L-PI-16-067, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors- Response to Request for Additional Information
| ML16230A554 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 08/17/2016 |
| From: | Northard S Northern States Power Co, Xcel Energy |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| CAC ME9734, CAC ME9735, L-PI-16-067 | |
| Download: ML16230A554 (14) | |
Text
Xcel Energy@ Prairie Island Nuclear Generating Plant 1717 Wakonade Drive East Welch, MN 55089 August 17, 2016 L-PI-16-067 10 CFR 50.90 10 CFR 50.48(c)
U.S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Prairie Island Nuclear Generating Plant Units 1 and 2 Dockets 50-282 and 50-306 Renewed License Nos. DPR-42 and DPR-60 License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors- Response to Request for Additional Information (CAC Nos. ME9734 and ME9735)
References:
- 1. NSPM letter, J.P. Sorensen to NRC Document Control Desk, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors, L-PI-12-089, dated September 28, 2012, ADAMS Accession No. ML12278A405.
- 2. NSPM letter, S. Sharp to NRC Document Control Desk, Supplement to License Amendment Request to Adopt NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactors, L-PI-14-045, dated April 30,2014 (ADAMS Nos. ML14125A106 and ML14125A149).
- 3. NRC email, T. Beltz to S. Chesnutt, Prairie Island Nuclear Generating Plant, Units 1 and 2 - NFPA 805 Requests for Additional Information and Response Timeline (TAG Nos. ME9734 and ME9735), dated March 30, 2015 (ADAMS Accession No. ML15089A157).
- 4. NSPM letter, K. Davison to NRC Document Control Desk, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors - Response to Request for Additional Information, L-PI-15-041, dated May 28, 2015 (ADAMS No. ML15153A018).
- 5. NSPM letter, K. Davison to NRC Document Control Desk, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors - Response to Request for Additional Information- 90-Day Responses, L-PI-15-052, dated June 19, 2015 (ADAMS No. ML15174A139).
Document Control Desk Page2
- 6. NSPM letter, K. Davison to NRC Document Control Desk, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors - Response to Final Request for Additional Information (PRA- Second Round), L-PI-15-059, dated October 22, 2015 (ADAMS No. ML15296A259).
- 7. NRC email, T. Beltz to A. Hazel hoff, Prairie Island Nuclear Generating Plant-Requests for Additional Information (AFPB) re: LAR to Adopt NFPA 805 (TAG Nos. ME9734 and ME9735), dated January 8, 2016 (ADAMS No.
- 8. NSPM letter, K. Davison to NRC Document Control Desk, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors - Response to Request for Additional Information, L-PI-16-004, dated January 20, 2016 (ADAMS No.
- 9. NSPM letter, K. Davison to NRC Document Control Desk, License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactors- Response to Request for Additional Information, L-PI 005, dated May 24, 2016 (ADAMS No. ML16152A046).
- 10. NRC email, R. Kuntz to A. Hazel hoff, Prairie Island Nuclear Generating Plant-Requests for Additional Information re: LAR to Adopt NFPA 805 (CAC Nos.
ME9734 and ME9735), dated July 26, 2016 (ADAMS No. ML16208A540).
In Reference 1, the Northern States Power Company, a Minnesota Corporation (NSPM) doing business as Xcel Energy, requested approval from the Nuclear Regulatory Commission (NRC) to transition the fire protection licensing basis for the Prairie Island Nuclear Generating Plant (PINGP) to 10 CFR 50.48(c), National Fire Protection Association Standard 805 (NFPA 805). Supplemental information was provided in letters dated November 8, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12314A144) and December 18, 2012 (ADAMS Accession No. ML12354A464).
In Reference 2, NSPM submitted a revised Fire Probabilistic Risk Assessment (PRA) in a supplement to the subject License Amendment Request (LAR). In Reference 3, the NRC staff provided requests for additional information (RAis) regarding this request and also provided a timeline and due dates for submitting responses within 60, 90, or 120 days after an on-site Audit that was conducted March 23-25, 2015. NSPM letter dated May 28, 2015 (Reference 4) provided responses to the 60-day RAis and one of the 90-day RAis (Fire Protection Engineering RAI 03). NSPM letter dated June 19, 2015 (Reference 5) provided responses to the remaining 90-day RAis.
