NL-14-0733, Response to Request for Additional Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c) - NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants
| ML14147A368 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 05/23/2014 |
| From: | Pierce C Southern Co, Southern Nuclear Operating Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14147A366 | List: |
| References | |
| NL-14-0733, TAC ME9741, TAC ME9742 | |
| Download: ML14147A368 (23) | |
Text
WITHHOLD FROM PUBLIC Charles R. Pierce Southern Nuclear DISCLOSURE UNDER Regulatory Affairs Director Operating Company, Inc. 10 CFR 2.390.
40 Inverness Center Parkway Post Office Box 1295 Birmingham, Alabama 35201 Tel 205.992.7872 Fax 205.992.7601 SOUTHERNA COMPANY May 23, 2014 Docket Nos.: 50-348 NL-14-0733 50-364 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, D. C. 20555-0001 Joseph M. Farley Nuclear Plant Response to Request for Additional Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c) - NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Ladies and Gentlemen:
By letter dated September 25, 2012, the Southern Nuclear Operating Company (SNC) submitted a license amendment request (LAR) for Joseph M. Farley Units 1 and 2 (Ref. TAC NOS. ME9741 and ME9742). The proposed amendment requests the review and approval for adoption of a new fire protection licensing basis which complies with the requirements in Sections 50.48(a) and 50.48(c) to Title 10 to the Code of Federal Regulations (10 CFR), and the guidance in Regulatory Guide (RG) 1.205, Revision 1, Risk-Informed, Performance-Based Fire Protectionfor Existing Light-Water Nuclear PowerPlants.
By letter dated December 12, 2012, the Nuclear Regulatory Commission (NRC)
Staff requested supplemental information regarding the acceptance of the license amendment (Adams Accession No. ML12345A398). SNC provided the requested information by letter dated December 20, 2012. The NRC staff subsequently completed the acceptance review by letter dated January 24, 2013, (Adams Accession No. ML13022A158).
By letter dated July 8, 2013, the NRC Staff formally transmitted a request for additional information (RAI) related to the referenced license amendment. SNC's responses to these RAIs are being provided by three submittals. By letter dated September 16, 2013, SNC provided the first set of responses. By letter dated October 30, 2013, SNC provided the second set of responses and by letter dated November 12, 2013, SNC provided the remaining set of responses. SNC provided that supplemental responses would be provided for nine of the RAIs.
U.S. Nuclear Regulatory Commission NL-14-0733 Page 2 By letter dated March 28, 2014, the NRC Staff formally transmitted the second round of requests for additional information related to the referenced license amendment request. By letter dated April 23, 2014, SNC provided the 30 day response to the second round of RAIs. Enclosure 1 to this letter provides the 60 day response to six of the eight remaining eight RAIs. PRA RAIs 06.a.01 and PRA 35 will be provided when the composite effect of the quantification of the PRA model is completed as agreed during the conference call with Shawn Williams on May 20, 2014. The supplemental responses identified in enclosure 1 of SNC's letter dated April 23, 2014 will be included in the response to PRA RAI
- 35. Enclosure 2 provides a supplemental response to the Safe Shutdown Analysis RAI 10.01.
Attachment G, Recovery Actions Transition provides a revision to page 9 which adds component numbers. Attachment G contains sensitive information and should be withheld from public disclosure under 10 CFR 2.390.
The No Significant Hazards Consideration determination provided in the original submittal is not altered by the RAI responses provided herein.
If you have any questions, please contact Ken McElroy at (205) 992-7369.
Mr. C. R. Pierce states he is Regulatory Affairs Director of Southern Nuclear Operating Company, is authorized to execute this oath on behalf of Southern Nuclear Operating Company and, to the best of his knowledge and belief, the facts set forth in this letter are true and correct.
ýX./f.
Respectfully submitted, C. R. Pierce Regulatory Affairs Director CRP/jkb/lac Stoandsubscribedbeforeme this Z day of ,2014.
Notary Public f My commission expires: / 2-0 / 9'
U.S. Nuclear Regulatory Commission NL-14-0733 Page 2
Enclosures:
- 1. Response to Probabilistic Risk Assessment RAIs
- 2. Supplemental Response to Safe Shutdown Analysis RAI Attachments: 1. Revision to Recovery Actions Transition- Attachment G cc: Southern Nuclear Operating Company Mr. S. E. Kuczynski, Chairman, President & CEO Mr. D. G. Bost, Executive Vice President & Chief Nuclear Officer Ms. C. A. Gayheart, Vice President - Farley Mr. B. L. Ivey, Vice President - Regulatory Affairs Mr. D. R. Madison, Vice President - Fleet Operations Mr. B. J. Adams, Vice President - Engineering RTYPE: CFA04.054 U. S. Nuclear Regulatory Commission Mr. V. M. McCree, Regional Administrator Mr. S. A. Williams, NRR Project Manager - Farley Mr. P. K. Niebaum, Senior Resident Inspector - Farley Mr. J. R. Sowa, Resident Inspector - Farley Alabama Department of Public Health Dr. D. E. Williamson, State Health Officer
Joseph M. Farley Nuclear Plant Response to Request for Additional Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c)
NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Enclosure 1 Response to Probabilistic Risk Assessment RAIs
Enclosure 1 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 01.01 LAR Attachment V, Table V.2-2, provides the results of the electrical cabinet fire severity sensitivity analysis for Unit 1, also indicating similar results for Unit 2.
There, the base CDF rose from 5.24E-5/y to 7.05E-5/y, an increase of 1.81 E-5/y.
