severe accident has no impact on LERF. The primary impact of continued operation of CS and CFCs is to protect the plant against late containment over-pressure failure events.
Two operator actions were credited in the FCS LERF model. Operator actions credited are (1) to reduce RCS pressure following core melt to avert a Thermally-Induced Steam Generator Tube Rupture (TI-SGTR) (this action credits the availability of the Power Operated relief Valves (PORVs) early in the severe accident) and (2) an operator error of commission to bump the Reactor Coolant Pumps (RCPs) following core melt (this action assumes operability of at least one
LIC-1 2-0083 Enclosure Page 156 of 164 RCP). This action increases the likelihood of TI-SGTR; although cautions are provided in the Severe, Accident Management Guidelines (SAMGs) to prevent operators from performing this action, it is included as a conservatism to bound the uncertainty associated with crediting SAMG actions. These operator actions were assigned bounding probabilities.
There is considerable uncertainty in the treatment of equipment survivability and operator actions in a post severe accident environment. Overall, the treatment of equipment survivability and operator actions in the LERF model may introduce minor conservatisms in the LERF assessment, but will not distort assessments for LERF. From the perspective of the FPRA, taking credit for equipment operating in a challenging environment would introduce additional equipment that could potentially be failed by the fire, thus increasing the importance of a given fire sequence. Similarly, crediting operator actions that require the use of equipment in the plant (e.g., PORVs to depressurize the Reactor Coolant System (RCS)) would also introduce additional equipment that could potentially be failed by the fire, thus increasing the importance of a given fire sequence. Therefore, it is judged that this treatment does not mask important insights or dependencies for the FPRA.
LE-C11:
FCS does not take credit for equipment operation or operator actions following a containment failure. While it is possible that operator actions to establish fission product scrubbing (e.g., sprays) or other beneficial effects could reduce fission products, it is assumed that these actions would not be effective to the point where a containment failure (with an effective break diameter greater than 2 inches) could be considered to have a small release. These model assumptions have a minimal impact on LERF calculation and are considered to be realistic; therefore, they are judged not skew the FPRA LERF results.
LE-C13:
Although a Containment Event Tree (CET) branch point for scrubbing is included for bypass sequences, this branch-point is set to zero such that no scrubbing is credited. This assumption primarily affects Steam Generator Tube Rupture (SGTR) sequences where secondary side heat removal is established in the affected Steam Generator (SG). In this case the release would be bubbled through a full SG prior to release out the Main Steam Safety Valve (MSSV). Because there are no SGTR initiating events included in the FPRA, the only sequence that would be categorized as a Level 1 SGTR is a pressure induced SGTR (for sequences with a high pressure differential across the SG tubes, e.g., Main Steam Line Break (MSLB) or Anticipated Transient without Scram (ATWS)). Therefore, the assumptions regarding credit for scrubbing of bypass sequences are considered to have a negligible impact on the FPRA results.
LIC-1 2-0083 Enclosure Page 157 of 164
- f. F&O DA-01-GA is identified in the 2007 PRA Self-Assessment report (SA 48) as Not Met. This F&O is not identified in LAR Attachment U, Table U-2 nor is it identified or dis positioned in the LAR, Table VI. Please explain how coincident maintenance on redundant equipment (both intrasystem as well as intersystem) was addressed for the IEPRA and how it impacted the FPRA.
If it was not addressed or only partly addressed, please provide the quantitative impact to the FPRA CDF and LERF, and ACDF and ALERF.
OPPD's Response to Probabilistic Risk Assessment RAI 15 f.:
Corrective and preventive maintenance of plant equipment is performed during at-power operations within the time periods specified by the Technical Specifications.
A plant procedure is in place to guard against the voluntary removal of redundant plant equipment from service at the same time. This procedure, which applies to most systems important to PRA, instructs operating personnel to verify the operability of redundant equipment prior to removing selected components from service. Based on this procedure, and good operating practices, coincident maintenance of redundant equipment is generally not allowed for important multi-train systems. Therefore, coincident maintenance on redundant components within the same system (i.e., intra-system coincident maintenance) was excluded from the PRA model for internal events. This same approach was used for the FPRA.
The risk management process, which is governed by a plant Standing Order and PRA guidelines, does not allow multiple high risk components in different systems to be taken out of service at the same time on a regular (planned) basis.
Additionally, operating experience has shown that coincident unavailability of multiple trains of different systems due to unplanned maintenance is rare.
Therefore maintenance data for coincident unavailability of inter-system equipment was not developed. For this reason, coincident maintenance on redundant components *in different systems (i.e., inter-system coincident maintenance) was also excluded from the PRA model for internal events. This approach was also used in the FPRA.
Coincident maintenance on redundant components is addressed in the IEPRA and the FPRA. Therefore, additional quantitative analysis of coincident maintenance for the FPRA is not warranted.
- g. The 2007 PRA Self-Assessment SA-07-48 identifies SR QU-E4 as Not Met against the 2005 version of the American Society of Mechanical Engineers (ASME) PRA standard, but it is not identified in LAR Attachment U-2 nor is it identified or dis positioned in LAR Table V1, so there appears to be no F&O against this SR. In addition, the content and wording for SR QU-E4 changed between the 2005 and 2009 versions of the ASME PRA standard.
Please identify how the IEPRA is affected by important model uncertainty and assumptions and the impact of this on the FPRA. If a finding exists against this SR, please provide the finding and corresponding resolution.
LIC-1 2-0083 Enclosure Page 158 of 164 OPPD's Response to Probabilistic Risk Assessment RAI 15 g.:
During the February 2006 review one F&O was issued that listed QU-E4 as a related SR QU-E4. The following discussion has been provided as resolution to this F&O and is used as the justification for meeting QU-E4. This assessment reflects the wording from ASME/ANS RA-Sa-2009 for this SR. Additionally, the RG 1.200, Rev. 2 clarification for this SR is judged to have no impact upon the current SR grade.
A plant specific analysis has been performed that identifies and characterizes sources of generic modeling uncertainty with respect to FCS. The FCS PRA documents model assumptions within an assumptions database that is kept up to date with the PRA model. Assumptions within this database can be sorted by items such as initiating event, accident sequence, ASME Technical Element and affected system. Assumption characterization involves evaluating the basis for the assumption, determining if reasonable alternative assumptions exist and quantitatively or qualitatively estimating the PRA model impact. Given that key assumptions and sources of uncertainty are application specific, the resulting characterization is in terms of importance (i.e., "Important," "Medium Importance,"
and "Non-Important") as this database applies to the baseline model. OPPD uses a procedure to document the process of entering and characterizing new assumptions.
Sensitivity studies for the internal events PRA model are evaluated as necessary to disposition specific PRA issues that are closed out within the Configuration Control Form (CCF) process. Examples of quantitative sensitivity studies that have been performed include:
- State of Knowledge Correlation impact on Inter-System Loss of Coolant Accidents (ISLOCAs)
- Sensitivity to several Level 2 assumptions
- Impact of a control room HVAC model
- Sensitivity of assumptions made in modeling the revised CS system operation Sensitivity studies were also performed for the following potential risk informed applications (note that these sensitivity studies were performed in support of the PWROG and neither application was pursued by OPPD).
- Evaluation of HPSI model uncertainty for NEI Initiative 4b
- Impact of implementing a staggered engineered safeguards test interval Data uncertainty is routinely evaluated by direct propagation of uncertainty distributions through CAFTA cutsets via UNCERT analyses for each model revision. The results of the FCS UNCERT analysis are documented in the PRA summary notebook.
Both the uncertainty characterization report and the assumptions database identify how the PRA model is affected by given assumptions or uncertainties (either explicitly or generally).
LIC-1 2-0083 Enclosure Page 159 of 164 As additional information for this RAI, the FCS assumptions and sources of uncertainty were reviewed with respect to the NFPA 805 application. The assumptions and uncertainties that were reviewed were compiled from a filter of the assumptions database for "Important" and "Medium Importance," assumptions (note that the criteria for identifying an assumption as, "Important," or, "Medium Importance," are documented in the PRA assumptions procedure) and from the areas of uncertainty with alternative assumptions identified in the uncertainty characterization report. These assumptions and uncertainties were reviewed with respect to the FPRA.
The result of this review showed that one area of uncertainty within the internal events PRA is considered "key" for the NFPA 805 application and should be quantitatively evaluated for the FPRA. Therefore, a sensitivity study was performed that utilized the EPRI methodology for incorporating the time associated with loading critical equipment following offsite power recovery into the Offsite Power Non-Recovery (OSPNR) analysis. To incorporate this, the time to core damage was reduced by 30 minutes to simulate the additional time for operators to load the critical equipment; this is consistent with EPRI TR-1009187, Revision 0 dated October 2003.
Table 3: Results of FPRA Sensitivity Study for PRA RAI 15g Incorporation of EPRI Method for time to Load Critical Equipment (30 min) following Offsite Power Recovery Base Fire Sensitivity Delta PRA* Study VFDR ACDF for NFPA 805 Transition (/yr) 5.72E-06 5.74E-06** 2.11 E-08 VFDR ALERF for NFPA 805 Transition (/yr) 6.67E-07 6.82E-07** 1.51 E-08 Total CDF (Internal Events, Internal Flood 6.01 E-05 6.02E-05 5.71 E-08
& Fire) (/yr) 6.01E-05 6.02E-05 5.71E-08 Total LERF (internal, flood, fire) (/yr) 4.82E-06 4.84E-06 1.64E-08
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
Sensitivity study case for Variance from Deterministic Requirement (VFDR) ACDF and ALERF is conservatively assessed by adding the change in the FPRA CDF and LERF, between the base and sensitivity case, to the VFDR ACDF and ALERF.
As shown in Table 3, the revised probabilities increased the total CDF and LERF by 5.71 E-08/yr and 1.64E-08/yr, respectively. In summary, QU-E4 is judged by OPPD to be met and the F&O associated with this issue is judged by OPPD to be closed. Additionally, an NFPA 805 specific assumptions and uncertainty review was completed and quantified the impact of a key source of uncertainty (treatment of human actions to restore power to critical equipment following off-site power recovery).
