ML14330A639
| ML14330A639 | |
| Person / Time | |
|---|---|
| Site: | Diablo Canyon  |
| Issue date: | 11/26/2014 |
| From: | Summy J Pacific Gas & Electric Co |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14332A120 | List: |
| References | |
| DCL-14-110, TAC MF2333, TAC MF2334 | |
| Download: ML14330A639 (44) | |
Text
Enclosure PG&E Letter DCL-14-110 However, issuance of NUREG/CR-7150, Volumes 1 and 2 renders DCPP's plant-specific assessment obsolete as the NUREG recommends screening out three-phase proper polarity hot shorts for all cases without any constraints or limitations.
Thus, the previous limitations in NUREG/CR-6850 for screening "High Consequence" components (i.e., ungrounded system and thermoplastic insulation) no longer exist, as discussed below. Subsequent to submittal of the DCPP LAR, NUREG/CR-7150 Volumes 1 and 2 have been issued. These two reports contain additional technical guidance related to consideration of three-phase proper polarity hot shorts. Specifically, the updated guidance recommends screening three-phase proper polarity hot shorts without any conditions or limitations.
This updated guidance is based on test results from the recent cable fire tests (CAROL-FIRE, JACQUE-FIRE), which provide further confirmation that fire-induced cable failure characteristics are not conducive to three-phase proper polarity hot shorts under any conditions.
NUREG/CR-7150, Volumes 1 and 2 classify the spurious operation of a three-phase AC motor due to proper polarity hot shorts on phase power cabling as "incredible." NUREG/CR-7150 recommends that cable failure modes classified as "incredible" not be considered by the Fire PRA, as established by the definition of "incredible" in Section 2.2.2 of NUREG/CR-7150, Volume 2: {(Incredible-The term {(incredible" used in conjunction with the phenomenon of a fire-induced circuit failure, signifies the PIRT panel's conclusion that the event will not occur. In these cases, the PIRT panel could find no evidence of the phenomenon ever occurring, and there were no credible engineering principles or technical arguments to support its happening during a fire." DCPP plans to revise the detailed circuit analysis and CFMLA based on the updated guidance in NUREG/CR-7150.
As part of this update, DCPP will adopt the NUREG/CR-7150 positions on other fire-induced cable failures, including three-phase proper polarity hot shorts. This update will be included in the Fire PRA models used to respond to PRA RAI-03. Based on the technically substantiated position in NUREG/CR-7150, DCPP's update to the CFMLA will exclude three-phase proper polarity hot shorts based on their classification of "incredible," which no longer has stipulations for high consequence equipment based on grounding type or insulation type. d) PG&E calculations (revised subsequent to the Peer Review) describe the treatment of thermoset or thermoplastic cables in the circuit failure analysis implemented for the DCPP Fire PRA . As established in this calculation, the CFMLA analysis follows the guidance in Chapter 10 of NUREG/CR-6850 for Option 1. Spurious operation probability values were based on insulation material classification.
Where the insulation material could not be confirmed it was conservatively assumed to be thermoplastic material.
In addition, the analysis includes no credit for control power transformers (CPTs). Specifically, Section 4.0 of the calculation states: 44 Enclosure PG&E Letter DCL-14-11 0 "The cable codes to determine if the cables of concern are single conductors, multiconductors, and whether the cable has insulation that is thermoset or thermoplastic are obtained from SETROUTE, DCPP's Cable and Raceway database." In summary, the cable and raceway database SETROUTE is used for determining if the cable insulation is thermoset or thermoplastic.
Based on this information, the appropriate circuit failure mode probabilities are selected following the guidance for Option 1 in Chapter 10 of NUREG/CR-6850 without CPT credit. Where insulation type could not be confirmed it was conservatively assumed to be thermoplastic material.
Subsequent to submittal of the DCPP LAR, NUREG/CR-7150 Volume 2 have been issued, which includes updated conditional spurious operation probabilities and the corresponding guidance for their application.
DCPP plans to revise the CFMLA based on the updated guidance in NUREG/CR-7150, Volume 2, which will include consideration of insulation materials as established in the new spurious operation probability tables. This treatment will be included in the Fire PRA model that is used to support the aggregate analysis in response to PRA RAI-03. e) Fire related F&O PRM-C1-01-2010 was addressed and dispositioned as closed prior to the LAR submittal.
As discussed in the LAR Attachment S, all of the planned modifications listed in Table S-2 are credited in the FPRA model, and planned modifications have been properly documented in the Development of Fire-Induced Risk Model notebook.
Table S-2 lists modifications that resolve VFDRs and non-VFDR modifications (i.e., Item S-2.3, RCP Seal Cooling Modification and Item S-2.4, Incipient Detection System Installation).
Table 1 below identifies modifications associated with VFDR resolutions and non-VFDR modifications.
Table 1-Identification of Modifications Associated with VFDRs Modification Unit Proposed Modification In Fire Resolves Non-VFDR that No. PRA VFDR Reduces Risk Electrical Raceway Fire Barrier System (ERFBS) Unit Installation to provide S-2.1 1 protection to Conduit YES Partial* NO KT351 associated with Steam Generator Level Transmitter MS-1-L T-529. 45 Enclosure
- PG&E Letter DCL-14-110 Table 1-Identification of Modifications Associated with VFDRs Modification Unit Proposed Modification In Fire Resolves Non-VFDR that No. PRA VFDR Reduces Risk Unit 1 Hot Shutdown Panel S-2.2 and YES YES NO Modifications S-2.3 S-2.4 S-2.5 Unit 2 Unit 1 Reactor Coolant Pump and (RCP) Seal Cooling YES NO YES Unit Modification 2 Unit 1 Incipient Detection and YES NO YES Unit System Installation 2 Electrical Raceway Fire Barrier System Unit (ERFBS). YES YES NO 2 (Re-route of cable associated with MS-2-FCV-95).
- The Unit 1 Electrical Raceway Fire Barrier Systems (ERFBS) modification partially resolves a VFDR, resulting in a reduction of delta risk for the associated fire area. f) PG&E has committed to the RCP seal cooling modification to reduce the risk of a LOCA resulting from a loss of RCP seal cooling as indicated in Item S-2.3, Table S-2 of AttachmentS of PG&E LAR 13-03 dated June 26, 2013. At the time of the submittal of the LAR, PG&E was planning to install the Westinghouse RCP SHIELD Passive Thermal SDS to satisfy this commitment.
The RCP seal model in the FPRA model was modified based on guidance in WCAP-17100-P, Supplement 1 and WCAP-17541-P.
The FPRA model with this modification was used to support the LAR dated June 26, 2013. Given unacceptable operating experience with the Westinghouse RCP SDS reported in a 10 CFR Part 21 letter from Westinghouse to NRC dated July 26, 2013, PG&E decided to install redesigned Westinghouse RCP Generation Ill SDS (GEN Ill) instead. A detailed description of the GEN Ill SDS can be found in TR-FSE-14-1, Revision 1, "Use of Westinghouse SHIELD Passive Shutdown Seal for FLEX 46 Enclosure PG&E Letter DCL-14-11 0 Strategies." Modeling of the RCP GEN Ill SDS applicable to the FPRA model as well as other hazard PRA models will be based on guidance in PWROG-14001-P, Revision 1, "PRA Model for the Generation Ill Westinghouse Shutdown Seal." PWROG-14001-P, Revision 1 has been submitted to NRC (Reference PWROG OG-14-211 dated July 3, 2014) but has not yet been approved.
The post-transition FPRA model will be verified to reflect the as-approved version of PWROG-14001-P before it is used as a basis in self-approval of post transition changes. This report provides the basis for the GEN Ill SDS credit in the PRA model in the response to PRA RAI3. The PRA model for the RCP GEN Ill SDS takes into account both the probability that the GEN Ill SDS fails to actuate when demanded and remains sealed after successfully actuated, and the probability that the operators fail to trip the RCPs in a timely manner following a loss of all seal cooling. The GEN Ill RCP seals will limit RCS inventory losses to very low leak rates when actuated in the event of a loss of RCP seal cooling. The low RCS leak rates during an accident or transient should lessen the burden on timeliness and level of mitigating actions. The Table below provides a comparison of the currently installed Westinghouse (W) RCP seals, the Westinghouse RCP SDS credited in PG&E LAR 13-03, and the GEN Ill SDS to be used in the integrated response to PRA RAI3. Comparison Original W As Credited In LAR 13-To Be Credited In PRA Seals 03 RAI-3 Guidance NUREG/CR-WCAP-17100-P, PWROG-14001-P, Used 5167(Rhodes Supplement 1 and Revision 1 (GEN Ill), Model) WCAP-17541-P WCAP-15603 Revision 1-A (WOG2000 model) Time to trip 13 minutes to 13 minutes to protect the 13minutes to protect the RCPs upon protect the SDS GEN Ill a loss of normal seal seal cooling package (#1 and Seal) 30 minutes to protect the injection normalsealpackage(#1 seal) 47 SDS Fails: 480 gpm** Enclosure PG&E Letter DCL-14-110 GEN Ill Failed: 480 gpm **
- SDS and GEN Ill Failed-to-actuate probability is independent of RCP Trip ** 480 gpm/pump is the maximum leakage possible ***Reference PWROG-14001-P, Revision 1, Section 2.5, Figure 2.5-1 ****The failure probabilities are proprietary information.
In order to avoid redacting the RAI response, the exact source of the values is specified.
- Reference PWROG-14001-P, Revision 1, Section 2.3.2 ******Reference PWROG-14001-P, Revision 1, Figure 2.5-2 48 Enclosure PG&E Letter DCL-14-110 g) As part of the initial fire modeling effort, plant walkdowns were conducted for every fire compartment where detailed fire modeling was performed.
During these walkdowns, in conjunction with the available plant data, electrical panels were investigated for their physical and electrical characteristics using the guidance provided by Section 6.5.6 of NUREG/CR-6850 and Section 8.2 of NUREG/CR-6850, Supplement 1 (FAQ 08-0042).