For A CDF, the base value rose from 8.80E-6/y to 1.03E-5/y, an increase of 1.50E-6/y. The analogous results for LERF and A LERF were as follows: (1) a LERF increase of 2.59E-6/y from 1.26E-6/y to 3.85E-6/y; (2) a A LERF increase of 9.90E-8/y from 4.14E-7/y to 5.13E-7/y. Subsequently, the LAR was supplemented by a sensitivity analysis which included the effect of removing credit for very early warning fire detection system (VEWFDS) in the main control room (MCR) in addition to the electrical cabinet fire severity adjustment. The results were as follows: (1) CDF now rose only 1.41 E-5/y (vs. the previous 1.81 E-5/y); (2) A CDF now rose only 1.18E-6/y (vs. the previous 1.50E-6/y); (3) LERF now rose more by 6.28E-6/y (vs. the previous 2.59E-6/y); (4) A LERF now rose more by 2.88E-7/y (vs. the previous 9.90E-8/y).
In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019), as justification for the smaller increase for CDF and A CDF with credit for both VEWFDS and electrical cabinet severity adjustment removed, the licensee indicated via Table 1 that, in addition to removing credit for VEWFDS in the MCR, the following additional refinements were now included: (1) refined main control board (MCB) fire scenarios (via App. L of NUREG/CR-6850); (2) more realistic probabilities for HGLs; (3) refined circuit analysis for selected fire scenarios; (4) correction to anomalies in fire ignition frequencies for selected fire scenarios. As a result, the CDF and A CDF increase for removing both VEWFDS and electrical cabinet factor credit were actually less than prior to removal of the VEWFDS credit alone. While the licensee's explanation is sound for these metrics, it remains unclear as to why the LERF and A LERF increases do not display the same trend as CDF and ACDF. If the CDF and A CDF showed a smaller increase with the additional refinements, why did not the LERF and A LERF as well? Explain why the increases in LERF and A LERF after removal of the VEWFDS credit and addition of the four refinements trended upward vs. the downward trend for the CDF and A CDF increases.
RESPONSE
During the review of the sensitivity and subsequent model refinement, more attention was placed on the change in risk associated with CDF rather than LERF. The reason for this is that the LERF risk and delta risk were much lower than the acceptable limits stated in RG 1.174. Since CDF was much closer to these limits, more time was spent reviewing the scenarios that were contributing to the change in CDF risk. Furthermore, as a result of this RAI and others, the electrical cabinet severity factor and credit for incipient detection in the MCB were removed from the baseline model altogether. The updated plant risk and delta risk results will be provided with the supplemental response for PRA RAI 35.
El-1
Enclosure 1 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 06.a.01 In a letter dated November 12, 2013 (ADAMS Accession No. ML13318A027) the licensee responded to PRA RAI 06(a) and stated that section V.2.2 of the LAR provides the details of the sensitivity analysis related to the bins that have an alpha that is less than or equal to one. Indicate ifthe acceptance guidelines of Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk- Informed Decisions on Plant-Specific Changes to the Licensing Basis," may be exceeded when this sensitivity study for those bins with an alpha less than or equal to 1 is applied to the integrated study of PRA RAI 35 (see below). If these guidelines may be exceeded, provide a description of fire protection or other measures that can be taken to provide additional defense in depth (DID) (see FAQ 08-0048).
RESPONSE
The types of scenarios and the location of the rooms that have the highest contribution of risk as it relates to this sensitivity are located in rooms that contain electrical cabinets. These types of ignition sources are most predominant throughout the plant and typically have the largest ZOI and therefore impacts the most targets. The types of rooms that contain these types of ignition sources are switchgear, electrical penetration rooms, and general Auxiliary building hallways.
The rooms that have the increased risk and delta risk have defined Defense in Depth actions in place, including safety margin evaluations. These are included in the Fire Risk Evaluations for the specific fire areas.
The updated base risk results includes the composite effect from the RAI responses referenced in PRA RAI 35, with the exception of the bin 15 fire frequency discussed in our previous response to PRA RAI 28.k, which has been judged to have insignificant impact on the risk results. Updated total plant, Fire PRA, and delta risk values will be submitted in PRA RAI 35 under separate cover once completed.
El -2
Enclosure 1 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 16.a.01 In a letter dated October 30, 2013 (ADAMS Accession No. ML13305A105), the licensee responded to PRA RAI 16.a and partially addressed some of the criteria for assuming damage within MCR panels to be limited to the initiating panel, namely the presence of no openings and a double wall with an air gap. However, Appendix S of NUREG/CR-6850 also states that there be no sensitive electrical equipment in the adjacent cabinet (or else such equipment to have already been "qualified" above 82C), even with the double wall with air gap. Otherwise damage to such equipment should be postulated. Explain whether these additional criteria are met or not. If the latter, explain how damage is modeled or, if not, the basis for assuming no damage. (Also see PRA RAI 33.a.01.)
RESPONSE
Damage to adjacent panels in the Main Control Room (MCR) area is postulated to be limited to the initiating panel provided there are no openings and the panels have a double steel wall and an air gap between them as described in the response to PRA RAI 16.a based on the guidelines provided in Appendix S of NUREG/CR-6850. If there are sensitive electronics within the adjacent panels, Appendix S of NUREG/CR-6850 notes that a damage time of ten minutes should be assumed, based on a comparison of the interior temperatures of the test panels as compared to the damage criteria for sensitive electronics provided in Appendix H of NUREG/CR-6850. Appendix S also notes that damage to sensitive electronics can be prevented if the fire is extinguished and the cabinet is cooled before ten minutes.
Sensitive electronics are present in most electrical panels in the Farley MCR within the main control board (MCB) area, as well as the back panel area. As such, the additional criteria described in Appendix S of NUREG/CR-6850 relating to sensitive electronics are applicable.