In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, when an alternative Loss of Offsite Power (LOOP) recovery assumption is applied to fire PRA supporting the LAR for this area of uncertainty.
LIC-1 2-0083 Enclosure Page 160 of 164 Probabilistic Risk Assessment RAI 16:
The 1999 Westinghouse and Combustion Engineering Owners Group (CEOG) peer review is the last full peer review of the IEPRA. A number of F&Os resulting from this peer review, based on the supplement to the LAR dated December 19, 2011, are not resolved or there was not enough information in the F&O disposition to understand if they were resolved. The following appear to have the potential to noticeably impact the fire PRA results.
- a. The 1999 peer review found that isolation of Component Cooling Water (CCW) to the Spent Fuel Pool (SFP) heat exchangers on a containment isolation signal was not modeled in the PRA (see F&O SY-11). The disposition for this F&O indicates that potential flow diversion is not considered to be significant, and so this flow diversion remains "not modeled." Please provide the impact on the fire CDF and LERF, and on ACDF and ALERF, if any, of not modeling this flow diversion.
OPPD's Response to Probabilistic Risk Assessment RAI 16 a:
Based on an engineering analysis, failure of the isolation of Component Cooling Water (CCW) to the Spent Fuel Pool heat exchangers does not prevent the CCW system from performing its function. Therefore, this failure is not modeled as a diversion flow path in the PRA.
The purpose of this engineering analysis was to determine the ability of the CCW and RW systems to fulfill credited nuclear safety functions in the event of a design basis accident coincident with a single active failure. The conclusions for the CCW system state that no single active failure was identified which would disable the CCW system or prevent the fulfillment of a credited nuclear safety function by the CCW system.
Based on the above, it can be shown that the RW and CCW systems successfully perform their PRA functions if the Spent Fuel Pool (SFP) heat exchanger remains unisolated. Therefore, additional quantitative analysis of the failure to isolate CCW to the SFP heat exchanger for the FPRA is not warranted.
- b. The 1999 peer review found that the AFW system model should include an additional demand for FW-10 (the secured pump) to start given the failure to run of an operating pump (see F&O SY-21). The disposition to this F&O indicates that this failure combination was not modeled in the PRA. Please provide the impact on the 'fire CDF and LERF, and on ACDF and ALERF, if any, of not modeling this failure combination.
OPPD's Response to Probabilistic Risk Assessment RAI 16 b:
A timeline for the operation of the emergency feedwater pumps from the start of a transient is as follows: the motor-driven AFW pump (FW-6) and the turbine-driven AFW pump (FW-10) start by the Auxiliary Feedwater Actuation Signal (AFAS).
LIC-1 2-0083 Enclosure Page 161 of 164 When FW-6 is determined to be running, operators will shutdown FW-10. If FW-6 were to experience a run failure, FW-10 would be restarted, thus introducing a second demand for which the pump may not successfully start. To determine impact of another demand on the fails to start failure mode for FW-10, a sensitivity study was performed that doubled the probability of the fails to start failure mode for FW-10 when present with an FW-6 run failure. This sensitivity study recognized that this additional demand is only applicable given a run failure of FW-6. The results of this sensitivity study are shown below in Table 4.
Table 4: Results of FPRA Sensitivity Study for PRA RAI 16b:
Inclusion of an Additional Demand Failure for FW-10 Given FW-6 Fails to Run Base Fire Sensitivity Delta PRA* Study VFDR ACDF for NFPA 805 Transition (/yr) 5.72E-06 5.73E-06** 7.48E-08 VFDR ALERF for NFPA 805 Transition (/yr) 6.67E-07 6.68E-07** 6.44E-10 Total CDF (Internal Events, Internal Flood 6.01 E-05 6.02E-05 4.95E-08
& Fire) (/yr) 6.01E-05 6.02E-05 4.95E-08 Total LERF (internal, flood, fire) (/yr) 4.82E-06 4.83E-06 2.24E-09
- Base Fire PRA results as reported in Section W.2 of LIC-1 1-0099.
The result of this sensitivity study shows that the impact of the second demand of FW-10 given a riJn failure of FW-6 increased the total CDF and LERF by 4.95E-08/yr and 2.24E-09/yr, respectively. The relative insensitivity of the fire PRA to this model change is primarily due to the low likelihood of fire sequences containing random failure of both FW-6 and FW-1 0. In the FPRA, for a given sequence, when this event combination is multiplied by its fire frequency and the probabilities of additional failures typically that can lead to core damage, the overall risk contribution becomes negligible compared to the sequences that dominate the fire PRA. In conclusion, the net change in risk is within the Region II acceptance criteria of RG 1.174, when the FW-10 failure to start event probabilities are multiplied by a factor of two when they appear in the same cutsets as FW-6 failure to start events.
Probabilistic Risk Assessment RAI 17:
The 1999 peer review of IEPRA and the FPRA peer review found several cases where dependency analysis had not been performed in the HRA for cut sets containing multiple operator errors (see DA-7 and FQ-A-01). Additional issues were identified that the HRA documentation was limited and that the process of obtaining operator review or input was not provided (see F&O HR-3 and HR-4). It appears that the HRA evaluations migrated to the EPRI HRA Calculator after the 1999 peer review. Migration of the HRA to the EPRI HRA Calculator would appear to be a PRA upgrade that would
LIC-1 2-0083 Enclosure Page 162 of 164 warrant a Focused Scope Peer Review per ASME/ANS-SA-Ra-2009, as endorsed by RG 1.200. Instead of a focused scope Peer Review, this upgrade was reviewed as part of the 2007 PRA Self-Assessment. Please explain how the self-assessment, as applied to the HRA element, meets the RG 1.200 endorsed ASME/ANSSA-Ra-2009 PRA standard and industry peer review guidance that requires focused scope peer reviews for PRA upgrades. If it is determined that this self-assessment does not meet the RG 1.200 endorsed standard and guidance as a focused scope peer review, please identify what actions will be taken to address this review deficiency.
OPPD's Response to Probabilistic Risk Assessment RAI 17:
Following the initial 1999 FCS Peer Review, the FCS HRA methodology was converted from a proprietary SAIC methodology to the EPRI HRA Calculator. OPPD agrees that this methodology change does warrant a peer review. For clarification, although this conversion was investigated as part of the 2007 Self-Assessment, this review was primarily focused on changes made to the HRA as a result of the NSSSRP. A more formal review of the HRA upgrade was performed in the February 2006 review.
The February 2006 review was performed generally in accordance with the guidance in Appendix B of Regulatory Guide (RG) 1.200, Rev. 0, considering regulatory interpretations of the ASME Standard as noted in Appendix A of RG 1.200. The differences between this review and the Peer Review requirements documented in Section 6 of ASME RA-Sb-2005 (the version of the ASME PRA standard available at the time) are minimal and are judged to not impact the technical adequacy of the review. Documentation of the review team details (e.g.,
reviewer resume, experience / knowledge base etc...) was not included and a single reviewer was utilized to review multiple PRA elements. However, this review followed the process documented in the original version of NEI 05-04 and was performed by an independent party who had substantial peer review experience and was highly experienced and knowledgeable in all PRA areas being reviewed. All relevant documentation of the PRA changes were transmitted prior to the review and adequate time was allotted for the review of each change.
While the primary focus of this review was to evaluate MSPI, the final evaluation contained a thorough review of all individual SRs from ASME RA-Sb-2005 (including all HRA SRs); a detailed review and the assigned capability category are documented for each SR. The February 2006 review generated a total of 16 B Level F&Os and 9 C Level F&Os, which included two Level B F&Os for HRA. The two HRA Level B F&Os have been addressed and closed for the Revision 11 PRA model. The Enclosure 1 of the LAR supplement dated December 22, 2011, (LIC-11-136), has been revised and includes the final dispositions of these F&Os.
Therefore, based on the quality of the review performed in the February 2006 review, this review is considered to meet the intent of performing a focused scope peer review for the change in HRA methodology. Therefore, no further actions are necessary regarding this issue.
LIC-1 2-0083 Enclosure Page 163 of 164 Probabilistic Risk Assessment RAI 18:
The LAR indicates that a full peer review was performed in 1999 for the IEPRA and in 2010 for the FRPA; and gap or self-assessments have been performed periodically since these peer reviews. The LAR also indicates that the PRA model has been revised periodically and some of these revisions included changes to address findings from these past peer reviews and self-assessments. Please identify any changes made to the IEPRA or FPRA since the last full-scope peer review of each of these PRA models that are consistent with the definition of a "PRA upgrade" in ASME/ANS-RA-Sa-2009, as endorsed by RG 1.200. Also, please address the following:
- a. If any changes are characterized as a PRA upgrade, identify if a focused-scope peer review was performed for these changes consistent with the guidance in ASME/ANS-RA-Sa-2009, as endorsed by Regulatory Guide 1.200, and describe any findings from that focused-scope peer review and the resolution of these findings for this application.
- b. If a focused-scope peer review has not been performed for changes characterized as a PRA upgrade, describe what actions will be implemented to address this deficiency.
OPPD's Response to Probabilistic Risk Assessment RAI 18 a. and b.:
The FCS PRA model was reviewed for PRA changes that meet the definition of a PRA upgrade as defined in non-mandatory Appendix 1-A of ASME/ANS-RA-Sa-2009.' Three changes (other than the HRA calculator upgrade, which is discussed in response to Probabilistic Risk Assessment RAI 17) were identified that are considered to warrant a focused scope peer review:
- The internal flooding analysis was completely revised for the Revision 11 PRA model.
" The LOOP recovery analysis was upgraded to a convolution methodology for calculating LOOP non-recovery factors in the Revision 3 PRA model.
" The LERF PRA model upgraded to the PWROG enhancement to NUREG/CR 6595, in December, 2005.
In December 2008, a focused scope peer review was performed specifically on the Internal Flooding SRs of AMSE RA-Sb-2005. The other two PRA upgrades were reviewed as part of the February 2006 review.