The maximum rated voltage, ventilation conditions, penetrations, robustness of panel enclosure, and any combustibles within inches of the cabinet were observed and documented for each panel. These walkdown results for each panel are documented in each detailed fire modeling report. Based on these observations, electrical distribution panels were screened as follows per NUREG/CR-6850:
- Well-sealed electrical cabinets that have robustly secured doors (and/or access panels) and that house only circuits below 440 V. o Cabinets with heavy gauge metal construction and doors which are reinforced with turned in edges which would prevent warping in the event of internal fire were considered well-sealed and robustly secured. o Panels with voltages above 440 V (e.g., MCCs) were not screened from the analysis
- Small, wall-mounted, fully enclosed cabinets, housing less than four switches.
These are located throughout the plant as local controllers for one or more pieces of equipment.
They contain little combustibles and are screened as nondamaging ignition sources. o Cabinets that would fall into this category typically include small 120 VAC distribution panels, small lighting panels, and small control panels. o These types of panels typically have a low combustible loading inside a relatively well-sealed cabinet (i.e., no openings, no unsealed penetrations, and no ventilation openings).
o The combustible materials (i.e., cables) inside the panel are normally separated internally from the breaker casings and therefore do not communicate directly with the breakers, which is the likely source of ignition.
o Based on these factors, the fire is expected to be substantially contained within the cabinet, behind the circuit breakers and panel door. No cabinets that are PRA components were screened.
A nonventilated cabinet that is also a PRA component is capable of damaging itself and all 49 Enclosure PG&E Letter DCL-14-110 cables that terminate at the cabinet. Therefore a fire scenario was still created if the cabinet was a PRA component or had terminating PRA cables. A confirmatory walkdown of screened electrical cabinets revealed that one cabinet in Compartment 5-A-4 and one cabinet in Compartment 5-B-4 required additional analysis due to cabinet doors that were not as robust as initially assumed. The fire modeling analyses for these compartments have been updated. The updated fire scenarios in the detailed fire modeling analysis will be reflected in the updated fire risk results that will be provided to the NRC after the Fire PRA is updated and additional quantification is performed in the response to PRA RAI-03. h) At DCPP, the MCR abandonment is considered both for loss of habitability scenarios and for loss of control (without loss of habitability).
The MCR abandonment due to loss of habitability is analyzed using detailed fire modeling of fires in the MCR to determine the probability of failing to extinguish the fire prior to operators being exposed to environmental conditions HGL height, temperature and/or smoke density) which preclude them from remaining and functioning in the MCR. The MCR abandonment due to loss of significant control from the MCR is modeled in a very similar way to the loss of habitability case. This significant loss of control is also addressed in this response by noting the differences between it and loss of habitability.
Since the human reliability aspects of both abandonment cases also have significant overlap, they are both addressed in the response to PRA RAI-22. Basis for assumptions made about smoke purge system The value of 1 o-3 used in the original MCR abandonment analysis submitted in the NFPA 805 LAR for the unavailability of the Smoke Purge function of the MCR ventilation system was an approximation based on the unavailability of the Normal mode of the MCR ventilation system. A review of the original MCR abandonment analysis identified that it did not consider or include the unavailability contribution from: (1) a failure of operator action to switch the MCR ventilation from the Normal mode to the Smoke Purge mode when required, and (2) routine maintenance of the Exhaust Fan (i.e., E-35). As part of an update to the MCR abandonment analysis in response to this RAI, FM RAI-01.i, and PRA RAI-08, the unavailability of the Smoke Purge system has been updated to include the unavailability of associated equipment due to their failure and maintenance/testing and the failure probability of operator action to switch to the Smoke Purge mode. Modeling of MCR transient fires and effect on abandonment The effect of transient fires on conditions within the MCR was originally analyzed using CFAST and was documented in the MCR abandonment analysis.
Multiple cases were run, including cases both with and without the ventilation system operating.
The results of the analysis showed that there 50 Enclosure PG&E Letter DCL-14-110 were no cases where the abandonment criteria were reached for a transient fire, either with or without the ventilation system operating.
This conclusion has been confirmed with the updated MCR abandonment analysis performed in response to FM RAI-01.i.
Description and justification of approach to crediting abandonment As a result of the additional analysis discussed in this RAI, FM RAI-01.i, and PRA RAI-08, the DCPP fire PRA will no longer assume a CCDP of 1.0 for the case of abandonment due to loss of habitability.
Instead, a loss of habitability abandonment model will be integrated into the PRA. The MCR abandonment analysis will address the following aspects:
- The time of abandonment, based on the detailed fire modeling performed for the loss of habitability calculation
- Updated unavailability of the Smoke Purge system due to random and fire induced failures of equipment and failure of operator action to switch to the Smoke Purge mode of the MCR HVAC system
- Fire propagation among ignition sources The development of the MCR abandonment fire scenarios will consider the following aspects:
- Location of a fire; MCR versus Cable Spreading Room, Main Control Board (MCB) versus non-MCB electrical cabinets in the MCR, and MCBs housing circuits (controls or AC power supports) associated with the Smoke Purge system versus MCBs not housing such circuits
- Fire-induced failures that create plant conditions that are beyond the capability of the remote shutdown equipment or procedures will fail remote shutdown.
Since the hot shutdown panel and the associated procedures only credit the operation of certain equipment, it is necessary to consider whether the fire-induced failures associated with an abandonment scenario cannot be obviated even if all actions are correctly taken and all equipment operates correctly.
- Nonrecoverable fire-induced and random failures of equipment that is required for remote shutdown, including the hot shutdown panel equipment, will be accounted for in the model. This is required in order to address the situation where the operators take all the correct actions in accordance with the abandonment procedure, but the equipment fails to operate correctly.
51 Enclosure PG&E Letter DCL-14-11 0
- Failure probability of operator actions to achieve and maintain safe and stable condition from the HSDP. Detailed discussion and justification of the Human Reliability Analysis (HRA) is provided in the response to PRA RAI-22.
- Impacts ori the opposite unit with no fire damage as the MCR abandonment due to a loss of habitability affects both units.
- Consistent with NRC's current position regarding loss of habitability abandonment, as documented in draft FPRA FAQ-0002, discussed in meetings, and accepted in past RAI responses from other plants, there is no need to account for the possibility that operators will fail to abandon the MCR (i.e., it will be assumed that the operators will not remain in the MCR when subjected to uninhabitable conditions).
The modeling for the loss of control case has two differences:
- There is no need to consider achieving remote shutdown of both units, as there are no scenarios that can result in a loss of control on both , units. Loss of habitability on both units that also involves a loss of control on the originating unit is covered by the fire-induced failure modeling that is incorporated in the abandonment modeling (i.e., the equipment failures would be captured as a consequential loss of control situation).
- A cognitive error for failure to abandon on loss of control in time to successfully achieve remote shutdown will be included in the model. Detailed discussion and justification of the HRA is provided in the response to PRA RAI-22. Both analyses will address the full range of possible damage scenarios associated with abandonment, from scenarios where there is essentially no fire-induced damage (which may occur for some loss of habitability scenarios) up to and including scenarios where the damage may be such that successful remote shutdown is not feasible.
The integrated analysis being performed in response to PRA RAI-03.a will include the impacts of the updated fire modeling and unavailability of the Smoke Purge system on the frequency of the MCR abandonment due to induced adverse MCR environment.
i) Fire wrap is credited in the FPRA for risk reduction in the following Unit 1 Fire Areas: 1, 3-BB (115 foot), 5-A-4, and 10. Fire wrap is credited in the FPRA for risk reduction in the following Unit 2 Fire Areas: 3-CC (1 00 foot), 3-CC (115 foot), 5-B-4, 9, 20, and 22-C. All credited fire wrap is qualified to at least a one-hour fire resistance rating or equivalent.
52 Enclosure PG&E Letter DCL-14-110 The technical justification for qualification of all fire wrap credited for risk reduction is documented as follows: Fire Areas Technical Justification Documentation 3-BB and 3-CC PG&E Calculation C-FP-1 04, "Evaluation of 3M Series E-50 Raceway Fire Barrier Wrap Configurations at DCPP" 5-A-4 and 5-B-4 PG&E Calculation C-FP-1 04, "Evaluation of 3M Series E-50 Raceway Fire Barrier Wrap Configurations at DCPP" 10 and 20 PG&E Drawing 502986, "Fire Protection, 12 kV Start-Up Switchgear Room, Area 'A"' 1 PG&E Drawing 515188 PG&E Design Change Notice 1-EA-47568 PG&E Design Change Notice 1-EA-47957 9 PG&E Design Change Notice 2-EA-48568 22-C PG&E Calculation C-FP-1 04, "Evaluation of 3M Series E-50 Raceway Fire Barrier Wrap Configurations at DCPP" Based on the above, the PG&E NFPA 805 LAR is revised to clarify that fire wrap (Electrical Raceway Fire Barrier System, ERFBS) is credited for risk reduction in Fire Areas 1, 9, and 22-C. This revision is applied to Table 4-3, "Summary of NFPA 805 Compliance Basis and Required Fire Protection Systems and Features," and to Attachment C, "NEI 04-02 Table B-3 Fire Area Transition." Also, the first paragraph in the disposition to F&O FSS-C8-01 (2008) in Attachment V to the PG&E NFPA 805 LAR is revised to read: "Fire Wrap information is available from the DCPP Cable/Raceway database (i.e., SetRoute Data) and PG&E Design Calculation 9000041072 (Legacy No 134-DC-T) (Attachment 2). Technical justification for qualification of the fire wrap is provided in PG&E Design Calculation 9000002223 (Legacy No. C-FP-104), Drawings 502986 and 515188, and Design Change Notices 1-EA-47568 , 1-EA-47957, and 2-EA-48568. fire wrap(s) in each Fire Area/fire scenario is documented in its risk modeling workbook (Appendix R of Calculation F.3.5). The surveillance test procedure, STP M-70D, provides requirements for visual inspection of wraps." j) DCPP PRA has reviewed and analyzed the DCPP specific maintenance and testing data for all credited fire suppression (carbon dioxide and wet pipe sprinkler systems) and detection systems. The degradation and/or unavailability of fire protection features is tracked via the Technical Specification (TS) and Equipment Control Guideline (ECG), and therefore, the DCPP TS tracking database was the source of the plant-specific data. The data selected for the review spans from 2008 through 2011. The results of the review showed that the unavailability of these systems credited in the DCPP FPRA were within the values prescribed in Section P.1.3, Appendix P (Detection and Suppression Analysis) of 53 Enclosure PG&E Letter DCL-14-110 NUREG/CR-6850.