Main Control Board:
The MCBs also contain sensitive electronics; however, because there is no internal separation between panels, a fire is assumed to propagate and damage adjacent targets in accordance with the methodology described in Appendix L of NUREG/CR-6850 (see response to RAI 33.a.01).
Non-Main Control Board Panels in the control room:
The Fire PRA does not postulate damage to sensitive electronics in adjacent panels that are separated by double wall construction with an air gap in the MCR (which includes the control room and the back panel area of the control room).
The basis for this treatment is that the MCR is continuously staffed and the control room operators are trained to quickly detect and mitigate the effects of a fire that may occur in the electrical equipment. Per Appendix S of NUREG/CR-6850, damage to sensitive electronics in an adjacent panel can be prevented if the fire is extinguished and the exposing cabinet is cooled within ten minutes.
Operations guidance for control room panel fires will be revised to emphasize the need to evaluate the initial fire and to open the panel doors if the potential exists for damage/overheating in an adjacent panel. In this case, the actions of the operators are credited for interrupting the fire growth (and therefore the peak El -3
Enclosure 1 Response to Probabilistic Risk Assessment RAIs internal temperature) and exposure hazards of the panel ignited, including opening the panel doors and cooling the adjacent panels. As noted in Appendix S of NUREG/CR-6850, the full scale fire tests documented in NUREG/CR-4527 indicated that open door panel fires did not cause the conditions in an adjacent panel to exceed the sensitive electronic temperature threshold if the panel door is open. Further, the actions of the operators are not necessarily postulated to extinguish the fire in the exposing panel; rather, they are assumed to delay or prevent further fire growth and propagation to adjacent panels until the time at which the fire brigade arrives and fully extinguishes the fire. A new implementation item has been added to the LAR Attachment S, Table S-3 to address the additional Operations guidance, and is attached to this RAI response.
This approach is further supported by the control room temperature data associated with the control room abandonment analysis. As described in Report 0005-0030-003-001, Rev. 1 ("Evaluation of the Control Room Abandonment Times at the Farley Nuclear Power Plant"), the abandonment time for the 9 8 'h percentile heat release rate closed panel non-propagating (separated by double wall air gap) fire scenarios is between 13 - 17 minutes, depending on the ventilation configuration. The hot gas layer temperature in the control room, which is indicative of the hot gas layer temperature in the back panel area, is approximately 60 0 C (140 0 F) at this time, which is below the damage threshold of 6500 (149°F) for sensitive electronics as described in Appendix H of NUREG/CR-6850. This ensures sufficient time for operator action to control the fire prior to control room abandonment. Ultimately the hot gas layer temperatures reach 80 -
90°C (176°F - 194°F) and the hot gas layer depth descends to the floor; however, the fire model does not credit the actions the operators would take to mitigate the fire prior to abandonment and are thus conservatively biased. In addition, the maximum temperature is comparable to the damage temperature threshold of 8200 (1 80 0 F) assumed in Appendix S of NUREG/CR-6850 for damage to sensitive electronics in adjacent panels.
As previously noted, the assumption that sensitive electronics are not damaged in adjacent panels separated by a double steel wall with an air gap in the control room area is considered applicable to panels in the main control board area as well as the back panel area. The basis for this assumption is that the control room is continuously staffed and that operators are trained to mitigate the effects of a fire using fire extinguishers and other fire protection equipment located in the general area. Note that the back panel area may not be continuously staffed, but the electrical panels in this area are within the control room HVAC envelope. This means that smoke generated from fires at these panels will be identified in their early stages and early initiation of fire mitigation actions by the operator and the fire brigade, including efforts to limit the fire to the initiating source and to cool any adjacent panels, will be implemented.
E1-4
Enclosure 1 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 21.a.01 In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019),
the licensee responded to PRA RAI 21 .a and confirmed that the three severity factors, 5.02E-4, 4.84E-4 and 0.00158, do not derive from Figure L-1 in NUREG/CR-6850 but are specifically calculated based on the type of ignition source, scenario location and abandonment time for the MCR abandonment analysis. The three severity factors correspond to the abandonment probabilities for transient ignition sources, equipment room fixed ignition sources and MCR fixed ignition sources, respectively. Provide a discussion of the derivation of these factors, including their bases, e.g., as given in Section 13.2.1 of the Farley Scenario Development Report, PRA-BC-1 1-014, and Section 6 of Units 1 and 2 Control Room Abandonment Times at the Joseph M. Farley Nuclear Plant, Rev 0.
RESPONSE
Overview The computation of the severity factors (control room abandonment probabilities) is derived from the analysis of the control room abandonment time for each bin of a heat release rate distribution (defined in Report 0005-0030-003-001 Rev. 1, "Evaluation of Control Room Abandonment Times at the Farley Nuclear Power Plant", based on the Heat Release Rate probability distributions defined in NUREG/CR-6850, Appendix E). A non-suppression probability is determined for each bin based on the time to abandonment calculated for that heat release rate bin from NUREG/CR-6850, Supplement 1, Chapter 14. The severity factor for each heat release rate bin is multiplied by the non-suppression probability corresponding to the time to abandonment for that heat release rate bin to obtain the probability of abandonment for that heat release rate bin. The probability by heat release rate bin is summed for all heat release rate bins to calculate the probability of abandonment for the particular control room fire configuration.
Details of the Severity Factor Derivation Methodology The analysis as described in PRA-BC-F-1 1-014 Rev. 5 has been updated to account for the responses to other RAIs. The MCR Abandonment calculation has been updated, Report 0005-0030-003-001, Rev. 1, "Evaluation of Control Room Abandonment Times at the Farley Nuclear Power Plant". This report provides the time at which the operators would abandon the control room given the abandonment temperature or visibility conditions defined in NUREG/CR-6850.