NSPM letter dated October 22, 2015 (Reference 6) provided responses to second round RAis and included an Attachment L request, "Approval Request 4- Wiring Above
Document Control Desk Page 3 Suspended Ceilings." In Reference 7, the NRC staff provided RAison this Attachment L request. In Reference 8, NSPM provided responses to these RAis. In Reference 9, NSPM provided the final RAI response. In Reference 10, the NRC staff provided additional RAis and these were discussed during an informal clarification call on July 21' 2016.
Enclosure 1 to this letter provides NSPM's response to the RAis received in Reference 10 and also provides one licensee identified issue.
This letter is submitted in accordance with 10 CFR 50.90. The additional information provided in this letter does not impact the conclusions of the No Significant Hazards Evaluation or Environmental Considerations Evaluation presented in Reference 2.
In accordance with 10 CFR 50.91, NSPM is notifying the State of Minnesota of this additional information by transmitting a copy of this letter to the designated State Official.
If there are any questions or if additional information is needed, please contact Gene Eckholt at 651-267-1742.
Summary of Commitments This letter contains no new commitments and makes no revisions to any existing commitments.
I declare under penalty of perjury that the foregoing is true and correct.
Executed on August 17, 2016.
~otfJJJ&_
Scott Northard Acting Site Vice President- Prairie Island Nuclear Generating Plant Northern States Power Company- Minnesota Enclosures (1) cc: Administrator, Region Ill, USNRC NRR Project Manager, PINGP, USNRC Resident Inspector, PINGP, USNRC State of Minnesota
L-PI-16-067 NSPM Enclosure 1 Response to Requests for Additional Information (RAis)
Regarding the License Amendment Request to Adopt National Fire Protection Association (NFPA) Standard 805 at Prairie Island Nuclear Generating Plant Units 1 and 2 and Submittal of Licensee Identified Item NRC Request By letter dated September 28, 2012 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML12278A405), Northern States Power Company, a Minnesota corporation (NSPM, the licensee), doing business as Xcel Energy, submitted a license amendment request (LAR) to transition its fire protection licensing basis at the Prairie Island Nuclear Generating Plant (PINGP), Units 1 and 2, from paragraph 50.48(b) of Title 10 of the Code of Federal Regulations (1 0 CFR) to 10 CFR 50.48(c),
National Fire Protection Association Standard NFPA 805 (NFPA 805). Supplemental information has been requested by the NRC staff and provided {by] NSPM.
The latest supplemental information was provided by NSPM in letter dated May 24, 2016 (Accession No. ML16152A046). Enclosure 1 to the letter provides NSPM's response to [Probabilistic Risk Assessment (PRA)] RAJ 03, including the response to PRA RAJ 01.h. Enclosure 2 provides licensee identified LAR changes. Enclosure 3 provides an updated LAR Attachment M, License Condition Changes. Enclosure 4 provides an updated LAR Attachment S, Plant Modifications and Items for Implementation. Enclosure 5 provides an updated LAR Attachment W, Fire PRA Insights. The U.S. Nuclear Regulatory Commission (NRC) staff has reviewed the May 24, 2016, letter and determined that additional information is required to complete its technical review.