The February 2006 review was performed generally in accordance with the guidance in Appendix B of Regulatory Guide (RG) 1.200, Rev. 0, considering regulatory interpretations of the ASME Standard as noted in Appendix A of RG 1.200. The differences between this review and the Peer Review requirements documented in Section 6 of ASME Ra-Sb-2005 (the version of the ASME PRA standard available at the time) are minimal and are judged to be inconsequential with respect to the technical adequacy of the review. Documentation of the
LIC-1 2-0083 Enclosure Page 164 of 164 review team details (e.g., reviewer resume, experience / knowledge base, etc.) was not included and a single reviewer was utilized to review multiple PRA elements. However, this review followed the process documented in the original version of NEI 05-04 and was performed by an independent party who had substantial peer review experience and was highly experienced and knowledgeable in all PRA areas being reviewed. All relevant documentation of the PRA changes were transmitted prior to the review and adequate time was allotted for the review of each change.
While the primary focus of this review was to evaluate MSPI, the final evaluation contained a thorough review of all individual SRs from ASME RA-Sb-2005 (including all Data and LERF SRs); a detailed review and the assigned capability category are documented for each SR.
The February 2006 review generated a total of 16 B Level F&Os and 9 C Level F&Os, which included two Level C F&Os for Data and two B Level F&Os for LERF. Neither of the two Data F&Os concerned the convolution approach. The two LERF F&Os have been addressed and closed for the Revision 11 PRA model. The Enclosure 1 of the LAR supplement dated December 22, 2011, (LIC-1 1-0136), has been revised and includes the final dispositions of these F&Os.
Therefore, based on the quality of the review performed in the February 2006 review, this review is considered to meet the intent of performing a focused scope peer review for the upgrades of the convolution and LERF methodologies. Therefore, no further actions are necessary regarding this issue.
LIC-1 2-0083 Enclosure, Attachment 1 Page 1 of 9 Attachment 1 - MP RAI 01-1 The NFPA 805 Transition LAR, Section 4.6 Overview of Post-Transition NFPA 805 Monitoring Program REVISED TEXT FROM:
4.6.1 Overview of NFPA 805 Requirements and NEI 04-02 Guidance on the NFPA 805 Fire Protection System and Feature Monitoring Program Section 2.6 of NFPA 805 states:
"A monitoring program shall be establishedto ensure that the availabilityand reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineeringanalysis remain valid."
The intent of the monitoring review is to confirm (or modify as necessary) the adequacy of the existing surveillance, testing, maintenance, compensatory measures, and oversight processes for transition to NFPA 805. This review will consider the following:
The adequacy of the scope of systems and equipment within existing plant programs, i.e., the necessary FP systems and features and nuclear safety capability equipment (NFPA 805 Section 1.5.1) are included.
The performance criteria for the availability and reliability of FP systems and features relied on to demonstrate compliance.
The adequacy of the plant corrective action program in determining causes of equipment and programmatic failures and in minimizing their recurrence.
4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program This section describes the overall Post-Transition NFPA 805 Monitoring Program process. The monitoring program will be implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805. The monitoring process will be conducted in four phases.
- Phase 1 will determine the scope which includes fire protection systems &
features and nuclear safety capability equipment.
- Phase 2 will establish performance criteria.
0 Phase 3 will determine risk significant fire protection program and defense-in-depth elements using criteria established in Phase 2.
- Phase 4 will implement the program after the scope and criteria are established.
Performance and availability monitoring criteria established in Phase 2 will be applied to the risk significant fire protection systems and features and a tracking program will be used on the remaining NFPA 805 required fire protection systems and features.
This process will result in development of a program that reviews the fire protection program performance and identifies trends in performance. The reviews will be based on specific performance goals established to measure the effectiveness of the
LIC-1 2-0083 Enclosure, Attachment 1 Page 2 of 9 fire protection program. Monitoring will ensure that assumptions in engineering analysis remain valid. The monitoring program will be documented in an administrative process (i.e., program manual or directive) that provides the process and sets clear guidelines to consistently measure the performance of the fire protection program.
The phases of the monitoring process are described as follows:
Phase 1 - Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements shall be included in the NFPA 805 monitoring program:
- Fire Protection Structures, Systems, and Components
" Fire protection systems and features required by the NSCA
" Fire protection systems and features modeled in the Fire PRA
" Fire protection systems and features required by Chapter 3 of NFPA 805
- Fire Protection Programmatic Elements
- Key Assumptions in Engineering Analyses As a minimum these fire protection features and systems will be included in the existing fire protection surveillance and system/program health programs. The following process is suggested to determine those fire protection systems and features that may require additional monitoring beyond normal surveillance activities.
- 1. Fire Protection Structures, Systems, and Components Monitoring of SSCs that are required to demonstrate compliance with NFPA 805 is required. These SSCs may include Detection and Alarm Systems, Fire Suppression Systems, Water Supply, Hydrants, and Valves, Fire Pumps, Stand Pipes, Hose Stations, and Hoses, or Fire Barriers, among others. Only those fire protection systems and features required by the NSCA or modeled in the Fire PRA would be considered in scope for the additional monitoring of the NFPA 805 program.
- 2. Monitoring of Fire Protection Programmatic Elements Monitoring of programmatic elements is required in order to "assess the performance of the fire protection program in meeting the performance criteria." Programmatic aspects include:
- Hot Work Control; Administrative Controls
- Fire Watch Programs; Program compliance and effectiveness
- Fire Brigade; Response Times Fire Protection Health Reports, Self-Assessments, regulator and insurance company reports provide inputs to the monitoring program. The monitoring of programmatic elements and program effectiveness may be performed as part of the management of engineering programs. This monitoring is more
LIC-1 2-0083 Enclosure, Attachment 1 Page 3 of 9 qualitative in nature since the programs do not lend themselves to the numerical methods of reliability and availability. These programs form the bases for many of the analytical assumptions used to evaluate compliance with NFPA 805 requirements.
- 3. Monitoring of Key Assumptions in Engineering Analyses The assumptions of the Fire PRA are the primary drivers of the need for monitoring levels of reliability and availability of the SSCs utilized in the risk informed performance based program. These SSCs are generally broken down into two groups, the NSCA (and PRA Internal Events) SSCs and the fire protection systems and features SSCs. Other analytical assumptions from the NSCA, Non-Power Operations and Radioactive Release evaluations may also increase the scope of Fire Protection SSCs or programmatic elements to be reviewed. The NFPA 805 Monitoring program shall be used to monitor the performance of these Fire Protection SSCs at either the component or the functional level.
NSCA and PRA internal events equipment and systems are generally monitored by the Maintenance Rule. It is anticipated that in most cases, for the NSCA type components, the existing Maintenance Rule performance goals will be bounding. Any NSCA equipment and systems not considered under the Maintenance Rule should be reviewed for inclusion in the Maintenance Rule.
Phase 2 - Establishing Risk Criteria Phase 2 of the process is establishing the risk significant criteria and screening for the FP SSCs and programmatic elements within the NFPA 805 monitoring scope.
This may be accomplished at the component, programmatic element, and/or functional level. Since risk is evaluated at the compartment level, criteria must be developed to determine those compartments (or analysis units) for which the FP SSCs are considered risk significant. The Fire PRA is the primary tool used to establish risk significance criteria and performance bounding guidelines. The screening thresholds used to determine risk significant fire compartments are those that meet the following example criteria:
" Risk Achievement Worth (RAW) of the monitored parameter > 2.0 (AND) either
- Large Early Release Frequency (LERF) x (RAW) > 1.0E-8 per year High Safety Significant (HSS) Fire Protection SSCs are those that meet or exceed the risk significant fire compartment screening criteria, and all required FP SSCs, programmatic elements and /or functions are included for each fire compartment.
Low Safety Significant (LSS) Fire Protection SSCs are those that do not meet the risk significant fire compartment screening criteria and are monitored via existing programs/processes. Additionally, the Expert Panel or reviewer may include other fire compartments (and required FP SSCs, programmatic elements and /or functions) that are not risk significant (per the Fire PRA screening criteria) but are included based on plant specific history and/or operational considerations.
LIC-1 2-0083 Enclosure, Attachment 1 Page 4 of 9 As an alternative to including the required FP SSCs, programmatic elements and /or functions in the entire fire compartment, fire protection equipment and features can be included based on a smaller analysis unit (ignition source). Basis needs to be provided when using this approach to ensure adequate monitoring is provided.
EXAMPLE (from NFPA 805 FAQ): For a plant, the power block definition included the turbine building. The Fire PRA had made the entire turbine building (four floors, open to the outside, approximately 52,800 square feet) one fire compartment. Values for CDF and LERF are greater than the threshold, so this fire compartment is screened into the monitoring program.
There are four significant fire sources identified (for CDF and LERF) for this fire compartment. Two fire sources are located in the general service switchgear room on the south side of the 261' elevation, one fire source is located on the northeast corner on the 261' elevation, and one fire source is in the electrical room on the south side of the 286' elevation. These four fire sources would contribute 350 detectors, 18 detector channels, 16 sprinkler valves, and ten manual pull stations into the scope of systems requiring additional monitoring. When just the impact from the four sources within the compartment is considered, the monitored equipment is 42 detectors, three detector channels, one sprinkler valve, and one manual pull station. This accounts for an almost 90% reduction in quantity of monitored equipment while still focusing on the important fire scenarios.
The more practical and realistic approach to this particular fire compartment would be to evaluate each of the four significant fire sources, determine exactly what equipment would mitigate the impact of the four significant fire sources, and to only include that equipment in the monitoring program.
Phase 3 - Risk Determination Phase 3 consists of utilizing the Fire PRA, or other processes as appropriate, to determine target values of reliability and availability for the HSS FP SSCs, programmatic elements and /or functions using the criteria established in Phase 2.
Failure Criteria are established by the Expert Panel or evaluation based on the required FP SSCs, programmatic elements and/or functions assumed level of performance in the supporting analyses. Action levels are established for the SSCs at the component level program level, or functionally through the use of the pseudo system or Performance Monitoring Group concept. The action level is determined based on the number of component, program or functional failures within a sufficiently bounding time period (-2-3 operating cycles). Adverse trends and unacceptable levels of availability, reliability, and performance will be reviewed against established action levels. Documentation of the Monitoring Program failure criteria and action level targets will be contained in the Expert Panel Meeting Minutes or other documented evaluation.