The following is the updated severity factor derivation for the MCR Abandonment scenarios at Farley.
The analysis evaluates cases for the following Heat Release Rates:
- Single cable bundle: closed electrical panel thermoset cable
- Multiple cable bundle: closed electrical panel thermoset cable
" Multiple cable bundle: closed electrical panel thermoset cable. The fire propagates to two adjacent panels after 10 minutes and two additional panels after 20 minutes E1-5
Enclosure 1 Response to Probabilistic Risk Assessment RAIs
- Transient fire in an open location
" Transient fire in a wall configuration
" Transient fire in a corner configuration It is assumed that the MCR will be ventilated by opening at least one door within 15 minutes. In the development of the analysis for the MCR Abandonment, some of the panels located in the control room area were opened in support of this analysis but many of the panels were not. For this reason all panels will be considered to consist of multiple cable bundles.
The FNP main control room area includes two distinct areas: the MCB panel area and the back panels or "equipment area." These areas are somewhat isolated by panels, partitions, doors and walls such that gases/smoke transport is limited between the two spaces. While the main control room is continuously occupied the equipment area is not. For this reason, there are three separate non-suppression probabilities used in the abandonment calculation. A lookup of these times in Chapter 14 of Supplement 1 to NUREG/CR-6850, for Control Room Fires, electrical and transient fires, provides a non-suppression value.
Abandonment cases are prepared for these areas/configurations individually; however the HVAC system is the same for both rooms and will therefore consider both rooms to be part of the main control room analysis.
Given the expected rapid fire brigade response due to continuous manning of the control room, the 15 minute timeframe for opening a door to the control room is considered appropriate. The 15 minute time frame is further justified in response to RAI FM 01a.
The abandonment times are documented in Report 0005-0030-003-001, Rev. 1, "Evaluation of Control Room Abandonment Times at the Farley Nuclear Power Plant". These times are used along with the non-suppression probabilities of Supplement 1 to NUREG/CR-6850 (ext) to determine a cumulative case NSP for all heat release rates. Note that the MCB electrical cabinet severity factors are the same as general electrical cabinets consisting of multiple cable bundles with propagating fires.
The computed NSPs are used to calculate a probability of abandonment that represents the probability that a given fire will cause the operators to abandon the control room. There were 13 abandonment severity factors calculated: five for the MCB panel area and eight for the Unit 1 & 2 equipment rooms and the outlying areas within Fire Area 044. In addition to the HRR cases identified above, there are also three HVAC configurations analyzed:
0 HVAC not operating
- HVAC operating normally
- HVAC in purge mode For a given fire scenario, one of the above configurations will be true.
The following severity factor types were calculated using the above HRR cases and HVAC configurations:
-MCREP.PROP.HVAC-OP: Main Control Room Electrical Panel Propagating, HVAC operating normally E1-6
Enclosure 1 Response to Probabilistic Risk Assessment RAIs
-MCREp-PROP-HVAC.INOP: Main Control Room Electrical Panel Propagating, HVAC not operating
-MCREP.PROP-HVAC-PURGE: Main Control Room Electrical Panel Propagating, HVAC in purge mode
-MCRTRAN-HVAC-OP: Main Control Room Transient, HVAC operating normally
-MCRTRAN-HVAC-INOP: Main Control Room Transient, HVAC not operating
-EREP.NOPROP.HVAC.OP: Equipment Room Electrical Panel Non-Propagating, HVAC operating normally
-EREP.NOPROP.HVAC.INOP: Equipment Room Electrical Panel Non-Propagating, HVAC not operating
-EREP-NOPROP-HVAC.PURGE: Equipment Room Electrical Panel Non-Propagating, HVAC in purge mode
-EREP.PROP-HVAC-OP: Equipment Room Electrical Panel Propagating, HVAC operating normally "EREP-PROP-HVAC-INOP: Equipment Room Electrical Panel Propagating, HVAC not operating
-EREP.PROP.HVAC.PURGE: Equipment Room Electrical Panel Propagating, HVAC in purge mode
-ERTRAN-HVAC-OP: Equipment Room Transient, HVAC operating normally
-ERTRAN-HVAC-INOP: Equipment Room Transient, HVAC not operating Where a propagating fire in the equipment room is defined as an ignition source involving two or more cubicles that are open to each other.
E1-7
Enclosure 1 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 29.01 In a letter dated September 16, 2013 (ADAMS Accession No. ML14038A019),
the licensee responded to PRA RAI 29 and indicated that 22 supporting requirements (SRs) fail to meet Capability Category (CC) II, 17 more than the staff was able to determine by review of LAR Attachment V, Table V-I. The licensee response refers to dispositions in LAR Attachment V, Table V-i, which, while indicating how the licensee addressed the related findings and observations (F&Os), do not specifically explain why failing to meet CC-Il is acceptable for transition under NFPA 805. Provide Table V-2 which explains the rationale for acceptability of less than CC-Il satisfaction for all 22 SRs.
RESPONSE
The response to this RAI was provided in SNC letter dated April 23, 2014.
E1-8
Enclosure 1 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 33.a.01 In a letter dated October 30, 2013 (ADAMS Accession No. ML13305A105), the licensee responded to PRA RAI 33.a and referenced PRA RAI 16.a. However, neither of the responses to PRA RAI 16.a or PRA RAI 33 discussed the timing for detection and manual suppression prior to fire spread to adjacent cabinets.