The review of the DCPP specific experience revealed no outlier behavior for any fire detection or fire suppression systems. k) In the uncertainty analysis of fire induced CDF SOKG was considered for all parameters in the Fire PRA model with the exception of fire ignition frequency and associated fire modeling parameters (e.g., nonsuppression probability).
The analysis was performed using the "Big Loop" Monte Carlo Calculation function of the RISKMAN software.
The method used in RISKMAN essentially samples all the parametric distributions in the model and applies each sample value of each parameter to all the basic events which have the same parameter.
The system/top event failure probabilities are next requantified using the sampled value from all the parameters (including component failure rates and unavailability values, human error probability, short probability, etc.). The fire induced accident sequences are next quantified using the updated system/top event split fraction values to obtain a sample CDF and LERF value. The process is repeated thousands of times to obtain sufficient number of CDF and LERF sample values to develop the distributions for the fire-induced CDF and LERF. The "Big Loop" Monte Carlo approach therefore inherently considers SOKC. In response to PRA RAI-03, the uncertainty analysis for the fire induced CDF and LERF using the Fire PRA Model will include the distributions for the fire ignition bin frequency and associated fire modeling parameters (e.g., nonsuppression probability).
These parameters will take into account SOKC. The mean CDF value and the mean LERF value from the uncertainty analysis will be compared to the CDF and LERF values calculated using the point estimate values in the Fire PRA model. I) The HRA dependency analysis performed subsequent to the 201 0 peer review was assessed to determine whether an HFE floor value should be used in the fire PRA. The analysis showed that most HFE combinations were above the 1 E-05 minimum joint HEP criterion.
In a few cases where the joint HEP was below the 1 E-05 criterion, HFE combinations were reviewed and found to involve nonconsequential HFE failures.
Nonconsequential failures are those that do not impact the accident sequence end state and are found in sequences that are complementary to core damage sequences involving success of the same HFE. If nonconsequential HFE failures were found, the HEP value for that HFE was omitted from the joint probability calculation.
Examples of these HFE combinations are:
- Fire induced Station Blackout (SBO) with HFE to initiate feed and bleed. Because feed and bleed is not possible under SBO conditions the sequence results in core damage regardless of feed and bleed outcome.
Because recirculation is not possible under SBO conditions, these 54 Enclosure PG&E Letter DCI--14-11 0 types of sequences result in core damage regardless of recirculation outcome. After correcting for the presence of nonconsequential HFE failures, the results of the assessment showed that the use of a floor value was not necessary.
The process to assess HFE dependencies will be performed again as part of the integrated analysis in response to PRA RAI-03 to ensure that changes made to the model since the previous dependency analysis have not altered the minimum joint HEP conclusions.
m) During the 2008 Fire PRA Peer Review, two F&O's related to Supporting Requirement (SR) Maintenance Update (MU) were issued. They were F&O MU-A 1-01 (2008) observing that the PRA administrative procedure did not address the FPRA per the guidance in the PRA Standard and F&O MU-A2-01 (2008) observing a lack of guidance on monitoring changes in PRA technology and industry experience.
In response to these F&Os, DCPP PRA prepared a new administrative procedure AWP E-028, Revision 0, "PRA Model Maintenance and Update." During a follow-on peer review of the FPRA in 2010, the Technical Element MU (per NEI 00-02) including these two SRs was rereviewed in entirety.
As documented in the 2010 FPRA Peer Review Report (Table 4-16 and Table B-13 of Appendix B of Westinghouse Letter L TR-RAM-11-11-004), the Peer Review team concluded that these two SRs were met based on the review of the DCPP administrative procedure AWP E-028, Revision 0, "PRA Model Maintenance and Update." Section 5.4 of AWP E-028 refers to Attachment 1, which requires review of industry experience and PRA technology on a periodic basis as well as monitoring of plant specific changes such as designs and procedures.
The requirements are general in nature such that it covers internal events and other hazards (including fire related events or changes).
NRC PRA RAI 3: 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 NRC. RG 1.17 4, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," provides quantitative guidelines on core damage frequency, large early release frequency, 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. The NRC staff review of the information in the LAR has identified additional information that is required to fully characterize the risk estimates.
55 Enclosure PG&E Letter DCL-14-11 0 The PRA methods currently under review in the LAR include:
- PRA RAI 2.a regarding systems or actions needed for safe and stable state at 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />
- PRA RAI 22 regarding CCDP developed for loss of control scenarios 56 Enclosure PG&E Letter DCL-14-11 0
- FM RAI 1. i regarding calculation of time to abandonment Please provide the following:
a) Results of an aggregate analysis that provides the integrated impact on the fire risk (i.e., the total transition CDF, LERF, .llCDF, .llLERF) of replacing specific methods identified above with alternative methods which are acceptable to the NRC. In this aggregate analysis, for those cases where the individual issues have a synergistic impact on the results, a simultaneous analysis must be performed.
For those cases where no synergy exists, a at-a-time analysis may be done. For those cases that have a negligible impact, a qualitative evaluation may be done. It should be noted that this list may expand depending on NRC's review of the responses to other RAis in this document.
b) For each method (i.e., each bullet) above, explain how the issue will be addressed in 1) the final aggregate analysis results provided in support of the LAR, and 2) the PRA that will be used at the beginning of the self-approval of post-transition changes. In addition, provide a method to ensure that all changes will be made, that a focused-scope peer review will be performed on changes that are PRA upgrades as defined in the PRA standard, and that any findings will be resolved before self-approval of post-transition changes. c) In the response, explain how the RG 1.205 risk acceptance guidelines are satisfied for the aggregate analysis.
If applicable include a description of any new modifications or operator actions being credited to reduce delta risk as well as a discussion of the associated impacts to the fire protection program. d) If any unacceptable methods or weaknesses will be retained in the PRA that will be used to estimate the change in risk of post-transition changes to support self-approval , explain how the quantification results for each future change will account for the use of these unacceptable methods or weaknesses.
PG&E Response:
During a clarification call with the NRC held on October 16, 2014, PG&E proposed extending the response submittal due date for RAI PRA-03, Integrated Analysis, to late February 2015, pending a confirmation from the NRC that the technical approach provided in the remaining 120-day responses due on November 28, 2014, are acceptable, applicable methods of analysis.
On October 28, 2014, the PG&E received approval from Mr. Barry Miller , NRC Project Manager , accepting the proposed extension of PRA-03, providing additional guidance to PG&E to arrange a clarification call with the NRC on January 16, 2015, to discuss the 120-day response applicability.
The approval and guidance received electronically as follows: 57 Enclosure PG&E Letter DCL-14-11 0 "The licensee's proposed approach is acceptable and we do not believe we need to hold another teleconference.
We can set a goal for us to review their 120-day response (which is due November 28th) and hold a call with them by January 16 to discuss if we find their analysis methods (i.e., their responses to the RAis) acceptable.
If not, we may need follow-up RAis. But if no follow-ups are needed, then we can set 6 weeks from the date of that call as a deadline to receive their response to PRA RAI 03, which is kind of a "roll-up" calculation that they can't finalize until we've found their other responses acceptable.
This is a very similar approach to what we do for other licensees, hence we don't believe a call is necessary." NRC PRA RAI 5: Though not specifically acknowledged in the LAR, it appears that reductions below the 98th percentile NUREG/CR-6850 HRR of 317 kW for transient fires are credited in the FPRA (e.g., the analysis shows that a 98th-percentile HRR of 142 kW was used for fire area 6-A-3). Please identify the fire areas for which reduced HRRs are credited and discuss the key factors used to justify the reduced rate below 317 kW per the guidance provided in the June 21, 2012, memorandum from Joseph Giitter to Biff Bradley (Recent Fire PRA Methods review Panel Decisions and EPRI 1022993, 'Evaluation of Peak Heat Release Rates in Electrical Cabinets Fires'," ADAMS Accession No. ML12171A583).
Include in this discussion:
a) Identification of any other fire compartments for where reduced HRR transient fires are credited.
b) For each location where a reduced HRR is credited, a description of the administrative controls that justify the reduced HRR including how specific attributes and considerations are addressed.
Provide a discussion of required maintenance for ignition sources in each location, and types/quantities of combustibles needed to perform that maintenance.
Also discuss the personnel traffic that would be expected through each location.
c) The results of a review of records related to violations of the transient combustible and hot work controls.
d) The impact of assuming reduced HRRs on the fire risk estimates.
Note that if a reduced HRR cannot be justified using these guidance criteria, then use the transient fire HRR values from Table G-1 of NUREG/CR-6850 in the integrated analysis performed in response to PRA RAI 3. PG&E Response:
a) Reduced HRRs for transient fires are credited in the following FCs:
- FC 4-A : Chemical Laboratory Area (69 kW)
- FC 6-A-3 : Unit 1 Battery, Inverter and DC Switchgear, {(H" Bus (142 kW) 58 Enclosure PG&E Letter DCL-14-110
- FC 8-G : Unit 1 Solid State Protection System (SSPS) Room (69 kW)
- FC 8-H : Unit 2 Solid State Protection System (SSPS) Room (69 kW) b) In FC 4-A, reduced transient HRRs (i.e., 69 kW) have been credited in the cable chase areas in the North West area of the compartment.