Phase 4 - Monitoring Implementation Phase 4 is the implementation of the monitoring program, once the monitoring scope and criteria are established. The corrective action process will be used to address performance of Fire Protection SSCs that do not meet Performance Criteria.
For HSS Fire Protection SSCs that are monitored, unacceptable levels of availability, reliability, and performance will be reviewed against the established action levels. If
LIC-12-0083 Enclosure, Attachment 1 Page 5 of 9 an action level is triggered, the Expert Panel approves the Corrective Action criteria and action level adjustment if more than usual monitoring is warranted.
A periodic assessment should be performed (e.g., at a frequency of approximately every two to three operating cycles), taking into account, where practical, industry wide operating experience. This may be conducted as part of other established assessment activities. Issues that should be addressed include:
" Review Systems with Performance Criteria. Do performance criteria still effectively monitor the functions of the system? Do the criteria still monitor the effectiveness of the Fire Protection System?
" Have the supporting analyses been revised such that the performance criteria are no longer applicable or new FP SSCs, programmatic elements and/or functions need to be in scope?
" Based on the performance during the assessment period, are there any trends in system performance that should be addressed that are not being addressed?
REVISED TEXT TO:
4.6.1 Overview of NFPA 805 Requirements for the NFPA 805 Monitoring Program Section 2.6 of NFPA 805 states:
"A monitoring program shall be established to ensure that the availabilityand reliability of the fire protection systems and features are maintained and to assess the performance of the fire protection program in meeting the performance criteria. Monitoring shall ensure that the assumptions in the engineeringanalysis remain valid."
4.6.2 Overview of Post-Transition NFPA 805 Monitoring Program The Monitoring program will be described in a procedure which will be prepared and implemented after the safety evaluation issuance as part of the fire protection program transition to NFPA 805 (see implementation item in Attachment S). The monitoring process is comprised of four phases.
- Phase 1 - Scoping
- Phase 2 - Screening Using Risk Criteria
- Phase 3 - Risk Target Value Determination
" Phase 4 - Monitoring Implementation The evaluation conducted as described below which includes these 4 Phases will be documented in an FCS calculation or engineering evaluation.
Phase 1 - Scoping In order to meet the NFPA 805 requirements for monitoring, the following categories of SSCs and programmatic elements shall be included in the NFPA 805 monitoring program:
LIC-12-0083 Enclosure, Attachment 1 Page 6 of 9
- 1) Structures, Systems, and Components "required" to comply with NFPA 805, specifically:
" Fire protection systems and features
- Required by the Nuclear Safety Capability Assessment
- Modeled in the Fire PRA
- Required by Chapter 3 of NFPA 805
" Nuclear Safety Capability Assessment equipment
- Nuclear safety equipment
- Fire PRA equipment
- Non Power Operations (NPO) equipment
- SSCs relied upon to meet radioactive release criteria
- 2) Fire Protection Programmatic Elements Phase 2 - Screening Using Risk Criteria Phase 2 of the process utilizes the risk significance of the SSCs and programmatic elements identified in Phase 1 to establish the appropriate monitoring process to be utilized. The categories of SSCs and programmatic elements from Phase 1 are each screened based on the ability to be able to explicitly establish their risk contribution.
Fire Protection Systems and Features The fire protection systems and features identified in Phase 1 as in scope are screened for risk significance. Risk significance is determined at the component, system, and/or functional level and evaluated on an individual fire area basis. Fire compartments smaller than fire areas may be used instead of fire areas provided the compartments are independent (i.e., share no fire protection SSCs) and sufficient basis is documented. The Fire PRA is used to establish the risk significant screening criteria based on the following thresholds, in accordance with FAQ 10-0059:
Risk Achievement Worth (RAW) of the monitored parameter > 2.0 (AND) either Core Damage Frequency (CDF) x (RAW) > 1.OE-7 per year (OR)
Large Early Release Frequency (LERF) x (RAW) > 1.OE-8 per year CDF, LERF, and RAW (monitored parameter) are calculated for each fire area. The
'monitored parameter' will be established at a level commensurate with the amenability of the parameter to risk measurement (e.g., a fire barrier may be more conducive to risk measurement than an individual barrier penetration).
Fire protection systems and features identified as in scope in Phase 1 will be screened and categorized as either HSS or LSS and included within the appropriate monitoring process based on the risk significance.
The HSS fire protection systems and features are those that meet or exceed the risk significant screening criteria. The HSS fire protection systems and features will be included in the monitoring program contained in the site Maintenance Rule Program described in procedure PED-SEI-34, "Maintenance Rule Program."
LIC-1 2-0083 Enclosure, Attachment 1 Page 7 of 9 The LSS fire protection systems and features are those that do not meet the risk significant screening criteria and are monitored via the existing inspection and test programs and in the existing system/program health program described in PED-SEI-19, "System Health Report Preparation."
Nuclear Safety Capability Assessment Equipment For fires originating during non-power operational (NPO) modes, the qualitative use of fire prevention to manage fire risk during Higher Risk Evolutions does not lend itself to quantitative risk measurement. Therefore, for NSCA Equipment credited for NPO only, no screening is performed and fire risk management effectiveness is monitored programmatically similar to combustible material controls and other fire protection programmatic elements using the existing inspection and test programs and system/program health programs.
All required NSCA equipment, except the NPO scope, identified as in scope in Phase 1 will be screened for safety significance using the fire PRA and the Maintenance Rule guidelines differentiating HSS equipment from LSS equipment.
The HSS NSCA SSCs are those that meet or exceed the risk significant screening criteria. The HSS NSCA equipment will be included in the monitoring program contained in the site Maintenance Rule Program. HSS NSCA equipment may already be appropriately monitored by the Maintenance Rule. A comparison of HSS NSCA equipment to the SSCs that are monitored in the Maintenance Rule program will be performed to determine what equipment will require additional NFPA 805 Monitoring. Also, the review will ensure current Maintenance Rule functions are consistent with the required functions of the HSS NSCA equipment.
All remaining NSCA equipment that is not screened HSS is considered to be LSS and is not included in the monitoring program.
SSCs Relied upon for Radioactive Release Criteria The evaluations performed to meet the radioactive release performance criteria are qualitative in nature. The SSCs relied upon to meet the radioactive release performance criteria are not amenable to quantitative risk measurement. Additionally, since 10 CFR Part 20 limits (which are lower than releases due to core damage and containment breach) for radiological effluents are not being exceeded, equipment relied upon to meet the radioactive release performance criteria is considered inherently LSS. Therefore, monitoring is conducted using the existing inspection and test programs and system/program health programs.
Fire Protection Programmatic Elements Monitoring of programmatic elements is required in order to "assess the performance of the fire protection program in meeting the performance criteria". These programs form the bases for many of the analytical assumptions used to evaluate compliance with NFPA 805 requirements.
Programmatic aspects include:
- Control of Combustible Materials - program compliance and effectiveness, transient exclusion zone effectiveness
LIC-1 2-0083 Enclosure, Attachment 1 Page 8 of 9
" Control of Ignition Sources - program compliance and effectiveness
- Impairment and Compensatory Measures - program compliance and effectiveness
- Industrial Fire Brigade - effectiveness Monitoring of programmatic elements and program effectiveness is more qualitative in nature since they do not lend themselves to the numerical methods of reliability and availability. Therefore, monitoring is conducted using the existing system and program health programs. Fire protection health reports, self-assessments, and regulator and insurance company reports provide inputs to this monitoring program.
Phase 3 - Risk Target Value Determination Phase 3 establishes target values for reliability and availability for the HSS fire protection systems and features and NSCA equipment.
HSS SSCs Reliability and availability criteria are established by evaluation based on the HSS fire protection system or features or NSCA equipment's assumed level of reliability/availability in the supporting analyses.
Action levels are established for the HSS SSCs at the component level, program level, or functionally through the use of the pseudo system or the 'performance monitoring group' concept. The actual action level is determined based on the number of component, program or functional failures within a sufficiently bounding time period (-2-3 operating cycles). In addition, the EPRI Technical Report (TR) 1006756, "Fire Protection Surveillance Optimization and Maintenance Guide for Fire Protection Systems and Features" may be used as input for establishing reliability targets, action levels, and monitoring frequency. When establishing the action level threshold for reliability and availability, the action level will be no lower than the fire PRA assumptions.
Where HSS SSCs are identified using the Maintenance Rule guidelines, the performance criteria may be established based on the Maintenance Rule, provided the criteria are consistent with Fire PRA assumptions.
LSS SSCs The LSS Fire Protection Systems and Features are included within the existing inspection and test programs and system and program health programs which ensure functionality and no reliability and availability criteria are assigned. LSS SSCs are not included in any monitoring process therefore reliability and availability criteria are not required. NPO equipment which is LSS is included within the existing inspection and test programs which ensure functionality and no reliability and availability criteria are assigned. Rad Release SSCs are LSS and are included within the existing inspection and test programs which ensure functionality, and no reliability and availability criteria are assigned. Fire protection programmatic elements are considered LSS and do not lend themselves to the numerical methods of reliability and availability so their effectiveness is based on the objective and anecdotal evidence evaluated by the engineers in charge of the programs.
LIC-1 2-0083 Enclosure, Attachment 1 Page 9 of 9 Phase 4 - Monitoring Implementation Phase 4 is the implementation of the monitoring program, once the monitoring scope and criteria are established.
For HSS fire protection systems and features and NSCA equipment that are monitored, the actual levels of availability, reliability, and performance will be reviewed against the established action levels. If an action level is triggered, the Corrective Action Program governed by SO-R-2, "Condition Reporting and Corrective Action," is used to identify the adverse condition. A corrective action plan will then be developed to ensure performance returns to the established level.
When applicable, a sensitivity study can be performed to determine the margin below the action level that still provides acceptable fire PRA results to help prioritize corrective actions if the action level is reached.