Furthermore, the response to PRA RAI 33.a indicates that all MCB panels are physically open to one another. Discuss the basis for assuming rapid enough detection and manual suppression prior to fire spread into the adjacent cabinet.
RESPONSE
The MCB panel fires are postulated in a manner in which a fire starts at a given point on the board and then propagates to the adjacent panel sections, regardless of the panel name or section. Damage to all targets in the scenario occurs at t=0. Appendix L of NUREG/CR-6850 is credited from a manual non-suppression standpoint. The MCB sections are open to each other and the fire is considered to propagate to the adjacent sections of the MCB. The analysis does not assume rapid enough detection and manual suppression prior to fire spread into the adjacent cabinet (or section panels) in MCB. No specific credit for detection and manual suppression is taken beyond that inherent in the NUREG/CR-6850, Appendix L Probability of Target Damage versus Distance curves.
For discussion of treatment of non-MCB electrical cabinets in the main control room, refer to the response to RAI PRA 16.a.01.
El -9
Enclosure 1 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 33.c.01 In a letter dated November 12, 2013 (ADAMS Accession No. ML13318A027), the licensee responded to PRA RAI 33.c and indicated an intent to revise its MCR abandonment calculation as follows:
"The CCDP for the abandonment scenario is based on failure of all actions in the control room. [A] conservative basis was used for determining the abandonment CCDP based on the calculated CCDP associated with panel damage and failure of the MCR actions. The intent of this criteria is to ensure that the abandonment CCDP is an appropriate bounding value given that, shutting down the plant from outside the control room has an inherently higher risk associated with it."
These criteria are presented as (1) using conditional core damage probability (CCDP) = 0.1 if FRANC calculates a CCDP < 0.001, (2) using CCDP = 0.2 if FRANC calculates a CCDP between 0.001 and 0.1, and (3) using 1.0 if FRANC calculates a CCDP > 0.1. These FRANC-calculated CCDPs are based on both MCB panel damage and failure of human actions in the MCR. Clarify how these human actions were quantified, including any detrimental effects (increased failure probabilities) due to fire effects in the MCR. If screening or other bounding values were used, specify their bases, e.g., screening/scoping approach from NUREG-1921, "Fire Human Reliability Analysis Guidelines" (or equivalent).
RESPONSE
The evaluation of the plant risk given abandonment of the main control room (MCR) involves a two-step process for assignment of the conditional core damage probability (CCDP) values. The first step in this process involves the use of FRANC to determine the characteristic CCDP for each of the fire scenarios treated in the MCR. There are approximately 100 scenarios that are individually evaluated for abandonment related risk contribution. Each of these scenarios are first quantified in FRANC assuming all MCR actions are 'failed' - set to TRUE.
Because the underlying risk model that is used in the FRANC quantification does not address the impact of the shift of command and control from the control room to the Hot Shutdown Panel for an abandonment scenario, this quantification effectively provides insights and characterization of the direct fire induced impacts given the originating fire. Based on this CCDP result, the second step in the quantification process determines the corresponding abandonment related CCDP for each of these individual scenarios by manually resetting the CCDP to one of the values provided in the original RAI response (0.10, 0.20, or 1.0), to account for the shift of command and control to the hot shutdown panel.
The assigned CCDP value (0.10, 0.20, or 1.0) is intended to reflect the combination of human reliability (HEP) and hardware failures. As provided in the original RAI response, a CCDP value of 0.10 is assigned only if the corresponding CCDP is less than 1 E-3. Given that such a CCDP value would have been determined without any credit for in-control room human actions (all MCR HFEs are set to TRUE), the resultant plant trip would be uncomplicated and non-fire affected plant systems would continue to operate until they are manually secured as part of the prescribed abandonment procedure steps. It is noted that:
El-10
Enclosure 1 Response to Probabilistic Risk Assessment RAIs
- The quantification provides no credit for primary bleed and feed since the associated HFE would have been set to TRUE
" Any consequential loss of coolant event would have yielded a CCDP of 1.0 as the human action to transition to recirculation from the containment sump would have also been set to TRUE.
- These considerations would infer that the time window for completion of the abandonment related action is relatively long.
- A resultant CCDP of less than 1 E-3 indicates that offsite power is available.
In this instance, the resultant assigned CCDP for the abandonment case of 0.10 easily bounds the sum of the abandonment HEP and hardware failure probability.
It is noted that a comparable result could be produced using the scoping guidance in NUREG-1921 and combining that result with the associated hardware failures. Although not formally evaluated and included in the analysis documentation, Figure 5-5 and Table 5-5 of NUREG-1 921 would suggest that an HEP of 0.04 would be appropriate. Since offsite power is available in this set of scenarios, the corresponding hardware failures would be expected to yield a value much less than 0.06 (0.1 - 0.04). Based on this comparison, the use of a 0.10 value as the characteristic abandonment CCDP for those instances where the FRANC quantification yielded a CCDP of less than 1 E-3 is considered to be very conservative.
A CCDP value of 1.0 would be assigned if the FRANC generated CCDP is 0.10 or greater. It is noted that any induced or consequential loss of coolant event would result in a FRANC generated CCDP of 1.0. This is because the dependent HFE to initiate recirculation from the containment sump will have been set to TRUE. For these cases, no HEP treatment is necessary as the scenario is conservatively assumed to result in core damage.
The third possible assigned CCDP of 0.20 is used if the FRANC generated CCDP is 1 E-3 or greater but less than 0.10. As noted earlier, the FRANC calculated CCDP is performed without any credit for in-control room human actions (all MCR HFEs are set to TRUE) and non-fire affected plant systems would continue to operate until they are manually secured as part of the prescribed abandonment procedure steps. This FRANC quantification provides no credit for primary bleed and feed since the associated HFE would have been set to TRUE. In addition, any consequential loss of coolant event would have yielded a CCDP of 1.0 as the human action to transition to recirculation from the containment sump would have also been set to TRUE. These considerations would infer that the time window for completion of the abandonment related action is relatively long.