NRC Request (PRA RAJ 01.h.01)
The response to PRA RAJ 01.h demonstrated through a sensitivity analysis that the total plant-level core damage frequency (CDF) for PINGP, Units 1 and 2, as well as the total delta (L\) CDF for PINGP, Unit 1 exceed RG 1.174, Revision 2, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant Specific Changes to the Licensing Basis" (ADAMS Accession No. ML100910006), risk acceptance guidelines should the generic fire ignition frequencies be based upon those in Table 6-1 of NUREG-CR-6850, "EPRIINRC-RES Fire PRA Methodology for Nuclear Power Facilities, Volume 2: Detailed Methodology." However, inconsistent with accepted guidance in Chapter 10 of NUREG/CR-6850, Supplement 1, "Fire Probabilistic Risk Assessment Methods Enhancements," the response did not describe fire protection, or related, measures that can be taken to provide additional defense-in-depth (DID);
rather, the response only indicated that DID was already evaluated as part of the fire risk evaluations (FREs) for each fire area. Moreover, while the response identified the general conservatisms in the analysis, it did not discuss the risk significance of these conservatisms for those fire scenarios or areas most impacted by the sensitivity analysis. Therefore, as originally requested in PRA RAJ 01.h, and commensurate with Page 1 of 11
L-PI-16-067 NSPM Enclosure 1 any potential increase in risk significance associated with use of generic fire ignition frequencies based upon those in Table 6-1 of NUREG/CR-6850, provide the following:
(1) describe the fire protection, or related, measures that will be taken to provide additional DID, and (2) discuss relevant conservatisms in the analyses for those scenarios most impacted by the sensitivity studies and the risk significance of those conservatisms.
NSPM Response (PRA RAI 01.h.01)
Defense-In-Depth (DID) is a balance of three echelons including:
- 1. Prevent fires from starting;
- 2. Detect, control, and extinguish promptly those fires that do occur;
- 3. Protect the reactor so that a fire that is not promptly extinguished by the fire suppression activities will not prevent the safe shutdown of the plant.
Defense in Depth features will be discussed for each of the Fire Areas with the greatest contribution to delta CDF below.
An evaluation of the sensitivity results against Regulatory Guide 1.174 indicates that the delta risks by fire area using the NUREG/CR-6850 ignition frequencies meet the risk acceptance guidelines defined for Region II of Figures 4 and 5 in Regulatory Guide 1.174 on an individual fire area basis. However, the sum of the CDF for Unit 1 and Unit 2 and delta CDF increase in risk for Unit 1 for the overall plant exceeds the acceptance guidelines. Therefore further discussion of Fire Protection Defense in Depth is provided. The Unit 2 delta CDF increase and total large early release frequency (LERF) and delta LERF results still meet Region II and Region Ill of Figures 4 and 5 in Regulatory Guide 1.174 and require no further discussion.
The fire ignition frequency sensitivity results indicate that most of the contribution to delta CDF is in the following Fire Areas:
Fire Area Description U1 ACDF Impact U2 ACDF Impact 13 Control Room 59% 71%
18 Relay and Cable 21% 28%
Spreading Room 20 U1 4.16KV BUS 16 14% 0%
Room Most of the risk impact is due to Fire Areas 13, 18, and 20, and therefore only these fire areas will be discussed. Below is a summary of the scenarios most impacted by the sensitivity analysis for each area, the conservatisms in the analyses for scenarios most impacted, and the risk significance of these model conservatisms.
Page 2 of 11
L-PI-16-067 NSPM Enclosure 1 Fire Area 13: Control Room The fire scenarios contributing to the delta CDF for the control room are in main control board panels 81, 82, C1, C2 and Panel G1. Panel G1, Safeguards Electrical Equipment Panel, contributes 50% of the delta CDF in FA 13. The scenario associated with Panel G1 that contributes to the delta CDF is a promptly suppressed fire generating a plant trip as noted in NSPM's June 19, 2015 response to PRA RAI 18 (ADAMS No.,
ML15174A139). Even though this fire scenario is promptly suppressed, Control Room abandonment is postulated due to loss of safeguards electrical equipment control, and the delta risk of recovery actions for Alternate Shutdown in this higher frequency fire scenario is the reason this is significant. The target set for this fire scenario includes a bounding localized impact for the panel. This treatment is conservative as the worst case target components are assigned as the target. The actual fire scenario could result in a less risk significant component failure on the panel. The PINGP Fire PRA quantifies the small fires in the main control board panels as affecting the worst possible target in the panel generating a plant trip. Assuming that even the small fires in the safeguards electrical control panel damage targets that force control room abandonment has a significant impact on delta CDF risk calculation because the delta risk of recovery actions for Alternate Shutdown is much higher than the delta risk for scenarios that don't require Alternate Shutdown outside the Control Room.