These areas are largely inaccessible (the space is mostly occupied by cable trays) and have only a limited floor area available to potentially accumulate transient combustibles.
Additionally, these areas have clear signage and/or painted floor areas indicating that combustible storage is prohibited.
Therefore, only a very small fire was postulated for these areas. In FC 6-A-3, there are two transient fire scenarios postulated along the north wall of the charger/inverter room where the 75th percentile HRR (i.e., 142 kW) was used. For a Compartment such as 6-A-3 with no pumps, motors or potential oil fires, the 75th percentile HRR bounds the possible transient ignition sources which were fire tested and identified in NUREG/CR-6850, Appendix G, Table G-7, with the exception of tests involving untreated wood (untreated wood is prohibited at DCPP in areas with safety-related equipment), airline trash bags with over 2 kg of paper products (such a large quantity of paper products will not be present in Compartment 6-A-3) or over 4 kg of straw/grass/eucalyptus duff (this type and quantity of combustible plant matter will not be present in Compartment 6-A-3). Since fires that are not bounded by the tests will not occur in Compartment 6-A-3, the 75th HRR was used for these transients.
Transient combustible materials are strictly controlled in FCs 8-G and 8-H in accordance with plant procedures and administrative controls.
Compartments 8-G and 8-H are classified as "No Combustible Storage" locations.
Therefore paper, cardboard, scrap wood, rags and other trash shall not be allowed to accumulate in these areas. FCs 8-G and 8-H are labeled combustible free zones, thus transient combustibles in the compartment are restricted.
These compartments have also been labeled hot work free zones. Therefore, a reduction in the transient fire size from the 317 kW recommended by NUREG/CR-6850 is justified in this area and transient fires have been characterized as a 69 kW fire. Since transient combustibles are strictly controlled, any temporary storage of transient materials for maintenance in excess of 15 pounds will require special permitting from the job supervisor.
Permitting requires controls on proper use and removal of transient materials as well as any compensatory measures that need to be enacted while the maintenance occurs. This enables proper notification and alert to the job supervisor of additional hazards and the duration for which this hazard exists to enable a prompt response in the event of an incident during maintenance.
A review of transient combustible permits from 2009 to 2014 for these compartments indicates that typical ignition source maintenance activities 59 Enclosure PG&E Letter DCL-14-11 0 requiring a permit include, but are not limited to, cleaning and inspecting battery chargers, load tests on battery chargers, and troubleshooting and repairing equipment.
The types and quantities of combustibles that may be introduced to perform these maintenance activities consist of limited amounts of testing wires, cable insulation, various test meters, plastic service carts, mixed plastic/rubber and metal tools, small tables, and chairs. The transient combustible permits for these activities have been reviewed by fire protection and found to be acceptable with no additional compensatory actions. Large combustible liquid fires will not occur in areas with a reduced transient HRR since activities in the compartment or applicable portions of the compartment (i.e., the 4-A cable chase) do not include maintenance of oil containing equipment.
All maintenance activities not requiring a permit (i.e., less than 15 pounds of transient material) are considered bounded by the transient analysis.
Access to these compartments is strictly controlled.
All four areas are within the power block, and thus access is restricted to those with unescorted access authorization.
While there is regular access to compartment 6-A-3, it is not in a normal travel path, so traffic is limited to watchstanders, engineers on walkdowns, maintenance personnel, etc. The cable chase areas in Compartment 4-A are largely inaccessible (the space is mostly occupied by cable trays) and are not in a normal travel path or frequented as part of normal operational activities.
Compartments 8-G and 8-H are only accessible from the MCR and have additional access controls in that the doors are locked and access can only be granted by an on-watch SRO. These access controls reduce the likelihood that transient combustibles will be inadvertently introduced into the FCs in any appreciable quantity.
c) Records dated between January 2009 and October 2014 identifying violations of hot work and transient combustible controls were reviewed.
During this period, there were no notifications written for hot work control issues or transient combustible control issues in FCs 8-G, 8-H, 6-A-3, or 4-A. d) All reduced HRRs credited in the Fire PRA have been justified, as documented in the DCPP responses to PRA RAI-05.a, b, and c. Reducing the transient HRR where reasonable and justifiable allows for a more realistic transient fire ZOI. This reduced ZOI for the subject transient fire scenarios prevents ignition of secondary combustibles, and results in a more limited Fire PRA target damage set. This ultimately provides for a more realistic CDF/LERF for the transient scenarios and respective FCs. NRC PRA RAI 8: a) The analysis provides a description of how MCB fires are modeled for the FPRA and includes an explanation that the probability of damage to a target set on the MCB is performed in accordance with the method described in Appendix L of NUREG/CR-6850.
The analysis also explains that the MCB for each unit is comprised of five adjoining vertical boards (i.e., VB1 to VB5) 60 Enclosure PG&E Letter DCL-14-110 arranged in an L-shape and three smaller Control Console sections (i.e., CC1, CC2, and CC3). The analysis also presents probabilities of MCB fire scenarios for different so-called Functional Target Sets (FTSs) based on the distance "d" between targets and that the distance "d" appears to be limited to targets contained within a vertical board (or the Control Console) and ranges from about 6" to 30." If targets are constrained to a single vertical board, please provide a discussion of barrier between vertical sections which limits propagation.
Otherwise, extend the target set across the vertical boards. b) The licensee's analysis indicates that the total adjusted fire ignition frequency is 1.65E-03/year for the two units. The analysis indicates that the MCB ignition frequency was divided between MCB sections according to the ratio of the section surface area to the total MCB surface area, but also shows that for MCR abandonment the whole MCB fire ignition frequency was used. Please explain this inconsistency and the correct ignition frequency that was used. If the MCB ignition frequency is divided by factoring in the MCB surface area, the NRG staff cannot complete its review. In this case, replace this treatment in the integrated analysis provided in response to PRA RAI 3, and provide an explanation of the approach used and the results in sufficient detail so the staff can make a conclusion regarding the use of the approach.
c) During the audit, the licensee indicated that supplemental analysis beyond NUREG/CR-6850, Appendix L, was developed to evaluate the risk due to horizontal raceways in the MCB. Please describe and justify the approach including the various aspects of the analysis and assumptions.
PG&E Response:
a) Target sets were not constrained to individual sections of the vertical board. The MCB fire modeling analysis considers a group of scenarios involving fires impacting multiple MCB vertical board sections.
For example, the target set for Fire Scenario VB23-1 includes targets in both Sections VB2 and VB3. As discussed in the MCR fire modeling analysis, when the value of "d" for a given functional target set is greater than the equivalent separation distance for damage to the corresponding cable raceway sections that run horizontally inside the MCB, the probability of damage to the Functional Target Set (FTS) is bounded by the probability of damage to the horizontal cable raceways, which is assumed to affect all components on the associated panels. This limits the separation distance "d" over which targets on the control board surface need to be considered.
A detailed description of the method used to calculate the probability of damage to the horizontal cable raceways due to a fire on the vertical board panel is provided in Section (c) below .. b) The sentence in the MCR Fire Modeling analysis, which states "Assuming that the frequency of a fire on any panel is proportional to the panel area, ... " is 61 Enclosure PG&E Letter OCL-14-11 0 misleading as the MCB ignition frequency has NOT been subdivided according to the panel area. The MCB ignition frequency of 8.24E-04 per unit was used in accordance with the method described in Appendix L of NUREG/CR-6850.
The statement was intended to explain the basis for the weighting factors used in calculating the conditional probability that, given a fire somewhere on the MCB, it will be on the upper or lower section. These weighting factors (i.e., conditional probabilities) were used to calculate the overall failure probabilities of the cable raceways, due to a fire on either the upper or lower MCB panel. The wording in the MCR Fire Modeling analysis has been modified to clarify the approach as reproduced below: "In the above Table 8, the overall failure probability of the upper or lower cable raceway due to a fire on either the upper or lower panel of a given section of the MCB, was calculated as a mean value, weighted according to the conditional probability that the fire occurs on the upper or lower MCB panel. Assuming that the probability of a fire on any panel is proportional to the panel height, the weighting factors are 0.6 and 0.4 respectively for the upper and lower panels, which are 48" and 32" in height respectively." c) The probability of damage to the horizontal raceways is calculated by estimating the intensity and duration of the radiant heat emanating from a fire source on the vertical board front panel. The radiant heat flux at the raceways depends on the distance from the source and this is estimated from a simplified representation of the geometry of the raceways relative to the vertical board geometry.
A section through the vertical board is shown schematically in Figure 1. The section is shown at the longitudinal position of the fire source along the panel (in the direction of the X-axis). Figure 1 also gives a schematic representation of the upper and lower raceways.
An estimate of the radiant heat flux at the raceways is derived by considering the upper and lower sets of raceways as single point targets in the 20 representation of Figure 1 and as line targets in 30. An upper bound to the radiant heat flux, considering all possible locations of a fire on the control panel, is obtained if the two horizontal lines representing the upper and lower raceways are positioned at the minimum distance from the vertical board fascia and at the mid-height of the respective control panel (i.e., upper or lower panel). The minimum horizontal distance between the upper raceways and the upper control panel is 15 inches and the mid-height of the vertical panel is 24 inches above the bottom of the panel. For simplicity, the lower control panel is also represented by a vertical panel with the minimum horizontal distance between the lower raceways and the lower control panel estimated to be a mean value of approximately 26 inches and with the mid-height of the panel being 16 Inches above the bottom of the panel.
- A plan view showing each section of the vertical board is shown schematically in Figure 2. The figure represents the upper or lower cable raceways showing the minimum distance between the fire source and each section of the raceways.