The scoping HEP of 0.04 from NUREG-1921 for the case when a CCDP of 0.10 is used, would also be applicable in this case. However, the hardware related failures involves a much wider range of possible values. An upper bound value of 0.10 would therefore be applicable in this case. When these two terms are combined, the result of 0.14 is still below the assigned value of 0.20. As such, the treatment remains conservative and bounding.
El-11
Enclosure 1 Response to Probabilistic Risk Assessment RAIs The process that was used in the Farley fire PRA for the treatment of MCR abandonment appropriately considered the combination of human error probability and equipment hardware failure likelihood. Although a formal detailed HRA was not performed nor was NUREG-1 921 specifically applied for the MCR abandonment scenarios, the results are expected to bound the results from a detailed HRA. As such, the resultant risk estimates are considered to have been adequately characterized for the purposes of this NFPA 805 application.
E1-12
Enclosure 1 Response to Probabilistic Risk Assessment RAIs Farley PRA RAI 35 Section 2.4.3.3 of the NFPA 805 standard incorporated by reference into 10 CFR 50.48(c) states that the PSA approach, methods, and data shall be acceptable to the AHJ, which is the NRC. Regulatory Guide (RG) 1.205, "Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants," identifies NUREG/CR-6850 as documenting a methodology for conducting a Fire PRA (FPRA) and endorses, with exceptions and clarifications, NEI 04-02, "Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c)," Rev. 2, as providing methods acceptable to the staff for adopting a fire protection program consistent with NFPA 805. RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities," describes a peer review process utilizing an associated ASME/ANS standard (currently ASME/ANS-RA-Sa-2009, "Addenda to ASME/ANS RA-S-2008, Standard for Leveil/Large Early Release Frequency Probabilistic Risk Assessment for Nuclear Power Plant Applications") as one acceptable approach for determining the technical adequacy of the PRA once acceptable consensus approaches or models have been established. In a letter dated July 12, 2006 to NEI (ADAMS Accession No. ML061660105), the NRC established the ongoing FAQ process where official agency positions regarding acceptable methods can be documented until they can be included in revisions to RG 1.205 or NEI 04-02.
Section 2.4.4.1 of NFPA 805 states that the change in public health risk arising from transition from the current fire protection program to an NFPA 805 based program, and all future plant changes to the program, shall be acceptable to the AHJ, which is the NRC. RG 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," provides quantitative guidelines on CDF and LERF and identifies acceptable changes to these frequencies that result from proposed changes to the plant's licensing basis and describes a general framework to determine the acceptability of risk-informed changes.
As stated on page B-1 of Appendix B of PRA-BC-F-1 1-004, "Fire PRA Logic Model," the new Westinghouse Shutdown Shield (SDS) was installed in fall 2010.
The internal events PRA (IEPRA), upon which the FPRA is based, takes credit for the SDS (failure rate of 0.0271/demand), limiting the leakage rate to 2 gpm where the faces of the SDS seal components remain in contact. The assumed leakage rate is increased to 19 gpm if the SDS actuates but the pump shaft continues to rotate if not tripped in a timely manner. Finally, if the SDS does not actuate at all, "existing" (Westinghouse Owners Group (WOG) 2000 or Rhodes Model) seal model leakage rates are applied. Given the July 26, 2013, 10 CFR Part 21 notification by Westinghouse concerning defects with the SDS performance, provide a sensitivity evaluation that removes all credit for the SDS package, including both probability and consequences as appropriate. Provide revised estimates of CDF, LERF, A CDF and A LERF, including non-fire hazards for CDF and LERF, as a result of removal of this credit. Should this result in any changes to conclusions regarding the transition satisfying RG 1.174 risk/A risk guidelines, address any changes that will be made to accommodate this.
When performing this analysis, include the composite effect from all previous re-evaluations, including any synergistic effects, specifically including the following:
E1-13
Enclosure 1 Response to Probabilistic Risk Assessment RAIs
- a. From the LAR and the December 20, 2012 LAR Supplement, sensitivities related to the electrical cabinet fire severity method (Section V.2.1) and use of control power transformer (CPT)
(Section V.2.3; also response to PRA RAI 08.a).
- b. From the RAI Responses dated September 16, 2013 (ADAMS Accession No. ML14038A019):
- i. PRA RAI 01 .a - Removal of credit for VEWFDS in the MCR (also PRA RAI 01.01) ii. PRA RAI 15.a - Revised seismic CDF based on 2008 USGS data iii. PRA RAI 28.k - Validity of current Ignition Bin 15 fire frequencies
- c. From the RAI Responses dated November 12, 2013 (ADAMS Accession No. ML13318A027):
- i. PRA RAI 07.e - Use of 0.1 CCDP for MCR Abandonment ii. PRA RAI 17.d- Turbine Building Collapse iii. PRA RAI 33.c - Revised MCR Abandonment analysis (also RAI PRA 33.c.01)
RESPONSE
Reactor Coolant Pump (RCP) Seal Model The Farley Internal Events PRA model and Fire PRA model for the NFPA 805 License Amendment Request (LAR) includes credit for the Westinghouse shutdown seal (SDS) as outlined in Revision 1 to Pressurized Water Reactor Owners Group Topical Report WCAP-1 7100-P/NP, "PRA Model for the Westinghouse Shut Down Seal." SNC is aware of the operating experience issues with the Westinghouse shutdown seals which resulted in the 10 CFR Part 21 notification, and how that experience does not match the SDS performance credited in the Farley model which is based on the topical report.