Defense-In-Depth for Fire Area 13, Control Room:
The first echelon of DID to prevent fires from starting includes features in the Fire Protection Program that control and limit combustibles and fire ignition sources, including hot work permits and robust design features that limit the likelihood of fires starting. Combustible materials are minimized in safety related areas like the control room.
The second echelon of DID to detect and quickly suppress fires is addressed by the fact that the main control room is constantly attended with qualified operators that can promptly suppress potential fires with extinguishers or fire hoses located near the area.
The main control room includes ionization fire detectors that were credited in the Fire PRA.
The third echelon of DID to protect safe shutdown of the plant is addressed by the fact there are diverse systems capable of removing decay heat from the reactor available in the control room. The procedure for Alternate Shutdown includes the actions to take to isolate the control room and shutdown outside the control room if the control room must be abandoned. Additionally, all of the actions taken outside the main control room for Alternate Shutdown were considered to be recovery actions and were included in the delta risk of recovery actions calculation. NFPA 805 FAQ 07-0030 allows actions taken at the Primary Control Station to be excluded from the delta risk calculation. As described in NSPM's May 28, 2015 response to SSA RAI 01 (ADAMS No.
ML15153A018), PINGP Hot Shutdown Panels include process monitoring indication for Source Range Neutron Flux, RCS Temperature, RCS Pressure, Pressurizer Level, and Page 3 of 11
L-PI-16-067 NSPM Enclosure 1 Steam Generator Level to support safe shutdown. The panel in the Auxiliary Feedwater Pump room is the primary command and control location after the Control Room is abandoned due to fire. This location contains controls for multiple components and is similar to an Alternate Shutdown Panel, but was not specifically approved as an Alternate Shutdown Panel in response to Generic Letter 81-12. Because the Hot Shutdown Panel was not considered to be a Primary Control Station as described in Regulatory Guide 1.205 and FAQ 07-0030, none of the PINGP Alternate Shutdown actions were credited to be Primary Control Station actions; therefore, the calculation of delta risk for recovery actions is a conservative calculation. If the Alternate Shutdown actions were considered to be Primary Control Station actions, the risk of Primary Control Station actions could be excluded from the delta risk of recovery actions calculation and the delta risk of recovery actions would be much lower.
The Fire PRA model for Alternate Shutdown conservatively assumes a Loss of Offsite Power and includes the delta risk of locally starting the Emergency Diesel Generators and re-powering the electrical buses. The Alternate Shutdown procedure includes steps for cases where offsite power is not available and the cases where offsite power remains available. For cases where offsite power remains available, the risk is much lower. Because the Alternate Shutdown procedure includes actions for cases where offsite power remains available and this was not credited in the Fire PRA, these actions in the Alternate Shutdown procedure provide additional Defense in Depth for safe shutdown.
Fire Area 18: Relay and Cable Spreading Room:
The scenarios contributing to the delta CDF in the Relay and Cable spreading room are several relay cabinet scenarios that propagate beyond the ignition source. The main model conservatism in this area is the conservative assumption that unless a conduit location has been specifically identified within the Fire Area, it was assumed to be within the Zone of Influence (ZOI) of all fire ignition sources in the Fire Area. Fire Area 18 at PINGP contains many cable tray stacks that extend from near the ignition sources up to the ceiling. There are also many conduits in the fire area that can be obstructed by the cable tray configurations in the area. Therefore, for conduits whose locations were not identified by walkdowns or drawings, a conservative assumption was applied (i.e.,
conduits were assumed to be failed in all the fire scenarios within the fire area). For cases where the quantification indicated that specific cables in conduits were risk-significant, detailed walkdowns and drawing reviews were performed to only map these conduits to the fire scenarios where the conduit could be damaged in the zone of influence of the fire initiator, thereby removing excessive conservatism. However, a number of conduits are still conservatively failed in all fire scenarios. This assumption causes more equipment to be failed in most of these fire scenarios resulting in conservative risk estimates.
The risk increase in FA 18 is due to the increased frequency of Bin 15, Electrical Cabinet fires. FA 18 contains many electrical cabinets in close proximity to electrical cable trays. The electrical cabinet fire scenarios that are not suppressed prior to fire Page 4 of 11
L-PI-16-067 NSPM Enclosure 1 propagation fail the cable trays within the Zone of Influence and fail all conduits in FA 18 that have not been specifically located by drawing review or walkdown.