62 Enclosure PG&E Letter DCL-14-11 0 The analysis was performed using Crystal Ball to determine the probability of damage to any section of the upper or lower raceways due to radiant heat from a fire source anywhere on the vertical board upper or lower panel averaged over all possible locations of the fire source along the whole length of the vertical board. The modeling approach is analogous to the methodology described in Appendix L of NUREG/CR-6850 insofar as the probability of target damage is calculated as a mean value taken over all possible locations of a fire on the vertical board upper or lower panels. The main differences in this approach are:
- the targets are not in the same plane as the fire source
- damage is caused by radiant heat, not the fire plume
- the damage probabilities are estimated by simulation, not calculated from formulae derived analytically The following parameters are inputs to the analysis and treated as random variables:
- fire peak heat release rate
- fire suppression time
- X-coordinate of the fire source
- Y -coordinate of the fire source. The peak HRR for each trial is sampled from a gamma distribution with parameters corresponding to those for a cabinet containing multiple bundles of qualified cable, consistent with the Appendix L method. The suppression time is sampled from an exponential distribution with parameter 0.33, corresponding to the mean suppression time for a control room fire as used in Appendix L. The HRR is assumed to increase in proportion to t2 from 0 to its peak value in 12 minutes (Appendix G of NUREG/CR-6850).
The HRR (Os) reached at the time of suppression (t) is thus given by: where Q s is the peak HRR of the fire source. 63 Enclosure PG&E Letter DCL-14-11 0 The longitudinal X-coordinate of the fire source for each trial is sampled from a uniform distribution between 0 feet and 57 feet (the total length of the vertical board). For a fire on the MCB upper panel, the vertical Y-coordinate of the fire source is sampled from a uniform distribution between 0 feet and 4 feet (the height of the MCB upper panel) and for a fire on the MCB lower panel, the Y-coordinate of the fire source is sampled from a uniform distribution or between 0 feet and 2.67 feet' (the height of the MCB lower panel). The minimum distance between the fire source and each section of the target raceway is calculated by first identifying, from the X-coordinate of the source, in which longitudinal section of the vertical board the fire is located. If the fire is in the same section of the vertical board as the section of the raceway being considered, the minimum horizontal distance is simply the perpendicular distance between the fire and the raceway (e.g., 15 inches for damage to the upper raceways from a fire source on the MCB upper panel). If the fire is not in the same section of the raceway, the minimum horizontal distance is the distance between the fire and the nearest end of the raceway section. This is illustrated in Figure 2, which shows the minimum horizontal distances between a fire in section VB2 of the vertical board and each section of the raceways. Using Pythagoras' Theorem, the minimum distance (r) is calculated for each section of the raceway and the maximum radiant heat flux (qr) is then given by: 0.4Q s q ,. =--2-4nr in which the factor of 0.4 is the radiant heat fraction.
The minimum radiant heat flux required for cable damage (11 kW/m 2) is taken from Appendix H of NUREG/CR-6850 for qualified cable. Appendix H also gives the time for damage to occur, dependent on the intensity of radiant heat flux. This relationship is modeled by a simple linear approximation.
For each section of the raceways, if the incident radiant heat flux is greater than the damage threshold and the suppression time is greater than the time to damage, that section of the raceways is damaged from the fire source being analyzed and this fire scenario is binned (or counted) as "target failure," otherwise it is binned as "no failure." Note that the event "target failure" for a given raceway section here includes events involving damage to one or more other sections adjacent to the given section. That is, as shown in Figure 2, a fire source in VB2 which damages raceways in Section VB2 could also damage the raceways located in VB1, VB3 , VB4 or VB5. The calculation is repeated over a large number of trials (typically 1 0,000) and the probability of damage to each raceway section, or combination of sections, is determined as the mean value of "target failure" for each. The results of this analysis give the probabilities of MCB fire scenarios involving damage to one or more sections of the upper or lower cable 64 Enclosure PG&E Letter DCL-14-110 raceways and are used together with probabilities derived from Figure L-1 of Appendix L of NUREG/CR-6850 for scenarios involving damage to components mounted on the MCB panels. Following the methodology described in Appendix L of NUREG/CR-6850, MCB scenarios are identified by considering only target sets of components on the MCB panel. In accordance with Figure L-1 of NUREG/CR-6850, the probability of damage to a given target set of components decreases as the maximum separation distance 'd' between the components increases.
As larger and larger target sets of components on the MCB are considered, there is a limiting value of 'd' at which the probability of damage to the target set is less than the probability of damage to the nearest section of the cable raceways.
Since damage to a section of the raceways is assumed equivalent to damage to all components on the associated section of the MCB panel, scenarios involving target sets of components on the MCB panel with values of 'd' greater than this limiting value are bounded by the scenario for damage to the corresponding raceways.
The limiting values of 'd' corresponding to each section of the upper and lower MCB panels are derived using Figure L-1 of NUREG/CR-6850 to determine the "equivalent distance" 'd' which equates to the probability of damage to each section of the upper or lower raceways.
The results of this calculation shows that the equivalent distance for the upper panels of MCB Sections VB1, VB2, VB3, and VB4 is 0. This means that damage to the upper raceways in these sections is more likely than damage to any target set of components on the corresponding section of the MCB upper panel and, since it is assumed that the consequences of damage to the upper raceways would be equivalent to damaging all the components on these sections of the MCB upper panel, this scenario bounds all scenarios involving subsets of components.
For other sections of the MCB, the distance 'd' is greater than 0, so it is necessary to consider scenarios involving subsets of components on the corresponding sections of the MCB, s.ubject to the corresponding limiting values of 'd'. Thus, for example, the limiting value of 'd' for the lower panel of MCB section VB2 is 29 inches. The target set for one of VB2 scenarios (Scenario VB2-7) has a value of 'd' equal to 21 inches and is therefore not bounded by damage to the lower raceways.
However, Scenario VB2-8 has a value of 'd' equal to 33 inches and is bounded by damage to the lower raceways, which is assumed to be equivalent to damaging all components in the lower (and upper) panels of VB2. Thus, all scenarios identified using the methodology of NUREG/CR-6850 Appendix L are included in the analysis.
If the scenario target set has a value of 'd' less than the corresponding limiting value, the frequency and fire impacts of the scenario are modeled explicitly and if the scenario target set has a value of 'd' greater than the corresponding limiting value, its frequency and fire impacts are included implicitly within the analysis of the bounding scenario of 65 damage to the cable raceways.
Figure -1 Schematic Section of Vertical Board Upper Raceways ---+-----. --Representative "Point" Target 15" 48" Representative "Point" Target Lower Raceways ----+---._ -Figure-2 Schematic Plan of Vertical Board Cable Raceways 0.00 11.33 \ 23.00 34.00 Fire Source 66 Enclosure PG&E Letter DCL-14-110 Fire source IX. Yl 48.00 57.00 X ft_
NRC PRA RAI 14: Enclosure PG&E Letter DCL-14-11 0 LAR Attachment W, Section W.2.1 provides a description of how the LlCDF and LlLERF associated with VFDRs is determined.
Please supplement the description by: a) Describing the types of model adjustments that were made to remove different types of VFDRs from the compliant case FPRA. In addition, identify any major changes made to the FPRA models and data for evaluating VFDRs. b) Describing how VFDR and non-VFDR plant modifications are credited in the compliant and post transition PRA models. c) Describing of the type of VFDRs identified, and providing justification that any VFDRs identified but not modeled in the FPRA do not impact the risk estimates.
PG&E Response:
a) Model adjustments performed to evaluate VFDRs were consistent with the guidance of FAQ 08-0054. Modeling of the compliant case is performed by protecting components associated with the VFDR from fire damage. This is implemented in the PRA model using one or more of the following methods:
- The basic event associated with the fire impacted component is set to a success state or the value of the basic event is set to 0, then the top event(s) containing the basic event is requantified.
- For top events that model fire impact on components but do not include random failures, a split fraction representing "success" (that is, no fire damage) of the component is used within the logic rule in order to model/reflect the condition that the component is protected from fire damage.
- For top events that model fire impact on components and include random failures, a split fraction representing the random failure of the componenUtrain is used to replace the split fraction representing the random failure and the impact of fire on the componenUtrain within the 67 Enclosure PG&E Letter DCL-14-110 logic rule in order to model/reflect the condition that the component is protected from fire damage. For VFDR components not modeled in the FPRA, the post-transition model is modified to simulate the impact of fire on the VFDR component(s) by failing "surrogate" components/functions in the FPRA that represent an equivalent or similar impact on the plant. For example the impact of fire on the pressurizer heaters (which are not modeled in the FPRA) which results in pressurization of the RCS was simulated by modeling the fire impact as a challenge to the pressurizer PORVs via macros representing conditions that challenge the PORVs. For the corresponding compliant case, the simulated impact of fire on the VFDR component(s) is not modeled. This is equivalent to the VFDR component(s) not impacted by fire in the compliant case. In addition, some VFDRs were not modeled in the FPRA due to their low risk contribution.
These are discussed in the response to part c of this RAI. No major changes were required to model the compliant or post-transition case when evaluating VFDRs. b) The details of the VFDR and non-VFDR plant modifications are provided in Section 4.8.2 of the Transition Report and summarized in Table S-2 of Attachment S of the report. A description of how each of the plant modifications is credited in the compliant and post transition models is as follows:
- Incipient Detection System (IDS) Installation.
This non-VFDR modification will improve the ability to detect incipient fires in cabinets in the CSRs and the Solid State Protection System (SSPS) rooms. This modification is credited in both the compliant and post transition FPRAs and was modeled by reducing the frequency of fire propagating out of an electrical cabinet installed with the IDS, which could lead to more severe fire impacts.
- Hot Shutdown Panel (HSDP) Modifications. This modification will ensure that the required HSDP functions are independent and electrically isolated from the MCR or CSR and Decay Heat Removal and Inventory and Pressure Control capabilities are available in case of a fire in the MCR or CSR. The modification is credited in both the compliant and post transition FPRAs by recovering selected MCR or CSR fire sequences leading to core damage via actions at the HSDP (a Primary Control Station).
These actions are modeled in the accident sequence recovery event tree through the use of appropriate recovery split fractions of in the logic rules. 68 Enclosure PG&E Letter DCL-14-110
- RCP Seal Cooling Modification.