A periodic assessment of the Monitoring Program will be included within the scope of the routine Fire Protection Program assessment which is described in LAR Section 4.7.3. The scope of the Monitoring Program assessment will include the following:
- Review systems with performance criteria. Do performance criteria still effectively monitor the functions of the system? Do the criteria still monitor the effectiveness of the fire protection and nuclear safety capability assessment systems?
- Have the supporting analyses been revised such that the performance criteria are no longer applicable or new fire protection or nuclear safety capability assessment SSCs, programmatic elements and/or functions need to be in scope?
- Based on the assessment period, are there any trends in monitored elements that should be addressed that are not being addressed?
- Has external Operating Experience and Internal Operating Experience when applicable been incorporated into the Monitoring program?
LIC-1 2-0083 Enclosure, Attachment 2 Page 1 of 18 Attachment 2 - SSD RAI 04-2 LAR Section 4.2.3 Licensing Action Transition Results REVISED BULLETED ITEM FROM (necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
- Separation of raw water pumps, discharge valves, and strainers inside the intake structure REVISED BULLETED ITEM TO (necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
- Separation of raw water pumps and discharge valves inside the intake structure LAR Attachment C Table B-3 Fire Area Transition Fire Area 30 Method of Accomplishment REVISED TEXT FOR DECAY HEAT REMOVAL FROM (necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
Align turbine-driven AFW pump FW-10 to provide feedwater supply from the EFWST to both steam generators through the AFW header, from the MCR.
REVISED TEXT FOR DECAY HEAT REMOVAL TO (necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
Align turbine-driven AFW pump FW-10 to provide feedwater supply from the EFWST to both steam generators through the AFW header, from the MCR. AFW flow can be established to both steam generators; and steam rejection to atmosphere can be established from both steam generators. Depending on the fire location within containment, each steam generator will have minimum of one channel of pressure and level indication available. Depending on fire location, the minimum RCS loop temperature indication available will be channels B and C of T-hot and T-cold for RCS loops 1A and 1B (for steam generator RC-2A), or channels A and D of T-hot and T-cold for RCS loops 2A and 2B (for steam generator RC-2B).
REVISED TEXT FOR PROCESS MONITORING FROM (necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
LIC-1 2-0083 Enclosure, Attachment 2 Page 2 of 18 RCS Loop 1 Hot Leg Temperature Channel A (A/TE-1 12H)
RCS Loop 1 Hot Leg Temperature Channel B (B/TE-1 12H)
RCS Loop 1 Hot Leg Temperature Channel C (CITE-i 12H)
RCS Loop 1 Hot Leg Temperature Channel D (DITE- 12H)
RCS Loop 1A Cold Leg Temperature Channel A (A/TE-1 12C)
RCS Loop 1A Cold Leg Temperature Channel C (C/TE-1 12C)
RCS Loop 1B Cold Leg Temperature Channel B (B/TE-1 12C)
RCS Loop 1B Cold Leg Temperature Channel D (D/TE-1 12C)
RCS Loop 2 Hot Leg Temperature Channel A (A/TE-122H)
RCS Loop 2 Hot Leg Temperature Channel B (B/TE-122H)
RCS Loop 2 Hot Leg Temperature Channel C (C/TE-122H)
RCS Loop 2 Hot Leg Temperature Channel D (D/TE-122H)
RCS Loop 2A Cold Leg Temperature Channel A (A/TE-122C)
RCS Loop 2A Cold Leg Temperature Channel C (C/TE-122C)
RCS Loop 2B Cold Leg Temperature Channel B (B/ITE-122C)
RCS Loop 2B Cold Leg Temperature Channel D (D/TE-122C)
REVISED TEXT FOR PROCESS MONITORING TO (necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
RCS Loop 1 Hot Leg Temperature Channel B (BITE-i 12H) and RCS Loop 1 Hot Leg Temperature Channel C (C/TE-i 12H) and RCS Loop 1A Cold Leg Temperature Channel C (C/TE-i 12C) and RCS Loop 1B Cold Leg Temperature Channel B (B/TE-1 12C)
Or RCS Loop 2 Hot Leg Temperature Channel A (A/TE-122H) and RCS Loop 2 Hot Leg Temperature Channel D (DITE-122H) and RCS Loop 2A Cold Leg Temperature Channel A (A/TE-122C) and RCS Loop 2B Cold Leg Temperature Channel D (D/TE-122C)
Fire Area 34B-1 Performance Goal / Method of Accomplishment / Comments REVISED TEXT FROM in "Comments" associated with RCS Inventory Control (self-identified change identified during the FCS NRC triennial fire protection audit in March 2012, to reflect that spurious operation of non-high-low pressure interface valve PCV-102-1 (PORV) is not postulated in the deterministic NFPA 805 analysis for the 2 conductor, 480VAC, two phase power cable in Fire Area 34B-1, consistent with and to similar treatment for the PORV power cable in containment, fire area 30):
See VFDR No. 34B-1-001, 34B-1-010, 34B-1-011, 34B-1-012, 34B 013, 34B-1-014, 34B-1-015, and 34B-1-016; Note that these RCS Inventory Control VFDRs also apply to RCS Pressure Control
LIC-1 2-0083 Enclosure, Attachment 2 Page 3 of 18 REVISED TEXT TO in "Comments" associated with RCS Inventory Control (self-identified change identified during the FCS NRC triennial fire protection audit in March 2012, to reflect that spurious operation of non-high-low pressure interface valve PCV-102-1 (PORV) is not postulated in the deterministic NFPA 805 analysis for the 2 conductor, 480VAC, two phase power cable in Fire Area 34B-1, consistent with and to similar treatment for the PORV power cable in containment, fire area 30):
See VFDR No. 34B-1 -001, 34B-1 -011, 34B-1 -012, 34B-1 -013, 34B 014, 34B-1-015, and 34B-1-016; Note that these RCS Inventory Control VFDRs also apply to RCS Pressure Control Variances from Deterministic Requirements (VFDR)
DELETED IN ENTIRETY (self-identified change identified during the FCS NRC triennial fire protection audit in March 2012, to reflect that spurious operation of non-high low pressure interface valve PCV-102-1 (PORV) is not postulated in the deterministic NFPA 805 analysis for the 2 conductor, 480VAC, two phase power cable in Fire Area 34B-1, consistent with and to similar treatment for the PORV power cable in containment, fire area 30):
VFDR No. 34-1-101 PCV 102 1 Pressurfizer power oper-ated- relief v~alve PCV 102 1 fails due.
to c-able damage (EC4175, EC3544, EC11429). Fire damage to cabl!
EC354and E- can
. .1120.. the a eto ca.use . .puriouI
.pen, bu,t Gca be mitigated from the main control room; however, fire damage to cabl.
EC475 can.. c.ause th pre . ..r power operated relief valve to spuriously open without the ability to mitigate fro~m a main control room switch. The pressuriz power op..atd relief. valve may Reed to be closed to prevent lo-ss Of RCS pressure and inventory and mnaintain the RCS pressure boundary. This componeRt supp...s RCS inventory an pressure co*nrol. Thi0 condition represen}ets a varianre from deterministic requirements of NEPP.A 805, Section 4.2.3. This is a separationise tion asos IA sceie A FeE)ey Idie ~*9 RA 6 for thi-s VFQ_ bhased- on an;;evaluaMtion4 of the VFDR's impact to defense-in-depth features Mithin fire area 314B1-.
The R.A has;- boonP dePmonsrEAated to be feasible and to net have adverse risk impact This recover,' action has been; credited for defense in depth only, a*d has not boon credi;te,4d in the Fire PRA quantificatin* foF this fire area. ConGesequently, the reliability of this reover,' action is implicitly zFero (i.e., not quantified) in the Fire PRA quantification for this fire area. Given that his eovy, act ion has been evaluated for both failtyand for
-ad-verserisk impact, the contribution to the quantified risqk ARceasoiated- wfith this fire area can only be between zero and a mnarginal positivke impovmet in the quantified risk baed on; anassuImed- finite value of reliability. Det-ails; of the evaluation for risk, safety mriad defens~e in deeth are loc-ated withi the Fire Risk Evlainfor the fire area.
LIC-12-0083 Enclosure, Attachment 2 Page 4 of 18 Fire Area 36C Fire Area Comments ADDED TEXT (necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
Note that 'fire area 360" is a credited ERFBS pyrocrete enclosure located within fire area 36B. "Fire area 36C" contains only train A cable trays and conduits. "Fire area 360" is not enclosed by three hour rated fire boundaries and is not a fire area per the rule; however, the designation
'lire area 360" is used in NFPA 805 Nuclear Safety Capability Assessment (NSCA) to identify the set of cables contained within the Pyrocrete enclosure and thereby facilitate the performance of NSCA and Non-power Operational modes (NPO) separation analyses for the set of cables contained within the pyrocrete enclosure.
LAR Attachment G Recovery Actions Transition Step 4 - Evaluation of the Feasibility of Recovery Actions Results of Step 4:
DELETED IN ENTIRETY (self-identified change during the FCS triennial fire protection audit in March 2012, to reflect that spurious operation of the non-high low pressure interface valve PCV-102-1 (PORV) is not postulated in the deterministic NFPA 805 analysis for the 2 conductor, 480VAC, two phase power cable in' Fire Area 34B-1, consistent with and to similar treatment for the PORV power cable in containment, Fire Area 30):
0 Fi*r, AreA 24 R-1 fir in poeot"ica"
. ponotration room open bre-aker MO31Al(in fire area 340) to fail closed PQR'. PC'.-
102 1 (Rionlatn AcRS for PreurreingVeto G4ry Control)actio Ruirod complete no later than 60 minuters (the neod- for this; action will be based on dentction in the Fe adra, and conStirgent iniAr r-0-oe 0 bhae~d_ On the obseryed ROS pressure fromp crdte ain conrol1 roominstumenatio i.e., operator observes RCS pressure decreasing, attempts- to c-lose bilock* valvos anmid PO-RIVs1 firoimmi conrol1 room, but does net observe RIGS pressure to stabilize).