The Fire PRA conservatively assumed the Heat Release Rate (HRR) for electrical cabinets in FA 18 fire modeling based on the 981h percentile HRR for vertical cabinets with qualified cable, and fire in more than one cable bundle because that is the bounding HRR for electrical cabinets. The HRR curve directly impacts the severity factor and probability of non-suppression for these scenarios. Less conservative assumptions regarding HRR for electrical cabinets would directly reduce the CDF and
.LlCDF of these fire scenarios.
Defense-In-Depth for Fire Area 18, Relay and Cable Spreading Room:
The first echelon of DID to prevent fires from starting includes features in the Fire Protection Program that control and limit combustibles and fire ignition sources, including hot work permits and robust design features that limit the likelihood of fires starting. Combustible materials are minimized in safety related areas like the relay and cable spreading room.
The second echelon of DID to detect and quickly suppress fires is addressed by the fact that FA 18 has ionization fire detectors in the area and will install a Very Early Warning Fire Detection System in the area to quickly detect fires that do start. An automatic Carbon Dioxide Fire Suppression system is installed in the area, which is also capable of being manually actuated as well. Finally, a qualified fire brigade is available at all times to manually suppress fires that do occur. The Fire PRA model conservatively assumed all cables in this area fail at thermoplastic damage thresholds even though most of the cable in the area has been qualified to IEEE 383-1974 which would justify using the higher damage threshold for thermoset cables. Current Engineering Manuals specify that IEEE 383 qualified cables should be installed going forward.
The third echelon of DID to protect safe shutdown of the plant is addressed by the fact there are diverse systems capable of removing decay heat from the reactor. The procedure for Alternate Shutdown includes the actions to take to isolate the control room and shutdown outside the control room if the control room must be abandoned.
Additionally, all of the actions taken outside the main control room for Alternate Shutdown were considered to be recovery actions and included in the delta risk of recovery actions calculation. NFPA 805 FAQ 07-0030 allows actions taken at the Primary Control Station to be excluded from the delta risk calculation. As described in NSPM's response to SSA RAI 01 (ADAMS No. ML15153A018), PINGP Hot Shutdown Panels include process monitoring indication for Source Range Neutron Flux, RCS Temperature, RCS Pressure, Pressurizer Level, and Steam Generator Level to support safe shutdown. The panel in the Auxiliary Feedwater Pump room is the primary command and control location after the Control Room is abandoned due to fire. This location contains controls for multiple components and is similar to an Alternate Shutdown Panel, but was not specifically approved as an Alternate Shutdown Panel in response to Generic Letter 81-12. Because the Hot Shutdown Panel was not Page 5 of 11
L-PI-16-067 NSPM Enclosure 1 considered to be a Primary Control Station as described in Regulatory Guide 1.205 and FAQ 07-0030, none of the PINGP Alternate Shutdown actions were credited to be Primary Control Station actions; therefore, the calculation of delta risk for recovery actions is a conservative calculation. If the Alternate Shutdown actions were considered to be Primary Control Station actions, the risk of Primary Control Station actions could be excluded from the delta risk of recovery actions calculation and the delta risk of recovery actions would be much lower.
The Fire PRA model for Alternate Shutdown conservatively assumes a Loss of Offsite Power and includes the delta risk of locally starting the Emergency Diesel Generators and re-powering the electrical buses. The Alternate Shutdown procedure includes steps for cases where offsite power is not available and the cases where offsite power remains available. For cases where offsite power remains available, the risk is much lower. Because the Alternate Shutdown procedure includes actions for cases where offsite power remains available and this was not credited in the Fire PRA, these actions in the Alternate Shutdown procedure provide additional Defense in Depth for safe shutdown.