As indicated in the response to PRA RAI-01.f, given the unacceptable operating experience with the Westinghouse RCP SDS reported in a 10 CFR Part 21 letter from Westinghouse to NRC dated July 26, 2013, PG&E decided to install redesigned Westinghouse RCP Generation Ill SDS (GEN Ill). This is a non-VFDR modification.
Modeling of the RCP GEN Ill SDS applicable to the FPRA model as well as other hazard PRA models will be based on guidance in PWROG-14001-P, Revision 1, "PRA Model for the Generation Ill Westinghouse Shutdown Seal." PWROG-14001-P, Revision 1, has been submitted to NRC (Reference PWROG OG-14-211 dated July 3, 2014) but has not yet been approved.
The post-transition FPRA model will be verified to reflect the as-approved version of PWROG-14001-P before it is used as a basis in self-approval of post transition changes. The Gen Ill (Reference PWROG-14001-P, Revision 1) to be installed at DCPP will be credited in both the compliant and post transition FPRAs. Refer to the response to PRA RAI-15 for a discussion of the "risk offset" associated with this modification.
- Unit 1 ERF8S Installation.
The protection of conduit KT351 in Fire Area 3-88-115 eliminates the potential for fire damage to MS-1-L T -529 and subsequent closure of AFW-1-LCV-111 ensuring the availability of a feedwater flowpath to SG 1-2 via AFW pump 1-2 and level control valve AFW-1-LCV-111.
The modification is credited in both the compliant and post transition FPRA models by removing the fire induced impact on MS-1-LT-529 from all 3-88-115 fire scenarios.
- Unit 2 ERF8S Installation.
Rerouting of cable associated with MS-FCV-95 ensures the availability of steam supply to the driven AFW pump 2-1 for fire events in Fire Areas 5-8-4 and 6-8-4 thereby reducing the potential for a of loss of decay heat removal due to fire in these fire areas. This modification is credited in both the compliant and post transition PRA models by removing the fire induced impact on MS-FCV-95 from all 5-8-4 and 6-8-4 fire scenarios.
c) The only type of VFDR identified resulted from a lack of adequate NFPA 805, Section 4.2.3 success path separation.
69 Enclosure PG&E Letter DCL-14-11 0 VFDRs that were not evaluated quantitatively are as follows:
- VFDRs associated with level indications for the CST (MU-1/2-L T-40) and RWST (SI-1/2-LT-920, SI-1/2-L T-921, and SI-1/2-LT-922).
The loss of water level indications does not result in the loss of the CST or RWST as a water source. In addition, this level indication is not required to ensure that the water source remains available.
Therefore, not modeling these indicators in the Fire PRA does not impact the risk estimates.
- VFDRs associated with fire damage to cables contained in conduit that is either wrapped (1-hour rating minimum) or embedded in concrete, but where the requirements of NFPA 805 (2001 edition), Section 4.2.3.3 are not met. Because, there are no high energy hazard sources or analyzed fires that could challenge the protection, the risk impact of these VFDRs is insignificant.
- VFDRs associated with loss of process instrumentation (RCS or SG temperature and/or pressure indications) due to a fire in the control room or CSR. These VFDRs were not evaluated quantitatively for the LAR submittal as their relative contribution to the transition risk increase due to a fire in the CR or CSR is insignificant as compared to the increase due to damages to other SSD equipment.
However, to account for the additional risk, these VFDRs will be evaluated quantitatively and will be included in the risk estimates to be provided in response to PRA RAI-03.
- VFDRs associated with the loss of excore neutron detectors.
Excore neutron detectors are not modeled in the PRA as an input to the reactor trip function.
However, reactor trip due to other diverse process instrumentation is considered in the PRA. Therefore, these VFDRs have an insignificant impact on risk estimates.
- VFDRs associated with normally closed Manual Valves (MU-Q-1557, MU-0-280 and MFW-1/2-FCV-437) that are required to be opened prior to CST depletion in order to provide a suction source for AFW pumps of both units. These VFDRs were not evaluated quantitatively for the LAR submittal as their contribution to the transition risk increase is expected to be low. However, to account for the additional risk, these VFDRs will be evaluated quantitatively and will be included in the risk estimates to be provided in response to PRA RAI-03. 70 NRC PRA RAI15: Enclosure PG&E Letter DCL-14-11 0 LAR Attachment W, Tables W-4 and W-5 report change-in-risk values for each fire area before crediting "beyond compliance" (i.e., non-VFDR related) modifications, and the total change-in-risk based on the sum of all the fire areas after crediting VFDR modifications by use of a "risk offset" value. It is not clear why the risk reduction for just the RCP seal upgrade is so substantial (i.e., 5.35E-05/year for Unit 1 and 6.58E-05/year for Unit 2). Given the significance of the risk reduction and the fact that a conservative calculation of the compliant plant CDF and LERF can lead to a non-conservative calculation of the t:.CDF and t:.LERF, please provide the following:
a) Clarification of what modifications are associated with the risk reduction presented at the bottom of LAR Attachment W, Tables W-4 and W-5, and an explanation of how this risk reduction was determined.
b) A summary of the risk significant scenarios for fire areas in the compliant
- plant model which are most significantly impacted by the risk offset. c) A discussion of the contribution of fire-induced failures for those risk significant scenarios.
d) For the risk significant scenarios, a discussion of the impact of any assumptions made that significantly contribute to the variant and the compliant case risk. PG&E Response: (a) The plant modification associated with the risk reduction presented at the bottom of LAR Attachment W, Tables W-4 and W-5 is the installation of the Westinghouse RCP SHIELD Passive Thermal Shutdown Seals (SDSs). The incipient detection committed modification is also a risk reduction modification.
However, its contribution to risk is included in the post-transition and compliant case. Its risk reduction is not included in the "risk offset" value. Modeling of the SDSs was based on guidance in WCAP-171 00-P, Supplement 1 (PRA Model for the Westinghouse Shutdown Seal (SDS)), and WCAP-17541-P (Implementation Guide for the Westinghouse Reactor Coolant Pump SHIELD Passive Thermal SDS). The following steps were followed in the determination of the risk reduction
("risk offset" value) presented at the bottom of LAR Attachment W, Tables W-4 and W-5 due to the installation of the Westinghouse RCP SDSs: 71 Enclosure PG&E Letter DCL-14-11 0 1. The CDF and LERF values were first calculated for each of the applicable fire scenarios for each fire area using the compliant Fire PRA model that credits the RCP SDS modification.
- 2. Using the same compliant Fire PRA model, credit for the RCP SDSs modification was removed from the model by setting the split fraction for only the RCP SDSs to a guaranteed failed state. 3. The compliant Fire PRA model with no credit for the RCP SDSs modification was then requantified to obtain the CDF and LERF values for each of the applicable fire scenarios for each fire area. 4. The differences in the CDF and LERF values from Step 1 and Step 3 for all applicable fire scenarios were then summed to obtain the risk reduction or "risk offset" value shown at the bottom of Tables W-4 and W-5. (b)&(c) A summary of the risk significant scenarios for fire areas in the compliant model that are most significantly impacted by the risk offset, and the contribution of fire-induced failures for the risk significant scenarios are provided in the tables below in order of CDF "risk offset" value: For Unit 1 Fire Areas and Scenarios Fire Risk Fire Description of Description of Fire-Offset induced failures for Area (CDF)* Scenario Fire Scenario Scenario 7-A 4.01 E-05 Z7ATS17F2 Whole room Fire induced spurious burn-up Phase B signal scenario for the resulting in isolation of Unit 1 Cable RCP thermal bar'rier Spreading cooling supply valves Room (Fire and fire damage of seal Area ?A) injection valve FCV-128. RCPs fail to trip and SDS failed due to random causes resulting in a seal LOCA. Fire induced failure of RHR pumps resulting in the loss of sump recirculation cooling. 72 Fire Risk Fire Description of Offset Area (CDF)* Scenario Fire Scenario Z7ARNP4FO Fire initiated in cabinets RN P4 and RN04 in the cable spreading room (Fire Area 7 A) and the fire damage is limited to the cabinets.
6-A-3 2.11 E-06 ZBTC131F2 Fire initiated at the backup Battery Charger BTC-1-BTC131-0P in the Unit 1 125 VDC BusH Room (Fire Area 6-A-3). Fire damage to thermoplastic and all thermoset cable targets within the fire's ZOI. 73 Enclosure PG&E Letter DCL-14-11 0 Description of Fire-induced failures for Scenario Fire induced failures of RCP thermal barrier cooling supply valves, and Seal Injection Valve FCV-128. RCPs tripped but SDS fail to actuate due to random failures resulting in a Seal LOCA (Rhodes Model). RHR pumps unavailable for sump recirculation cooling due to operator failure to initially trip the pumps (fire caused degraded instrumentation).
Fire induced spurious Phase B signal resulting in isolation of RCP thermal barrier cooling supply valves and RWST draindown resulting in loss of Seal Injection.
RCP tripped but SDS fail to actuate due to random failures resulting in a Seal LOCA (Rhodes Model). RHR pumps unavailable for sump recirculation cooling due to operator failure to initially trip the pumps (fire caused degraded instrumentation).
Fire Risk Fire Description of Offset Area (CDF)* Scenario Fire Scenario 5-A-2 1.76E-06 ZTHG10F1 Fire initiated in cabinet THG1 0 in the 480 V Bus G MCC Room (Fire Area 5-A-2) with fire damage limited to the fire's ZOI. Enclosure PG&E Letter DCL-14-11 0 Description of Fire-induced failures for Scenario Fire induced failures of RCP thermal barrier cooling supply valves and Charging Pump 1-1. Fire induced loss of power supply to Charging Pumps 1-2 and 1-3 resulting in loss of RCP seal injection. RCPs tripped but SDS fail to actuate due to random failures resulting in a Seal LOCA (Rhodes Model).