In response to the issues which resulted in the 10 CFR Part 21 notification, Westinghouse has developed a new version of the SDS which has undergone a rigorous testing program. SNC plans to install this next generation shutdown seal from Westinghouse as a corrective action resolution for the issues identified in the Part 21 notification. Installation of this next generation shutdown seal in each FNP RCP will be added as an installation item to Attachment S.
The current modeling of the SDS performance in the PRA for NFPA 805 reflects the original PRA modeling guidance for the Westinghouse shutdown seal (WCAP-171 00-P/NP Rev. 1). PRA modeling guidance specific to this next generation (Generation Ill) SDS is in the development and review process for submittal to the NRC. SNC expects the RCP SDS risk benefit using this model to be consistent with the benefit currently being determined using the previous WCAP-1 71 00-P/NP, Rev. 1 model. SNC will monitor Westinghouse and PWR Owners Group efforts relative to issuance of this modeling guidance and NRC review.
As stated in the License Amendment Request Section 4.8.2, "Plant Modifications and Items to be Completed During the Implementation," following the installation of modifications and the as-built installation details, additional refinements El-14
Enclosure 1 Response to Probabilistic Risk Assessment RAIs surrounding the modification may need to be incorporated into the Fire PRA model, and the Fire PRA will verify the validity of the reported change-in-risk on as-built conditions after the modification is completed. This is captured under implementation item 30 in Table S-3, "Implementation Items," and is applicable to the installation of this next generation shutdown seal in each FNP RCP. The model will be revised to incorporate any new data or modeling requirements and further refinements or modifications will be addressed as necessary.
Composite Effect from all Previous Re-Evaluations a) From LAR and the December 20, 2012 LAR Supplement:
i) NFPA 805 LAR Submittal Section V.2.1 - The electrical cabinet fire severity method is no longer used in the Farley Fire PRA.
ii) NFPA 805 LAR Submittal Section V.2.3 and RAI PRA 08(a) (ML12359A051) - The hot short-induced spurious operation conditional likelihood estimates used in the development of the Fire PRA has been updated to reflect the guidance provided in ML14086A165, "Supplemental Interim Technical Guidance on Fire-Induced Circuit Failure Mode Likelihood Analysis". This guidance has been incorporated into the base model. The original response to RAI PRA 08(a) indicated that this issue would be addressed via a sensitivity analysis. This response supersedes the previous response and will incorporate this change into the baseline Fire PRA.
b) From RAI Responses dated September 16, 2013:
i) PRA RAI 01 .a - Credit for VEWFDS has been eliminated from the control room main control board area (fire zone 044). Credit for VEWFDS modifications is taken for electrical panels located behind the main control room vertical boards within the control room envelope. (See RAI 01.a response, including reference to supporting document RBA 13-006-F, Sensitivity Study of VEWFDS credited in the MCR, dated August 9, 2013) ii) PRA RAI 15.a - The updated seismic contribution to the total plant risk is being used.
iii) PRA RAI 28.k - The number of vertical sections of 1 F Switchgear in Room 0335 was corrected to 15 from 16 and sensitivity analysis was performed for the discrepancy of total plant count of electrical cabinets. The sensitivity analysis provided in the RAI 28.k submittal demonstrated that the associated discrepancy had an insignificant impact on the Fire PRA. Therefore, this change is not included in the updated fire PRA risk evaluation in support of the NFPA 805 LAR.
This change will be incorporated in conjunction with other future changes as appropriate.
c) For RAI Responses dated November 12, 2013:
i) PRA RAI 07.e - The methodology of applying a 0.1 CCDP for all control room abandonment scenarios is no longer credited in the Fire PRA. See also PRA RAI 33.c response in the November 12, 2013 submittal as well as the response to PRA RAI 33.c.01 accompanying this RAI response submittal.
El-15
Enclosure 1 Response to Probabilistic Risk Assessment RAIs ii) PRA RAI 17.d - The ignition frequency and suppression terms for the Turbine Building Collapse scenarios have been updated.
iii) PRA RAI 33.c - See response to PRA RAI 33.c.01 for a discussion and clarification of the methodology used for the MCR Abandonment analysis.
The updated base risk results includes the composite effect from the RAI responses referenced above, with the exception of PRA RAI 28.k above, which has been judged to have insignificant impact on the risk results.
Updated total plant, Fire PRA, and delta risk values will be submitted under separate cover once completed. This will include a revised NFPA 805 LAR, Attachment W as well as updates to Attachments C, G, S, and V.
El-16
Joseph M. Farley Nuclear Plant Response to Request for Additional Information Regarding License Amendment Request for Transition to 10 CFR 50.48(c)
NFPA 805 Performance Based Standard for Fire Protection for Light Water Reactor Generating Plants Enclosure 2 Supplemental Response to Safe Shutdown Analysis RAI
Enclosure 2 Supplemental Response to Safe Shutdown Analysis RAI Farley Safe Shutdown Analysis (SSA) Request for Additional Information (RAI) 10.01 In a letter dated October 30, 2013 (Agencywide Documents Access and Management System (ADAMS) Accession No. ML13305A105), the licensee responded to SSA RAI 10 and indicated that no recovery actions (RAs) were omitted from license amendment request (LAR) Attachment G, and that the correlation of RAs to variances from deterministic requirements (VFDRs) were provided in LAR Attachment G, Table G-1, of the LAR supplement dated December 20, 2012 (ADAMS Accession No. ML12359A050).