Fire Area 20: Unit 1 4.16KV Safeguards Switchgear (Bus 16)
The scenarios contributing to the delta CDF are several transient fires and the high energy arcing fault fire scenario for Bus 16. The ZOI identified for the HEAF scenarios is the same as the ZOI identified for the peak HRR of the electrical cabinet. This approach is conservative as Appendix Min NUREG/CR-6850 identifies a ZOI of 5 ft vertically and 1 ft horizontally from the cabinet's front or rear panel (versus the 10 ft by 10ft horizontal ZOI assigned to fixed ignition sources in the PINGP Fire PRA). A reduced ZOI for the HEAF scenario would result in lower risk contribution for the corresponding fire scenarios in Fire Area 20 if fewer targets are mapped to them and therefore, the current ZOI is conservative.
Defense-In-Depth for Fire Area 20, Unit 1 4.16KV Safeguards Switchgear (Bus 16):
The first echelon of DID to prevent fires from starting includes features in the Fire Protection Program that control and limit combustibles and fire ignition sources, including hot work permits and robust design features that limit the likelihood of fires starting. Combustible materials are minimized in safety related areas like the Bus 16 .
room. Additional defense in depth is provided by operator tours in areas containing safety-related cables or equipment at least once each shift to detect fire hazards, including combustible materials.
The second echelon of DID to detect and quickly suppress fires is addressed by the fact that FA 20 has ionization fire detectors in the area. A qualified fire brigade is available at all times to manually suppress fires that do occur. The Fire PRA model conservatively assumed all cables in this area fail at thermoplastic damage thresholds even though most of the cable in the area has been qualified to IEEE 383-1974 which would justify using the higher damage threshold for thermoset cables. Current Page 6 of 11
L-PI-16-067 NSPM Enclosure 1 Engineering Manuals specify that IEEE 383 qualified cables should be installed going forward.
The third echelon of DID to protect safe shutdown of the plant is addressed by the fact there are diverse systems capable of removing decay heat from the reactor.*
Modifications have been completed to ensure Train A Process Monitoring safe shutdown equipment remains unaffected by a fire in FA 20. The delta risk is due to a potential Multiple Spurious Operation (MSO) scenario where fire damage to cables causes the 12 Motor Driven Auxiliary Feedwater Pump (MD AFWP) to spuriously start and loss of power to discharge Motor Operated Valves to isolate flow. Failure to isolate flow could cause an over-fill the Steam Generator and subsequent failure of the Turbine Driven Auxiliary Feedwater Pump. The Fire Risk Evaluation concluded that the delta risk of this MSO was low and did not credit a recovery action to mitigate this VFDR.
Conclusion The risk increase resulting from the fire ignition frequency sensitivity study has been identified and the appropriate balance of defense in depth features provides assurance that risk is reasonable. Most of the delta risk increase is due to the risk of recovery actions for Alternate Shutdown due to fires in the Control Room and Relay and Cable Spreading Room. The assumption that all of the risk of Alternate Shutdown actions is included in the calculation of the risk of recovery actions ensures the risk is bounded (and not excluded by classifying them as Primary Control Station actions). The assumption of a loss of offsite power is conservative and the fact that the Alternate Shutdown procedure provides guidance for scenarios where offsite power remains available provides additional defense in depth for the Control Room and Relay and Cable Spreading Room. The 4.16KV Bus 16 room contributes a smaller percentage to the delta risk. Defense in depth features for the 4.16KV Bus 16 room include the fact that it is a vital area, storage of transient combustible materials is minimized, and operator tours are performed in areas containing safety-related cables or equipment at least once each shift to detect fire hazards, including combustible materials.
References
- 1. FPRA-PI-FQ, Fire Risk Quantification Notebook, rev 3
- 2. FPRA-PI-MCR, Main Control Room Analysis
- 3. FPRA-PI-RRA, Relay Room Analysis
- 4. PINGP Procedure F5 Appendix B, Control Room Evacuation (Fire)
- 5. PINGP 195, Turbine Building Data
- 7. EC 26796, Fire Risk Evaluation Page 7 of 11
L-PI-16-067 NSPM Enclosure 1 NRC Request (PRA RAJ 20)
The discussion provided under Item 14 of Licensee-Identified Issue 5, included in to the letter dated May 24, 2016 (see page 68 of 74 of Enclosure 2),
explains that the scope of modifications described in LAR AttachmentS, Table S-2, Modification Item S2-10 has been reduced. As a result, the FPRA appears to have been updated to explicitly model the risk of the resulting breaker coordination and cable protection vulnerabilities and associated VFDRs. Item 22 of Licensee-Identified Issue 5 (see page 70 of 74) discusses similar issues. In parlicu/ar to circuit overcurrent protection inadequacies, the letter discusses that they are associated with fire damage to external direct current (DC) control cables that supporl overcurrent protection for medium voltage (4.16KV) switchgear breakers. While Items 14 and 22 of Licensee-Identified Issue 5 provide some discussion of how these inadequacies are modeled in the FPRA, this discussion is limited and does not provide sufficient detail for the NRC staff to complete its review. Provide the following information:
a) Explain how inadequate coordination and overcurrent protection was modeled in the FPRA, and justify that this treatment addresses the failures that could occur as a result of the identified circuit inadequacies.