Table G-1 Recovery Actions and Activities Occurring at the Primary Control Station DELETED IN ENTIRETY (self-identified change during the FCS triennial fire protection audit in March 2012, to reflect that spurious operation of the non-high low pressure interface valve PCV-102-1 (PORV) is not postulated in the deterministic NFPA 805 analysis for the 2 conductor, 480VAC, two phase power cable in Fire Area 34B-1, consistent with and to similar treatment for the PORV power cable in containment, Fire Area 30):
Fare Area.: 34B-1
LIC-1 2-0083 Enclosure, Attachment 2 Page 5 of 18 Component: PGV-102-1 Component Decr~iption: Proccurizor Powcr Opcrated Relief Value Actfione:s Manually, opon breaker MCC 30-G1 -AG! (in fire arca 34G) to fail clocod PORV PCV 102 1 (isolate RGS for pressure/inventory control).
VFDR: 341. 1 010 RAIPGS! RA LAR Attachment K Existing Licensing Action Transition Licensing Action - Appendix R Exemption, Electrical Penetration, Lack of area wide suppression (III.G.2 and III.G.3 criteria)
Basis REVISED TEXT FROM (Self-identified change to provide more complete descriptionof originallicensing action basis):
In lieu of pressurizer auxiliary spray, AFW is being credited for RCS pressure control to depressurize the RCS to SDC (LPSI Shutdown Cooling) entry conditions. Note that the "alternate shutdown capability" for a fire in this area referred to the utilization of AFW in lieu of pressurizer auxiliary spray. The use of AFW in this capacity is no longer considered to represent an "alternate shutdown strategy", and as such the existing portion(s) of NRC exemption NRC-85-0200 for fire area 34B-1 regarding the use of AFW in this capacity is not deemed to be required for NFPA 805 transition. Note that loss of all pressurizer heaters in fire area 34B-1 has been identified as a Variance from the Deterministic Requirements (VFDR) of NFPA 805 due to the difficulty presented in maintaining the plant in hot shutdown, "safe and stable" for NFPA 805.
REVISED TEXT TO (Self-identified change to provide more complete description of originallicensing action basis):
In lieu of pressurizer heaters and/or pressurizer auxiliary spray, AFW is being credited for RCS pressure control and to depressurize the RCS to SDC (LPSI Shutdown Cooling) entry conditions. Note that the "alternate shutdown capability" for a fire in this area referred to the utilization of AFW in lieu of pressurizer heaters and/or pressurizer auxiliary spray. The use of AFW in this capacity is no longer considered to represent an "alternate shutdown strategy", and as such the existing portion(s) of NRC exemption NRC-85-0200 for fire area 34B-1 regarding the use of AFW in this capacity is not deemed to be required for NFPA 805 transition. Note that loss of all pressurizer heaters in fire area 34B-1 has been identified as a Variance from the Deterministic Requirements (VFDR) of NFPA 805
LIC-1 2-0083 Enclosure, Attachment 2 Page 6 of 18 due to the difficulty presented in maintaining the plant in hot shutdown, "safe and stable" for NFPA 805.
LAR Attachment T Clarification of Prior NRC Approvals Prior Approval Clarification Request 1 Current Licensing Basis REVISED TEXT FROM (To better discriminate and better explain the separation compliance strategy for RCS temperature instrumentation as opposed to the separation compliance strategy for steam generator pressure and level instrumentation, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC in March 2012.)
Steam generator level and pressure instrumentation, RCS temperature instrumentation, and source range neutron flux monitoring instrumentation is located inside containment. Each process monitoring variable has four redundant channels. The cables for these instruments are routed such that at least one redundant channel is routed over 20 feet from any other redundant channel. However, this minimum separation of 20 feet is not in all cases free of intervening combustibles. The NRC approved exemption identifies that intervening combustibles may be present where there is 20 feet of separation; but this is justified based on there being limited quantities of intervening combustibles, and the existing fire protection. Based on the NRC approved exemption for containment dated July 3, 1985 (NRC-85-0200), a minimum of one division of safe shutdown instrumentation can be credited to remain free of fire damage for a containment fire. The following instrumentation and associated cables are located in containment:
SG Level Instrument Supporting Cables Penetration Raceways A/LT-911 EA12135A A-11 48C, 47C, 2C, 3C, 4C, 5C A/LT-912 EA12151A A-11 48C, 47C, 2C, 3C, 4C B/LT-911 EB12139A A-4 61C, 60C, 7C, 9C B/LT-912 EB12155A A-4 61C,60C,7C,9C, 12C C/LT-911 EC12143A D-5 59C, 58C, 30C, 31C, 34C, RISER C/LT-912 EC12159A D-5 59C D/LT-911 ED12147A D-10 50C,49C,20C,28C, 25C, 26C, 27C, RISER, 15C D/LT-912 ED12163A D-10 50C
LIC-12-0083 Enclosure, Attachment 2 Page 7 of 18 SG Pressure Inst. Supportingq Cables Penetration Raceways A/PT-913 EA1 2137A A-11 48C, 47C, 2C, 3C, 4C, 5C A/PT-914 EA12153A A-11 48C, 47C, 2C, 3C, 4C B/PT-913 EB12141A A-4 61C,60C,7C,9C B/PT-914 EB12157A A-4 61C,60C,7C,9C, 12C C/PT-913 EC12145A D-5 59C, 58C, 30C, 31C, 34C, RISER C/PT-914 EC12161A D-5 59C D/PT-913 ED12149A D-10 50C, 49C, 20C, 28C, 25C, 26C, 27C, RISER, 15C D/PT-914 ED12165A D-10 50C RCS Temp. Inst. Supportingq Cables Penetration Raceways AITE-1 12C EA1 1488C A-11 CONDUIT AITE-1 12H EA11491C A-11 CONDUIT A/TE- 122C EA 1492C A-11 CONDUIT AITE-122H EA11495C A-11 CONDUIT B/TE-i 12C EB1 1496C A-4 CONDUIT B/TE-1 12H EB11499C A-4 CONDUIT B/ITE-122C EB1 1500C A-4 CONDUIT B/TE-122H EB1 1503C A-4 CONDUIT CITE-1 12C EC1 1504C D-5 CONDUIT C/TE-1 12H EC11505C D-5 CONDUIT C/TE-122C ECi 1506C D-5 CONDUIT C/TE-122H EC11507C D-5 CONDUIT D/TE-112C ED1 1508C D-10 CONDUIT D/TE-1 12H ED11509C D-10 CONDUIT DiTE-122C ED1151OC D-10 CONDUIT D/TE-1 22H EDl1511C D-10 CONDUIT Neutron Flux Inst. Supportincq Cables Penetration Raceways NE-O01 EA3019A B-11 CONDUIT EA3019B N/A CONDUIT EA3020A B-1I CONDUIT EA3020B N/A CONDUIT NE-002 EB3012A B-4 CONDUIT EB3012B N/A CONDUIT
LIC-12-0083 Enclosure, Attachment 2 Page 8 of 18 EB3013A B-4 CONDUIT EB3013B N/A CONDUIT NE-003 EC3026A C-5 CONDUIT EC3026B N/A CONDUIT EC3027A C-5 CONDUIT EC3027B N/A CONDUIT NE-004 ED3005A C-10 CONDUIT ED3005B N/A CONDUIT ED3006A C-10 CONDUIT ED3006B N/A CONDUIT REVISED TEXT TO (To better discriminate and better explain the separation compliance strategy for RCS temperature instrumentation as opposed to the separation compliance strategy for steam generator pressure and level instrumentation, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC in March 2012.)
Steam generator level and pressure instrumentation, and source range neutron flux monitoring instrumentation is located inside containment.
Each process monitoring variable has four redundant channels. The cables for these instruments are routed such that at least one redundant channel is routed over 20 feet from any other redundant channel.
However, this minimum separation of 20 feet is not in all cases free of intervening combustibles. The NRC approved exemption identifies that intervening combustibles may be present where there is 20 feet of separation; but this is justified based on there being limited quantities of intervening combustibles, and the existing fire protection. Based on the NRC approved exemption for containment dated July 3, 1985 (NRC 0200), a minimum of one division of safe shutdown instrumentation can be credited to remain free of fire damage for a containment fire. The following instrumentation and associated cables are located in containment:
SG Level Instrument Supporting Cables Penetration Raceways A/LT-911 EA12135A A-11 48C, 47C, 2C, 3C, 4C, 5C A/LT-912 EA12151A A-11 48C, 47C, 2C, 30, 4C B/LT-911 EB12139A A-4 61C, 60C, 7C, 9C B/LT-912 EB12155A A-4 61C, 60C, 7C, 9C, 12C C/LT-911 EC12143A D-5 59C, 58C, 30C, 31 C, 34C, RISER C/LT-912 EC12159A D-5 59C D/LT-911 ED12147A D-10 50C,49C,20C,28C, 25C, 26C, 27C, RISER, 15C
LIC-1 2-0083 Enclosure, Attachment 2 Page 9 of 18 D/LT-912 ED12163A D-10 50C SG Pressure Inst. Supporting Cables Penetration Raceways A/PT-913 EA12137A A-11 48C, 47C, 2C, 3C, 4C, 5C A/PT-914 EA12153A A-11 48C, 47C, 20, 3C, 4C B/PT-913 EB12141A A-4 61C,60C,7C,9C B/PT-914 EB12157A A-4 61C, 60C,7C, 9C, 12C C/PT-913 EC12145A D-5 59C, 58C,30C,31C, 34C, RISER C/PT-914 EC12161A D-5 59C D/PT-913 ED12149A D-10 50C, 49C, 20C, 28C, 25C, 26C, 27C, RISER, 15C D/PT-914 ED12165A D-10 50C Neutron Flux Inst. Supporting Cables Penetration Raceways NE-001 EA3019A B-11 CONDUIT EA3019B N/A CONDUIT EA3020A B-11 CONDUIT EA3020B N/A CONDUIT NE-002 EB3012A B-4 CONDUIT EB3012B N/A CONDUIT EB3013A B-4 CONDUIT EB3013B N/A CONDUIT NE-003 EC3026A C-5 CONDUIT EC3026B N/A CONDUIT EC3027A C-5 CONDUIT EC3027B N/A CONDUIT NE-004 ED3005A C-10 CONDUIT ED3005B N/A CONDUIT ED3006A C-10 CONDUIT ED3006B N/A CONDUIT Reactor coolant system hot leg and cold leg temperature instrumentation is located inside containment. Each steam generator has four redundant channels of reactor coolant system hot leg and cold leg temperature instrumentation. The cables for channels B and C of reactor coolant system hot leg and cold leg temperature instrumentation associated with steam generator RC-2A are routed over 20 feet away from the cables for channels A and D of reactor coolant system hot leg and cold leg
LIC-12-0083 Enclosure, Attachment 2 Page 10 of 18 temperature instrumentation associated with steam generator RC-2B.