However, the NRC staff is providing the following examples of inconsistences between: LAR Attachment C, Table C-1 and LAR Attachment G, Table G-1; with the LAR supplement and the RAI responses:
- a. The LAR Attachment G, Table G-1 submitted on December 20, 2012, is missing many fire areas compared to the original LAR Attachment G, Table G-1 provided in September, 2012 (e.g., most of U2-040, U2 2-040, U2 2-041, U2 2-075, and U2 2-076... up to U2-2-021).
- b. Components identified in the new LAR Attachment G, Table G-1 don't correspond to VFDRs in LAR Attachment C, Table C-1 (e.g.,
Q1P16V0530 and Q1P16V0593).
- c. Components corresponding to VFDRs in the LAR Attachment C, Table C-1 are not identified in the new LAR Attachment G, Table G-1 (e.g., VFDRs: U1-044-PCS-040 and U1-1-040-PCS-186), and there are potential duplicate VFDRs (e.g., U1-1-040-PCS-145 &
146, U1-044-PCS-127 & 128 and U2-044-PCS-079 & 155)
- d. A partial LAR Attachment G, Table G-1 was submitted with the RAI responses on November 12, 2013 (ADAMS Accession No. ML13318A027), which includes LAR pages G-9 through G-
- 26. This table has corrections and multiple entries needing to be removed as duplicates. However, this LAR Attachment G, Table G-1 still includes several components (e.g., OP-RECOV-XXXX) that are not in LAR Attachment C, Table C-1.
Provide LAR Attachment G, Table G-1 and LAR Attachment C, Table C-1 that are up-to-date and correlate accordingly.
SUPPLEMENTAL RESPONSE:
- b. The component(s) identified in LAR Attachment G, Table G-1 are component(s) made available by application of recovery action(s). In cases where a VFDR requires a recovery action (to satisfy the risk or defense-in-depth criteria of NFPA 805, Ch. 4.2.4) alternate or redundant components are restored to meet the performance criteria. These recovered components are identified in Attachment G, Table G-1. The components associated with the VFDR are documented in Attachment C, Table C-1. The previous LAR Att. G tables that were submitted did not include the recovery action type to VFDR to recovered component E2-1
Enclosure 2 Supplemental Response to Safe Shutdown Analysis RAI correlation; however, to facilitate more complete mapping this has been revised with the updated Attachment G included in this response.
To specifically address the examples in the RAI the following detail is provided:
SE(I.0t51 W601-008 Atijacti,mntl FRt foF Unit I (Aitoewate Sbutdowni) ý ue Af t*a 044 Repred MU Recovery Action C91om1ome"t1 Cogin of t Omoti. Wt-R Reovera cavi Adcawer w*visu OIPi6smis Train a OW to Diese &i"V 114511511 Remrove power ftm senic wate veless Q1P116W22 UWi 2 SW $upl 10o EDO 18 asOC V cooi fat EDO lB by openirg ted 01PI1W830 Uri 2 SW Return ftm EDO tie supply dra*i tlu at MCC IT. Power 01PI6VW23 Unit 1SW SUPp to EDO 18 to these vae Isramoved Pio to starln B OIPIOVU3I1 WUi tSWR ftoMr EDOG 1B EDO to rO*ed pmtu aPan of lIe vaves 01IP16v0636 Tran 8 SW Rfetm from Diesel won EDG 18 Isp i serwce.
nacd 01P16V038 GO1PVW54 Train 8 sW Discierge to PoNd nSW to EDO Is not avaal" foM Unt 1t.
Train 8 SW DischaMe to River manwsW opw lhe Vto2 SW supply ad run vatv to EDO 18 Thes vaves so manual*
VOWed becaMe tI"ea nomalny dDsd These ac s ov-ue ta SW ispovded to the EDG tom t lest one of ft urit SW system Table G-1 (Componeft Recovered - Fire Area Associated VFORs Reference Nlot Necessarly WOUR Cofiponent) (Components Affcted by Fire Listed m At& C)
QIP16VO530 UI 044 145-151 SE-C051326701-O08 (FRE)
Table 6-1 (Component Recovered - Fire Area Associated VFORs Reference Not Necessarily VFDR Component) (Components Affected by Fire Listed in AtLC)
Q1P16V0593 U2 044 7,51-55,145,146 SE-C051326701-.08 (FRE)
St 0~r3A6?0 008r- Atnhm [RE $mUnwit ? (Afternate MShuldown) i mi A~a 0)44 Rer*4ud OR ReCOMeY Acuone compoOent U) componen Oescn1 t VFOR Roe"" "Con 086"I1116 QP6V051 Train sW to Diese auk** 7.51- Remove pmwr krn sv wate valves Q1P16VU9W thit 1SW SuPply to EDO 28 55, asmodoad wlbh cooiq tor EOG 28 by opening Ite Q0P16W" 3 Un I 9W RPeltur tm EDG 2B 145,146 spl dmcdi brears A MCC IP, 2DO, 2V md 02P16VU59 Wli 2 SW 91pl to EDG 28 2T. Power to th"ese vetoe is removed rIor to 02PI8V03 Wi 2 SW Redun trom EDO 28 staring the EDO to prevent spurious operaon of C2P16VO536 Train 8 SW Retu tom "th Desel vamves one EDG 291 piaced in sevice Q2P06O38 BDWO Q2P16V-545 TraMi 8 W Obdise to Pond If SW to EDO 2B is not avaal e eomUit 2, Train S8W DOidwe to River ,nar"Aly open oh Ut sw Upply mad end vaales to EDO 28. The vaes me mU* i uni rsysoem because thwey onarmaly dosed These acbon enans tha SW IsprovkWe to the EDO tOM at Weast one of fte WWls SW system, E2-2