b) Include a description of the circuit failure modes addressed and how associated component failures were modeled in the FPRA. Also, describe and justify assumptions made in the FPRA about how fire-induced faults associated with inadequately coordinated/protected circuits impact upstream and downstream components from the fault.
c) Given the lack of electrical coordination and cable protection as a result of the reduced scope of modifications discussed in Items 14 and 22 of Licensee-Identified Issue 5, include an explanation of how the potential for secondary fires was addressed in the FPRA. If secondary fires were not modeled and fire-induced faults in inadequately protected circuits could lead to secondary fires, then justify this modeling exclusion.
d) In line with the issues described above, the NRC staff a/so notes that a risk-informed (RI) approach was used in association with a reduction in scope of circuit-protection-related modifications (e.g., Item S2-10 of LAR AttachmentS, Table S-2) by modeling overcurrent failure modes and secondary fires. These modifications would have provided overcurrent trip protection for ceria in circuits.
As a result,
- i. Describe how the plant response model addresses fire-induced faulting of one or more load circuits at the same time that DC control power has been lost due to fire damage.
ii. Explain how the lack of circuit protection and the resulting potential for common enclosure issues are analyzed.
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L-PI-16-067 NSPM iii. Discuss the potential for secondary fires and describe how secondary fires are modeled (e.g., fir.e size, zone of influence, propagation, etc.).
iv. Discuss the potential for high energy arcing faults to be created as a result of the inability to clear fire-induced load faults at the load breaker, and describe how such faults are modeled (fault location, zone of influence, fire propagation, etc.). Without DC control power being available at the switchgear, circuit breakers upstream may have to clear the fault and those breakers will very likely have a much higher overcurrent setpoint than that required to protect the integrity of the cables being faulted.
NSPM Response (PRA RAI 20) a) The fire induced loss of over-current protection on 4.16KV breakers could lead to loss of breaker coordination for the 4.16KV and over-current conditions on 4.16KV power cables leading to over-heating of the cables and potentially causing secondary damage to other cables in the common enclosure (raceway, panel, junction, etc.). The Fire PRA model includes basic events to model failure of the credited 4.16KV buses. In fire scenarios where the DC control power and AC power cable are both affected, the basic event that represents failure of the upstream 4.16KV bus is set to true to model failure of the bus due to loss of breaker coordination for that bus. The Fire PRA model also fails the basic events for the components that are powered from the associated cables in the common raceway with the potentially over-heated 4.16KV load cable. The Fire PRA model was updated to set the basic events to true (failed) for the associated components powered by cables in the common raceway in fire scenarios where the DC control power was affected by the fire. The Fire PRA results bound the worst case failures by setting the basic events to true for failure of the entire upstream bus (which propagates to failure of all loads powered from that bus due to dependencies in the fault tree) and setting the basic events to true for failure of the components powered by cables that share a common enclosure with the affected load cables.
b) The fire-induced loss of DC control power to 4.16KV breaker protective circuitry could cause a loss of electrical coordination of the 4.16KV bus which ultimately could result in the source to the 4.16KV bus de-energizing the whole bus to clear the fault on the load that lost breaker coordination with the source breaker.* The Fire PRA model basic event to fail the entire bus is set to true in these fire scenarios where fire causes loss of coordination on the 4.16KV bus. The upstream bus is modeled to fail and all loads powered by the upstream bus are failed by loss of power from the bus in the Fire PRA model. Fire-induced loss of electrical coordination is assumed to fail the entire electrical bus and all loads powered from the bus were assumed to lose power. This assumption is bounding because it encompasses the worst case failure mode that the entire Page 9 of 11
L-PI-16-067 NSPM bus is unavailable to provide power to any of the loads normally powered from this bus.