However, this separation of more than 20 feet is not in all cases free of intervening combustibles. The NRC approved exemption identifies that intervening combustibles may be present where there is 20 feet of separation; but this is justified based on there being limited quantities of intervening combustibles, and the existing fire protection. Based on the NRC approved exemption for containment dated July 3, 1985 (NRC 0200), a minimum of channels A and D of reactor coolant system hot leg and cold leg temperature instrumentation associated with steam generator RC-2A or channels B and C of reactor coolant system hot leg and cold leg temperature instrumentation associated with steam generator RC-2B can be credited to remain free of fire damage for a containment fire. The following instrumentation and associated cables are located in containment:
RCS Temp. Inst. Supporting Cables Penetration Raceways A/TE-1 12C (for RC-2A) EA11488C A-11 CONDUIT A/TE-1 12H (for RC-2A) EA11491C A-11 CONDUIT BITE-1 12C (for RC-2A) EB11496C A-4 CONDUIT B/TE-i 12H (for RC-2A) EB1 1499C A-4 CONDUIT CITE-1 12C (for RC-2A) ECi 1504C D-5 CONDUIT C/TE-112H (for RC-2A) EC11505C D-5 CONDUIT D/TE-1 12C (for RC-2A) ED1 1508C D-10 CONDUIT D/TE-1 12H (for RC-2A) ED11509C D-10 CONDUIT A/TE-122C (for RC-2B) EA11492C A-11 CONDUIT A/TE-122H (for RC-2B) EA11495C A-11 CONDUIT B/TE-122C (for RC-2B) EB11500C A-4 CONDUIT BITE-122H (for RC-2B) EB1 1503C A-4 CONDUIT C/TE-122C (for RC-2B) EC11506C D-5 CONDUIT C/TE-122H (for RC-2B) ECI 1507C D-5 CONDUIT DITE-122C (for RC-2B) ED1151OC D-10 CONDUIT DITE-122H (for RC-2B) EDl1511C D-10 CONDUIT Request REVISED TEXT FROM (To better discriminate and better explain the separation compliance strategy for RCS temperature instrumentation as opposed to the separation compliance strategy for steam generator pressure and level instrumentation, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC in March 2012.)
Based on the cable routing (drawings 11405-E-92 Sh. 1, 11405-E-93 Sh.
1, and 11405-E-98) in containment, the associated electrical penetrations for train A (A-11, B-11) and train D (C-10, D-10) instrumentation are separated by more than 20 feet (horizontally) from the associated
LIC-1 2-0083 Enclosure, Attachment 2 Page 11 of 18 electrical penetrations for train B (A-4, B-4) and train C (C-5, D-5) instrumentation. Furthermore, the cables are routed (in trays or in conduits) in containment with at least 20 feet of separation such that at least one channel each of steam generator pressure and level, RCS temperature, and source range neutron flux instrumentation will remain free of fire damage, although the same channel may not necessarily be available for all of the required instrumentation in a particular fire location.
The NRC approved exemption identifies that intervening combustibles may be present; however, the exemption should clearly identify the instruments, cables, and routes that form the basis of instrumentation availability (as identified in the tables above).
REVISED TEXT TO (to better discriminate and better explain the separation compliance strategy for RCS temperature instrumentation as opposed to the separation compliance strategy for steam generator pressure and level instrumentation, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC in March 2012.)
Based on the cable routing (drawings 11405-E-92 Sh. 1, 11405-E-92 Sh.
2, 11405-E-93 Sh. 1, 11405-E-93 Sh. 2, and 11405-E-98) in containment, the associated electrical penetrations for train A (A-11, B-1i1) and train D (C-10, D-10) instrumentation are separated by more than 20 feet (horizontally) from the associated electrical penetrations for train B (A-4, B-4) and train C (C-5, D-5) instrumentation. Furthermore, the cables are routed (in trays or in conduits) in containment with at least 20 feet of separation such that at least one channel each of steam generator pressure and level, and source range neutron flux instrumentation will remain free of fire damage, although the same channel may not necessarily be available for all of the required instrumentation in a particular fire location. A minimum of channels A and D of reactor coolant system hot leg and cold leg temperature instrumentation associated with steam generator RC-2A or channels B and C of reactor coolant system hot leg and cold leg temperature instrumentation associated with steam generator RC-2B can be credited to remain free of fire damage for a containment fire. Furthermore, for any containment fire, auxiliary feedwater flow can be established to both steam generators; and steam rejection to atmosphere can be established from both steam generators.
The NRC approved exemption identifies that intervening combustibles may be present; however, the exemption should clearly identify the instruments, cables, and routes that form the basis of instrumentation availability (as identified in the tables above).
REVISED TEXT FROM (To specifically include steam generator pressure and level instrumentation and pressurizer pressure and level instrumentation in the paragraphs discussing the requested clarification for instrument sensing lines, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC in March 2012.)
LIC-12-0083 Enclosure, Attachment 2 Page 12 of 18 The NRC approved exemption must also be clarified to identify that the exemption basis provides justification for why the existing separation between the instrument sensing lines for redundant channels of instrumentation is adequate to provide reasonable assurance that one channel of instrumentation will remain free of fire damage. The exemption basis for this includes the following justification:
Within the containment, the redundant instrument channels and sensing lines have a minimum of 20 feet of horizontal separation with minimal intervening combustibles. The intervening combustibles consist of lightly loaded cable trays. The sensing lines have common point of origin (i.e.,
steam generators and pressurizer). At the points of origin it is not possible to achieve physical separation. From the point of origin the lines are routed in different directions to the transmitters which have a minimum of 20 feet of horizontal separation. The steam generator instruments are located in separate quadrants within containment and typically have 50 feet of separation. This separation is consistent with the separation discussed and credited in the NRC SER dated July 3, 1985 (NRC-85-0200). This SER grants an exemption from 20 feet of separation with no intervening combustibles for certain areas within containment. The SER specifically addresses the pressurizer bays and areas where the intervening combustibles are made up of IEEE-383 qualified cables. The instrument sensing line routings meet these criteria and therefore are considered to be covered under this exemption. Based on this, the instrument sensing lines have adequate separation to support NFPA 805 safe shutdown requirements for providing at least one channel of reliable indication for process monitoring of pressurizer level and pressure, and steam generator level and pressure.
REVISED TEXT TO (to specifically include steam generator pressure and level instrumentation and pressurizer pressure and level instrumentation in the paragraphs discussing the requested clarification for instrument sensing lines, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC).
The NRC approved exemption must also be clarified to identify that the exemption basis provides justification for why the existing separation between the instrument sensing lines for redundant channels of steam generator pressure and level instrumentation and pressurizer pressure and level instrumentation is adequate to provide reasonable assurance that one channel of instrumentation will remain free of fire damage. The exemption basis for this includes the following justification:
Within the containment, the redundant instrument channels and sensing lines have a minimum of 20 feet of horizontal separation with minimal intervening combustibles. The intervening combustibles consist of lightly loaded cable trays. The sensing lines have common point of origin (i.e.,
steam generators and pressurizer). At the points of origin it is not possible to achieve physical separation. From the point of origin the lines
LIC-1 2-0083 Enclosure, Attachment 2 Page 13 of 18 are routed in different directions to the transmitters which have a minimum of 20 feet of horizontal separation. The steam generator instruments are located in separate quadrants within containment and typically have 50 feet of separation. This separation is consistent with the separation discussed and credited in the NRC SER dated July 3, 1985 (NRC-85-0200). This SER grants an exemption from 20 feet of separation with no intervening combustibles for certain areas within containment. The SER specifically addresses the pressurizer bays and areas where the intervening combustibles are made up of IEEE-383 qualified cables. The instrument sensing line routings meet these criteria and therefore are considered to be covered under this exemption. Based on this, the instrument sensing lines have adequate separation to support NFPA 805 safe shutdown requirements for providing at least one channel of reliable indication for process monitoring of steam generator pressure and level and pressurizer pressure and level.
Prior Approval Clarification Request 4 Current Licensing Basis DELETED TEXT (deletion of superfluous information not germane to the requested clarification, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
Howoyer, the 6eparati.n between ,-cabl*s for redrundant n-trum-, I TL l 04X and L=T 404Y is only 1 8' at thc rsuie bay, Which is not justified by the NRC approved xeme~ption. UAs resuwlt of thc lacgk of adequate separation c botwonh,,,L.-- iX a-ndl LT "IlY, it i6 possible that neitheF ntRumcnet Will bo available for a fire ocurnn the pressurize bay (L 106 will remain fre o. fire damage in this; c.as.e). Pre6surizer heater availability r-equire either IT-X o-r LX TTlVY to remain free, of fire damage. A plant moedificattioen isceic orpaeH 0 ihafu position switch, where the now position isolates6 the coil Of relay 63X'LIC!
101 4fro LIC alarm output (or similar des'-ign change to mneet sae intent). The modiji*ation ,Fo will allow plant operators to isolate th spuiou sgnals from L=T- 40X and L=T- 10Y and control the pressurizer heaters from the main control room.