- The Unit 1 fire areas with dominant LERF offset are similar to those for CDF offset For Unit 2 Fire Areas and Scenarios Risk Fire Description Description of Fire-Fire Area Offset Scenario of Fire induced failures for (CDF)* Scenario Scenario 7-8 4.00E-Z78TS17F4 Whole room Fire induced spurious 05 burn-up Phase 8 signal resulting scenario for in isolation of RCP Unit 2 Cable thermal barrier cooling Spreading supply valves and fire Room (Fire damage of Seal Area 7-8) Injection Valve FCV-128. RCPs fail to trip and SDS failed due to random causes resulting in a seal LOCA. Fire induced -failure of RHR pumps resulting in the loss of sump recirculation cooling. 74 Risk Fire Description Fire Area Offset Scenario of Fire (CDF)* Scenario Z7BPK109F1 Fire initiated in cabinet PK1-9 in the Unit 2 Cable Spreading Room (Fire Area 7-B). Fire damage is limited to that which occurred prior to auto-suppression.
TB-12 1.09E-Z23CTS3F2 Transient fire 05 in the Unit 2 4160 V BusH Cable Spreading Room (Fire Area 23-C) with damage to thermoplastic and thermoset cable targets within the fire's ZOI 75 Enclosure PG&E Letter DCL-14-11 0 Description of Fire-induced failures for Scenario Fire induced spurious Phase B signal resulting in isolation of RCP thermal barrier cooling supply valves and RWST draindown resulting in loss of Seal Injection.
RCPs tripped but SDS fail to actuate due to random failures
- resulting in a Seal LOCA (Rhodes Model). RHR pumps unavailable for sump recirculation cooling due to fire damage of RHR pump trains. Fire induced loss of power from startup transformer, and both redundant CCW auto-start pressure switches resulting in the loss of CCW. RCP tripped but SDS fail to actuate due to random failures resulting in a Seal LOCA (Rhodes Model). Note that the risk offset value in the corresponding Unit 1 Fire Area TB-6 is less significant since the corresponding Unit1 fire scenario has fire damage to only one of the two redundant CCW auto-start pressure switch. CCW is not Risk Fire Description Fire Area Offset Scenario of Fire (CDF)* Scenario 6-B-2 3.41E-ZBTC22F2 Fire initiated at 06 the Battery Charger BTC-2-BTC22-0P in the Unit 2 125 VDC Bus GRoom (Fire Area 6B2). Manual fire suppression successful and fire damage limited to the fire's maximum ZOI. 5-B-2 2.49E-Z5B2TG10F1 Fire initiated in 06 cabinet THG10 in the Unit 2 480 V Bus G MCC Room (Fire Area 5-B-2) with fire damage limited to the fire's ZOI. Enclosure PG&E Letter DCL-14-11 0 Description of Fire-induced failures for Scenario disabled by fire impact and CCW flow is still available to the* RCP seals -therefore reducing the probability of a seal LOCA. Fire induced failures of RCP thermal barrier cooling supply valves, and Seal Injection valve FCV-128. RCPs tripped but SDS fail to actuate due to random failures resulting in a Seal LOCA. No charging due to pump random failure and loss of support, with fire induced failure of Sl. Fire induced failures of RCP thermal barrier cooling supply valves and Charging Pump 2-1. Fire induced loss of power supply to Charging Pumps 2-2 and 2-3 resulting in loss of RCP seal injection.
RCPs tripped but SDS fail to actuate due to random failures resulting in a Seal LOCA (Rhodes Model).
- The Unit 2 fire areas with dominant LERF offset are similar to those for CDF offset 76 Enclosure PG&E Letter DCL-14-11 0 (d) There are no specific assumptions made that significantly contribute to the. post transition (variant) and the compliant case risk. NRC PRA RAI21: LAR Attachment V, Table V-1 presents the results of a sensitivity study on the updated fire ignition bin frequencies provided in NUREG/CR-6850, Supplement 1 (i.e., FAQ 08-0048) using the mean of the fire frequency bins contained in Section 6 of NUREG/CR-6850 for those bins with an alpha value less than or equal to one. It is not clear why the percent increase for LlCDF and LlLERF is higher than the percent increase for the total fire CDF and LERF (e.g., the percent increase in CDF for Unit 2 is 57%) while the percent increase in LlCDF for Unit 2 is 74°/o), given that fire ignition frequencies impact both the compliant and post-transition plant case accident sequences the same, and do not affect CCDP and CLERP values. Please provide the following:
a) An explanation of the anomaly cited above and whether the reported in-risk values in LAR Attachment V, Table V-1 are correct. b) An updated sensitivity study based on the integrated analysis performed in response to PRA RAI 3. Include in the sensitivity study any adjustments needed to correct anomalous results from the initial sensitivity study. c) An indication of whether the acceptance guidelines of RG 1.17 4 may be exceeded when this sensitivity study with respect to FAQ 08-0048 is applied to the integrated study of PRA RAI 3. If these guidelines may be exceeded, provide a description of fire protection, or related, measures that can be taken to provide additional defense in depth, as discussed in FAQ 08-0048. PG&E Response:
a) The difference in the percent increase between the total fire CDF (and LERF) and the transition LlCDF (and LlLERF) reported in Table V-1, Attachment V of the DCPP NFPA 805 LAR submittal is not an anomaly. The reported change in risk values in Table V-1, which were based on the Fire PRA model used for the LARin June 2013, are correct. A detailed review of the results of the sensitivity study, which were the basis for the reported values in Table V-1, is performed to identify the sources of this difference (e.g., the percent increase in CDF for Unit 2 is 57 percent while the percent increase in LlCDF for Unit 2 is 7 4 percent).
The review focuses on the difference observed in Unit 2 CDF and LlCDF as similar differences were observed in Unit 1 as well as for LERF and LlLERF in both units. 77 Enclosure PG&E Letter DCL-14-11 0 The review shows that a higher percent increase in LlCDF compared to the total CDF in Unit 2 sensitivity study is mostly due to conservatisms in the methods introduced as part of: (1) the sensitivity study, and (2) the transition risk calculation (i.e., Fire Risk Evaluation (FRE)). In the sensitivity study on CDF, each fire scenario in the FPRA was mapped to a fire ignition bin (with an alpha of less than or equal to 1 in Table 2-2 of EPRI 1 016735) in most part based on its fire modeling information such as initiating fire source (e.g., Batteries, MCB, Electrical Cabinet, etc.). The relative increase in CDF over the post-transition FPRA was then calculated based on this mapping information.
This mapping information, however, could not be directly applied to some of the fire scenarios considered in .LlCDF calculation.
In the FRE and therefore in .LlCDF calculation of a fire area, the fire scenario(s) with low CDF were grouped into a "pseudo" scenario.
The transition change in risk (LlCDF) from such scenarios was estimated by conservatively assuming the zero CDF value for the risk contribution from the compliant model; that is the .LlCDF represents the GDF value of the post-transition model. Because a pseudo scenario normally encompasses multiple scenarios with potentially different ignition sources, a single ignition bin representing the pseudo scenario could be selected or mapped to it. As a conservative approach, this pseudo scenario was mapped to the ignition bin with the highest frequency adjustment factor (i.e., Bin 31). This pseudo scenario and its treatment were unique to the .LlCDF calculation and results in a higher percent increase over that of the total CDF. A conservative estimation of the .LlCDF in the FREs as discussed above, which assumes the zero CDF for the compliant model for those pseudo scenarios, further magnified the difference in the percent increase.
Without this assumption of the zero CDF for the compliant model, the CDF of the compliant model would have been a non-zero value estimated with the same ignition bin (i.e., Bin 31) applied to the post-transition model. With the higher CDF for the compliant model, the resulting .LlCDF would have been smaller than that reported in the sensitivity study and the difference in the percent increase between the CDF and .LlCDF would have been closer. In summary, a higher percent increase for .LlCDF and LlLERF is accounted for and is not an anomaly. There is no adjustment or correction to be made to the aggregate model in response to PRA RAI-21 (a). b) Once the results (i.e., the total transient CDF, LERF, LlCDF, LlLERF) of the aggregate analysis in response to PRA RAI-03 is available, a sensitivity study on the updated fire ignition bin frequencies provided in NUREG/CR-6850, 78 Enclosure PG&E Letter DCL-14-11 0 Supplement 1 (i.e., FAQ 08-0048) will be performed.
A review of the results from the initial sensitivity study in response to PRA RAI-21 (a) shows that a higher percent increase for LlCDF and LlLERF reported in the initial sensitivity study as compared to CDF and LERF is not an anomaly. This review identified no corrections or adjustments that need to be made to the aggregate analysis or the sensitivity study. c) The results of the updated sensitivity study in response to PRA RAI-21 (b) will be compared to the acceptance guidelines of RG 1.17 4. If these guidelines are exceeded, the results of the sensitivity study will be reviewed to obtain risk insights such as dominant fire areas, fire scenarios or accident sequences, and differences in risk profile between the sensitivity study cases. Based on these risk insights, the existing defense-in-depth measures will be reviewed if they provide adequate measures to counter potential risk increase, or additional defense-in-depth measures will be identified if the existing measures are determined to be inadequate.
NRC PRA RAI 22: During the audit, the licensee explained that the analysis could better address the timing of the fire scenarios that are supported by thermal-hydraulic analysis and the resulting cues for operator actions. As a result, please evaluate where abandonment due to loss of control is credited and provide an assessment of MCR abandonment through establishing a bounding scenario or a set of representative scenarios.
Ensure the potential complexity of fire-induced damage, including spurious operations, are incorporated into these scenarios.
Evaluate the timing supported by thermal hydraulics for these scenarios and the effect on the HRA including the effect on cues. Provide a comparison of the results of this analysis with the values from the PRA supporting the LAR, and explain any potential differences.
PG&E Response:
The discussion of where abandonment due to loss of control is credited in the model, and how the complexity of fire-induced damage incorporated into these scenarios, is addressed in the response to PRA RAI-01.h.
The response to PRA RAI-22 focuses on the HRA portion of the MCR abandonment.