The fire-induced loss of DC control power to 4.16KV breaker protective circuitry could cause a loss of over-current protection which could allow the 4.16KV load cable to over-heat and cause secondary damage to other associated cables in the common enclosure. Failure of these associated cables is modeled in *the Fire PRA by setting their respective basic events for the components powered by these power cables to true in each fire scenario where the DC control power and AC power cables are affected.
c) The 4.16KV AC power cable could over-heat due to the loss of over-current protection and is postulated to cause secondary fire damage to other cables that share a common enclosure. All of the cables in the common enclosure (raceway, panel, junction, etc.) going back to the upstream 4.16KV bus are failed in the Fire PRA model by setting the basic events associated with the cables in the common raceway to true.
Fire PRA FAQ 13-0005 provides clarification on treatment of self-ignited cable tray fires. FAQ 13-0005 reviewed the events in the Fire Event Database related to self-ignited cable tray fires and found that two earliest events at San Onofre Nuclear Generating Station (SONGS) were outliers due to the significant (more than a year) pre-heating of the cables due to over-loading of the trays and under-rated cables. Aside from the two events at SONGS in 1968 (which should not recur due to changes in cable ampacity limits), the other 50 events in the Fire Events Database did not propagate beyond their ignition point and either self-extinguished or were manually extinguished. The method in FPRA FAQ 13-0005 is to postulate that the self-ignited or hot work initiated cable tray fires do not damage cables beyond the raceway of origin because this comports with fire events recorded and fire experiments.
Therefore, the impact of secondary fire damage due to 4.16KV cables over-heating and potentially damaging the other associated cables in the common enclosure (raceway) is bounding and there is no need to postulate secondary fire damage zone of influence beyond the common enclosure (raceway).
d)
- i. See response to parts a and b.
ii. See response to parts a and b.
iii. See response to part c.
iv. 4.16KV circuits are considered to be high energy circuits in the Fire PRA model. Many of the Fire Scenarios are "Full Compartment Burn" (FCB) scenarios which means that all targets in the fire compartment were Page 10 of 11
L-PI-16-067 NSPM Enclosure 1 assumed to be failed by all fire ignition sources in the compartment with no credit for fire modeling and no limit on the Zone of Influence. In non-FCB fire compartments with detailed fire modeling, the Zone of Influence of the initial fire ignition source is sufficiently large (10ft by 10ft and floor to ceiling in the vertical dimension) to encompass the Zone of Influence of a High Energy Arcing Fault (5 feet vertical and 3 feet horizontal) at the location of the faulted 4.16KV cable.
References
- 1. Generic Letter 81 Fire Protection Rule (45 FR 76602, November 19, 1980)
- 2. NFPA 805, Performance Based Standard for Fire Protection for Light Water Electric Generating Plants, 2001 Edition
- 3. NF-40107-2, Cable Tray System Ground Floor Plan Turbine Room Class I Area, Rev 76
- 4. NF-40112-1, Cable Tray System Ground Floor Plan AUX Bid G Unit 1, Rev 76 Licensee Identified Issue An error in the May 24, 2016 submittal was discovered in Enclosure 5, Attachment W, Fire PRA Insights. Specifically, in Table W-6, Unit 1 Fire Area Risk Summary, under Fire Area 8 (Turbine Building), the VFDR column should be revised to indicate "Yes" because there is a VFDR in Fire Area 8. There are no recovery actions and no delta risk in Fire Area 8 because this VFDR is resolved by modification and therefore, Fire Area 8 meets NFPA 805, Section 4.2.3.2, Deterministic Compliance. No change is required to Table W-7, Unit 2 Fire Area Risk Summary, because there are no VFDRs affecting Unit 2 in Fire Area 8 (Turbine Building).
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