Request REVISED TO CORRECT TYPOGRAPHICAL ERROR FROM (necessary correction identified during the LAR audit RAI breakout meeting between OPPD and NRC):
Pzr. Level Inst. Penetration LT-101X B-5 LT-106 C-4
LIC-12-0083 Enclosure, Attachment 2 Page 14 of 18 REVISED TO CORRECT TYPOGRAPHICAL ERROR TO (necessary correction identified during the LAR audit RAI breakout meeting between OPPD and NRC):
Pzr. Level Inst. Penetration LT-101X C-4 LT-106 B-5 Prior Approval Clarification Request 6 Request REVISED TEXT FROM (deletion of superfluous information not germane to the requested clarification, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC in March 2012):
.... bay - adequate separation is provided based on the floor slab at el.
1013' between the tray sections for the train A (at el. 1013' below) and train B heaters (at el. 994' above), which effectively functions as a radiant energy shield. The inadequate separation between the cables for LT-101X and LT-101Y at the pressurizer bay must be addressed. LT-101X or LT-101Y need to remain free of fire damage for the backup heaters to work from the main control room. A plant modification is credited to replace HC-101-1 with a four position switch, where the new position isolates the coil of relay 63X/LIC-101 from LIC alarm output (or similar design change to meet same intent). The modification will allow for plant operators to isolate the spurious signals from LT-101X and LT-101Y and control the backup groups of pressurizer heaters from the main control room.
REVISED TEXT TO (enhanced description of requested clarification for separation of redundant pressurizer heater power cables, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
...bay - for the redundant pressurizer heater power cables not in the immediate vicinity of the pressurizer cubicle or the electrical penetrations (penetrations A-1, A-2, D-1, D-2), the vertical separation between the cable trays for the redundant pressurizer heater power cables is greater than 24 feet, and the cable trays for the redundant pressurizer heater power cables are separated by a combination of floor slab and grated steel platform, which effectively function as a radiant energy shield.
Prior Approval Clarification Request 7 Request REVISED TEXT FROM (text in 2 nd bullet to remove credit for Raw Water Strainer availability from the discussion of an intake structure fire event at el. 985' (at the intake structure stairwell), necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC).
LIC-1 2-0083 Enclosure, Attachment 2 Page 15 of 18 For a fire at intake structure el. 985' (at the intake structure stairwell),
a three-hour rated fire barrier enclosure has been provided for the power and/or control cables associated with all four RW pumps, their associated discharge valves, and their associated strainers. This enclosure is identified as fire area 31A in the 10 CFR 50 Appendix R analysis and the NFPA 805 Nuclear Safety Performance Criteria (NSPC) analysis. The following cables are contained within the three-hour rated fire barrier enclosure:
Component Supporting Cables AC-10A EA66 AC-10B EB67 AC-10C EC68 AC-10D ED69 HCV-2850 EA7302 HCV-2851 EB7309 HCV-2852 EC7316 HCV-2853 ED7323 HCV-2874A EA7306 HCV-2874B EB7307 HCV-2875A EA7313 HCV-2875B EB7314 HCV-2876A EC7320 HCV-2876B ED7321 3B3/1 B3B-6 A227 4C4/1 B4C-7 B228 The July 3, 1985 NRC approved exemption (NRC-85-0200) is being clarified accordingly to explicitly define the safe shutdown compliance strategy for the intake structure, and to specify that a three-hour rated fire barrier enclosure has been provided for the power and control cables associated with all four RW pumps, their associated discharge valves, and their associated strainers at intake structure el. 985' (at the intake structure stairwell).
REVISED TEXT TO (text in 2 nd bullet to remove credit for Raw Water Strainer availability from the discussion of an intake structure fire event at el. 985' (at the intake structure stairwell), necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC).
For a fire at intake structure el. 985' (at the intake structure stairwell),
a three-hour rated fire barrier enclosure has been provided for the power and/or control cables associated with all four RW pumps and their associated discharge valves. This enclosure is identified as fire area 31A in the 10 CFR 50 Appendix R analysis and the NFPA 805 Nuclear Safety Performance Criteria (NSPC) analysis. The following cables are contained within the three-hour rated fire barrier enclosure:
LIC-1 2-0083 Enclosure, Attachment 2 Page 16 of 18 Component Supporting Cables AC-10A EA66 AC-10B EB67 AC-10C EC68 AC-10D ED69 HCV-2850 EA7302 HCV-2851 EB7309 HCV-2852 EC7316 HCV-2853 ED7323 HCV-2874A EA7306 HCV-2874B EB7307 HCV-2875A EA7313 HCV-2875B EB7314 HCV-2876A EC7320 HCV-2876B ED7321 3B3/1 B3B-6 A227 4C4/1 B4C-7 B228 The July 3, 1985 NRC approved exemption (NRC-85-0200) is being clarified accordingly to explicitly define the safe shutdown compliance strategy for the intake structure, and to specify that a three-hour rated fire barrier enclosure has been provided for the power and control cables associated with all four RW pumps and their associated discharge valves at intake structure el. 985' (at the intake structure stairwell).
REVISED TEXT FROM (text in 4 th bullet to clarify that credit for Raw Water Strainer availability in Prior Approval Clarification Request 7 only applies to the discussion of a fire event at pull boxes 128T and 129T, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC).
For a fire at pull boxes 128T and 129T el. 998' (south wall of auxiliary building), the NRC approved exemption dated July 3, 1985 (NRC 0200) was granted on the basis that the limited fire loading and widely dispersed combustible materials in the vicinity of the pull boxes will ensure that fire propagation is limited and will provide reasonable assurance that at least one train of the RW system will remain free of fire damage prior to fire brigade intervention, and that no other safe shutdown systems will be fire damaged. Pull boxes 128T and 129T are located outside at grade elevation 998' on the south wall of the auxiliary building. The pull boxes are separated by approximately 3 1/2 feet. Based on an evaluation of the RW pumps, their associated discharge valves and strainers, and the routing of their associated power and control cables, it has been conservatively determined by OPPD that only pump AC-10A (train A) or pump AC-10D (train B) can be credited for safe shutdown within the bases of the NRC approved exemption, whereby one train of the RW system is assumed to have sustained fire damage prior to successful manual suppression by the fire brigade, with the other train of the RW system having remained
LIC-1 2-0083 Enclosure, Attachment 2 Page 17 of 18 unaffected by the fire (for a fire located in the vicinity of the pull boxes). Pump AC-10A, associated discharge valve HCV-2850, and associated strainer AC-12A (MCC-3B3), or pump AC-1OD, associated discharge valve HCV-2853, and associated strainer AC-12B (MCC-4C4) are the only equipment combinations available for the RW system that do not require cross train cables. Train A cables EA66 (AC-10A), EA7302 (HCV-2850), and A227 (AC-12A) are routed together in the duct banks through pull box 129T; similarly, train B cables ED69 (AC-10D), ED7323 (HCV-2853), and B228 (AC-12B) are routed together in the duct banks through pull box 128T (reference drawing 11405-E-320). The July 3, 1985, NRC approved exemption (NRC-85-0200) is being clarified accordingly to explicitly define the safe shutdown compliance strategy for the intake structure, and to specify that either pump AC-10A or AC-10D is credited in conjunction with its associated discharge valve and strainer, based on the similar cable routes to the RW pump 4KV power cables.
REVISED TEXT TO (text in 4 t" bullet to clarify that credit for Raw Water Strainer availability in Prior Approval Clarification Request 7 only applies to the discussion of a fire event at pull boxes 128T and 129T, necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC).
For a fire at pull boxes 128T and 129T el. 998' (south wall of auxiliary building), the NRC approved exemption dated July 3, 1985 (NRC 0200) was granted on the basis that the limited fire loading and widely dispersed combustible materials in the vicinity of the pull boxes will ensure that fire propagation is limited and will provide reasonable assurance that at least one train of the RW system will remain free of fire damage prior to fire brigade intervention, and that no other safe shutdown systems will be fire damaged. Pull boxes 128T and 129T are located outside at grade elevation 998' on the south wall of the auxiliary building. The pull boxes are separated by approximately 3 1/2 feet. Based on an evaluation of the RW pumps, their associated discharge valves and strainers, and the routing of their associated power and control cables, it has been conservatively determined by OPPD that only pump AC-10A (train A) or pump AC-10D (train B) can be credited for safe shutdown within the bases of the NRC approved exemption, whereby one train of the RW system is assumed to have sustained fire damage prior to successful manual suppression by the fire brigade, with the other train of the RW system having remained unaffected by the fire (for a fire located in the vicinity of the pull boxes). Pump AC-10A, associated discharge valve HCV-2850, and associated strainer AC-12A (MCC-3B1), or pump AC-10D, associated discharge valve HCV-2853, and associated strainer AC-12B (MCC-4C1) are the only equipment combinations available for the RW system that do not require cross train cables. Train A cables EA66 (AC-10A), EA7302 (HCV-2850), and A227 (AC-12A) are routed together in the duct banks through pull box 129T; similarly, train B cables ED69 (AC-10D), ED7323 (HCV-2853), and B228 (AC-12B) are
LIC-12-0083 Enclosure, Attachment 2 Page 18 of 18 routed together in the duct banks through pull box 128T (reference drawing 11405-E-320). The July 3, 1985, NRC approved exemption (NRC-85-0200) is being clarified accordingly to explicitly define the safe shutdown compliance strategy for the intake structure, and to specify that either pump AC-10A or AC-10D is credited in conjunction with its associated discharge valve and strainer (one strainer or the other is credited for a fire at pull boxes 128T and 129T), based on the similar cable routes to the RW pump 4KV power cables.
Prior Approval Clarification Request 10 Request ADDED TEXT (necessary clarification identified during the LAR audit RAI breakout meeting between OPPD and NRC):
Note that 'lire area 36C" is a credited ERFBS pyrocrete enclosure located within fire area 36B. "Fire area 360" contains only train A cable trays and conduits. "Fire area 36C" is not enclosed by three hour rated fire boundaries and is not a fire area per the rule; however, the designation
'lire area 36C" is used in NFPA 805 Nuclear Safety Capability Assessment (NSCA) to identify the set of cables contained within the Pyrocrete enclosure and thereby facilitate the performance of NSCA and Non-power Operational modes (NPO) separation analyses for the set of cables contained within the pyrocrete enclosure.