An assessment of the bounding scenarios will be provided in PRA RAI-03, including the results of the calculation of CCDP and comparison to the LAR values. This analysis is an update to the MCR abandonment evaluation performed to support the LAR and addresses review comments from the NRC. For the case of abandonment on loss of control, a detailed HRA is being conducted using NUREG-1921 accepted methods to evaluate Human Error Probabilities 79 Enclosure PG&E Letter DCL-14-11 0 (HEPs) for the range of loss of habitability and loss of control scenarios identified in the response to PRA RAI 01.h. The loss of control HEP will include the cognitive error for the decision to abandon the MCR, as well as the execution of all actions required to achieve and maintain a safe and stable state. Since the issue is not whether the MCR is ever abandoned, but whether it is abandoned in time to successfully achieve remote shutdown using the procedures being developed for NFPA 805, a timeline is being established to assess the following key timing elements that provide input to the calculation of the HEP for the cognitive HFE:
- Thermal-Hydraulic (T -H) calculations performed in support of the PRA success criteria evaluation are being used to determine how long the operators have to evacuate the control room. The PRA evaluates the timing of various scenarios using T-H analyses and certain TCOAs have been identified in the post-fire SSD analysis with time requirements for their performance.
These T -H analysis and TCOAs are being reviewed as part of the MCR abandonment to determine the most applicable timing for the cognitive action for abandonment or System Time Window (Tsw). Considerations for the application of the timing are those scenarios that are most relevant to the abandonment scenarios.
These can be as specific as the minimum time that the operators are expected to establish control at the Hot Shutdown Panel (HSDP) in case of fire in the MCR or cable spreading room. Alternatively, they can be based on overall objectives such as the latest time at which the operators can -start AFW and still avoid core damage or the time to establish containment cooling, and subtracting the time required for execution actions from this total to estimate the time within which the decision to abandon must be taken. Among the T-H analyses that will be evaluated for relevance will be the following:
- 4. Establish feed and bleed for loss of feedwater scenarios without AFW 5. Recovery of power and AFW flow for Station Blackout (SBO) scenarios.
- "Tdelay" is the time from the start of the fire to the arrival of the cue that there has been a loss of control such that the MCR must be abandoned.
Operator interviews and walkthroughs will be used to determine how long it takes from receipt of a fire alarm to perform the visual confirmation of the fire location 80 Enclosure PG&E Letter DCL-14-11 0 and its severity.
By this time the Shift Manager will have the information needed to make the decision to abandon based on the fire confirmation and the fire induced failures (e.g., spurious actuations) that will have manifested themselves in the MCR.
- The DCPP CSR fire response procedure is in the process of being modified to specify the loss of control abandonment criteria.
Operators will be directed to the MCR abandonment procedure OP AP-8A for scenarios involving a loss of AFW, charging, electrical power or a stuck open PORV. Operator interviews will be used to determine the time taken for the diagnosis and decision-making (included in the HRA calculator as T1/2) once the delay time has ended and given these conditions.
- Additional timing for the issuance to operators of procedure attachments and direction to obtain keys to the operators who would be implementing actions at the HSDP or locally in the plant will be evaluated.
This will be obtained through timed simulator run-throughs by DCPP operations.
This will be included in the calculation of the cognitive HEP as part of what is usually termed the "manipulation time" (Tm) and represents the time between the decision to abandon and the actual start of the execution actions. The timing will be verified during the implementation phase (LAR AttachmentS, Table S-3, Items S-3.16 and S-3.17), prior to self-approval.
Simulator exercises to observe the diagnosis process will be conducted after the operators are trained on the new procedures.
The HRA Calculator uses this timing along with analyst evaluations based upon operator interviews and observations during walkthroughs as input information to calculate an HEP using several methods, primarily the Cause Based Decision Tree Method (CBDTM) and the Human Cognitive Reliability (HCR) method. The final HEP is generally based upon the highest value (unless the analyst feels there is a compelling reason to specify a particular method).
- In addition to the decision to abandon modeled by the cognitive portion of the HFE, the procedure-directed execution actions required to achieve and maintain a safe and stable state will then be evaluated using the execution portion of the HFEs for the range of scenarios discussed in the response to PRA RAI 01.h. This will be a reevaluation of the execution actions assessed for the MCR abandonment HRA results submitted with the LAR to incorporate any procedure changes or specific distinctions.
Consideration is given in these actions to the need to address spurious actuations of equipment, as covered in the procedures.
In addition, the overall time window for the execution actions will consider the T -H analyses and TCOAs discussed previously.
Timing for the manipulation portion of the actions will reflect simulator training insights.
81 Enclosure PG&E Letter DCL-14-11 0 Finally, it is noted that there is no specific guidance on performing the HRA and modeling for abandonment on loss of control. However, DCPP is confident that implementing the approach discussed above is within the current capabilities of both PRA modeling and HRA quantification techniques, and that this will constitute an acceptable analysis for these scenarios.
Further, it should be noted that the strategy being implemented by DCPP in the new procedures is designed to encourage abandonment in those cases where it affords the best opportunity to reach a safe, stable state. 82 ABS AC AED AFW AM SAC AOV APCSB ASDV ASME ASW ATWS BAST BTP 8CDF CCDP CCF CCP ccw CDF CFAST CFMLA CP CPT CR Acronym List ABS Consulting Inc. Alternating Current Automatic External Defibrillator Auxiliary Feedwater ATWS Mitigation System Actuation Circuitry Air Operated Valve Auxiliary Power Conversion Systems Branch Atmospheric Steam Dump Valve American Society of Mechanical Engineers Auxiliary Saltwater Anticipated Transient Without Scram Boric Acid Storage Tank . Branch Technical Position Delta Core Damage Frequency Conditional Core Damage Probability Common Cause Failure Centrifugal Charging Pump Component Cooling Water Core Damage Frequency Consolidated Model of Fire and Smoke Transport Circuit Failure Mode Likelihood Analysis Carbon Dioxide Casualty Procedure Control Power Transformer Control Room 1 Enclosure Attachment 1 PG&E Letter DCL-14-11 0
CSR CST DCM DCPP DFOTP DG DID ECCS ECG EDG EMT EPRI ERFBS F&B F&Os FAQ FDS FC FCV FDT FHB FIVE FM FP FPE FPRA FRE Cable Spreading Room Condensate Storage Tank Design Criteria Memorandum Diablo Canyon Power Plant Diesel Fuel Oil Transfer Pump Diesel Generator Defense-in-Depth Emergency Core Cooling System Equipment Control Guideline Emergency Diesel Generator Electrical Metallic Tubing Electric Power Research Institute Electrical Raceway Fire Barrier Systems Feed and Bleed Facts and Observations Frequently Asked Question Fire Dynamics Simulator Fire Compartment Flow Control Valve Fire Dynamics Tools Fuel Handling Building Fire Induced Vulnerability Evaluation Fire Modeling Fire Protection Fire Protection Engineering Fire Probabilistic Risk Assessment Fire Risk Evaluation 2 Enclosure Attachment 1 PG&E Letter DCL-14-11 0
FWST HEAF HFE HGL HSDP HRA HRR HVAC lA lAS IC lEPRA ISLOCA kPa KSF kW LAR 8LERF LERF LOCA LLOCA LTCW MAAP MCA MCB MCC MCR Feedwater Storage Tank High Energy Arching Fault Human Factors Engineering Hot Gas Layer Hot Shutdown Panel Human Reliability Analysis Heat Release Rate Heating, Ventilation, and Air Conditioning Instrument Air Instrument Air System Incident Commander Internal Events Probabilistic Risk Assessment Interfacing System Loss of Coolant Accident Kilo pascals Key Safety Function Kilowatt License Amendment Request Delta Large Early Release Frequency Large Early Release Frequency Loss of Coolant Accident Large Loss of Coolant Accident Long Term Cooling Water Modular Accident Analysis Program Multi-Compartment Analysis Main Control Board Motor Control Center Main Control Room 3 Enclosure Attachment 1 PG&E Letter DCL-14-110 MDAFW MFW MLOCA MOV MQH MWS NEC NFPA NPO NRC NSCA NSPC OR PAU PFP PG&E PORV PRA PVC RCA RA RAFA RAI RCP RCS RG RHR Motor Driven Auxiliary Feedwater Main Feedwater Medium Loss of Coolant Accident Motor Operated Valve McCaffrey, Quintiere, and Harkleroad Makeup Water System National Electric Code National Fire Protection Association Non-Power Operation or Non-Power Operational Nuclear Regulatory Commission Nuclear Safety Capability Assessment Nuclear Safety Performance Criteria Operations Responder Physical Analysis Unit Pre-Fire Plans Pacific Gas and Electric Company Power-Operated Relief Valve Probabilistic Risk Assessment Po lyvi nylch lo ride Radiologically Controlled Area Recovery Action Recovery Action Feasibility Assessment Request for Additional Information Reactor Coolant Pump Reactor Coolant System Regulatory Guide Residual Heat Removal 4 Enclosure Attachment 1 PG&E Letter DCL-14-11 0
RIPB RPE RTI RWSR RWST SAT SDS SFPE SG SLOCA SR SSA sse SSD SSPS SWGR TCOA TDAFW TG TH UFSAR VAC VFDR V&V ZOI Risk-Informed, Performance-Based Replacement Parts Evaluation Response Time Index Raw Water Storage Reservoir Refueling Water Storage Tank Systematic Approach to Training Shutdown Seals Society of Fire Protection Engineers Steam Generator Small Loss of Coolant Accident Supporting Requirements Safe Shutdown Analysis Systems, Structures, and Components Safe Shutdown Solid State Protection System Switchgear Time-Critical Operator Actions Turbine Driven Auxiliary Feedwater Turbine Generator Thermal Hydraulic Updated Final Safety Analysis Report Volts Alternating Current Variance from Deterministic Requirements Verification and Validation Zone of Influence 5 Enclosure Attachment 1 PG&E Letter DCL-14-110