ML14154A496
| ML14154A496 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 05/30/2014 |
| From: | James Shea Tennessee Valley Authority |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML14154A495 | List: |
| References | |
| L44 140530 002, TAC MF1185, TAC MF1186, TAC MF1187 | |
| Download: ML14154A496 (19) | |
Text
Security-Related Information - Withhold from Public Disclosure in accordance with 10 CFR 2.390.
Attachments 2, 3, and 4 of the enclosure contain Security-Related Information. Upon removal of Attachments 2, 3, and 4 from the enclosure, this letter is uncontrolled.
L44 140530 002 1101 Market Street, Chattanooga, Tennessee 37402 CNL-14-067 May 30, 2014 10 CFR 50.90 10 CFR 2.390 ATTN: Document Control Desk U.S. Nuclear Regulatory Commission Washington, D.C. 20555-0001 Browns Ferry Nuclear Plant, Units 1, 2, and 3 Renewed Facility Operating License Nos. DPR-33, DPR-52, and DPR-68 NRC Docket Nos. 50-259, 50-260, and 50-296
Subject:
Update to the License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (TAC Nos. MF1185, MF1186, and MF1187)
References:
- 1. Letter from TVA to NRC, "License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants (2001 Edition) (Technical Specification Change TS-480)," dated March 27, 2013 (ADAMS Accession No. ML13092A393)
- 2. Letter from TVA to NRC, "Response to NRC Request to Supplement License Amendment Request to Adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants for the Browns Ferry Nuclear Plant, Units 1, 2, and 3 (TAC Nos.
MF1185, MF1186, and MF1187)," dated May 16, 2013 (ADAMS Accession No. ML13141A291)
By letter dated March 27, 2013 (Reference 1), Tennessee Valley Authority (TVA) submitted a license amendment request (LAR) for Browns Ferry Nuclear Plant (BFN), Units 1, 2, and 3, to transition to National Fire Protection Association Standard (NFPA) 805. In addition, by letter dated May 16, 2013 (Reference 2), TVA provided information to supplement the Reference 1 letter.
The enclosure to this letter provides an update to portions of the NFPA 805 LAR consisting of marked up pages reflecting changes identified by TVA, with descriptions and justifications for each change. Attachments 2, 3, and 4 to the enclosure contain security-related information and should be withheld from public disclosure under 10 CFR 2.390.
U.S. Nuclear Regulatory Commission Page 2 May 30, 2014 Consistent with the standards set forth in Title 10 of the Code of Federal regulations (10 CFR), Part 50.92(c), TVA has determined that the additional information, as provided in this letter, does not affect the no significant hazards consideration associated with the proposed application previously provided in Reference 1.
There are no regulatory commitments contained in this submittal. Please address any questions regarding this submittal to Mr. Edward D. Schrull at (423) 751-3850.
I declare under penalty of perjury that the foregoing is true and correct. Executed on this 30th day of May 2014.
JiW.jBhea fee President, Nuclear Licensing
Enclosure:
NFPA 805 License Amendment Request Update cc (Enclosure):
NRC Regional Administrator- Region II NRC Project Manager - Browns Ferry Nuclear Plant NRC Senior Resident Inspector- Browns Ferry Nuclear Plant State Health Officer, Alabama State Department of Health
ENCLOSURE Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2, and 3 NFPA 805 License Amendment Request Update As a result of continuing reviews and implementation activities, Tennessee Valley Authority (TVA) has determined that several portions of the National Fire Protection Association Standard (NFPA) 805 License Amendment Request (LAR) must be revised to update or correct information provided in the LAR. Each issue has been entered into the TVA Corrective Action Program for evaluation and disposition. Attachment 1 to this enclosure provides descriptions, justifications and marked up pages related to LAR Attachment A, "NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements." Attachment 2 to this enclosure provides descriptions, justifications and marked up pages related to LAR Attachment C, "NEI 04-02 Table B Fire Area Transition," Table C-1, "NFPA 805 Ch 4 Compliance (NEI 04-02 Table B-3)." Attachment 3 to this enclosure provides descriptions, justifications and marked up pages related to LAR Attachment G, "Recovery Actions Transition." Attachment 4 to this enclosure provides descriptions, justifications and marked up pages related to LAR Attachment S, "Modifications and Implementation Items," Table S-2, "Plant Modifications Committed." Attachment 5 to this enclosure provides descriptions, justifications and marked up pages related to LAR Attachment U, "Internal Events PRA Quality," Table U-1, "Internal Events PRA Peer Review - Facts and Observations."
Attachments 2, 3, and 4 of the enclosure contain Security-Related Information.
E-1
ATTACHMENT 1 TO ENCLOSURE Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2, and 3 NFPA 805 License Amendment Request Attachment A Update Change TVA-1
==
Description:==
LAR, Attachment A, "NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program &
Design Features," NFPA 805 Ch. 3 References for 3.9.1 (1) NFPA 13, Standard for the Installation of Sprinkler Systems are revised to correct the Revision number of MDQ099920100005, "NFPA 13 Code Compliance Evaluation - 1985 Edition," from Revision 1 to Revision 2.
Justification:
TVA determined that code compliance reviews conducted to support the NFPA 805 LAR submittal failed to review the Diesel Generator Buildings' sprinkler systems (Problem Evaluation Report (PER) 744029). A code review of the sprinkler systems applicable to NFPA 13 was performed to support the LAR. However, the NFPA 13 Code Compliance Evaluation - 1985 Edition calculation did not include the sprinkler systems in the Unit 1/2 and Unit 3 Diesel Generator Building Pipe and Electrical Tunnels. The drawings indicate that Unit 1 has 8 heads while Unit 3 has 24 heads. The Unit 3 system is located in an area where there is a water spray suppression system also installed. This water spray system for Unit 3 was included in a separate code compliance review.
The NFPA 13 Code Compliance Evaluation - 1985 Edition calculation has been appropriately revised to include fire suppression systems protecting the Unit 1/2 and Unit 3 Diesel Generator Building Pipe and Electrical Tunnels. The revised calculation includes a complete review with respect to the existing systems applicable to the code. The change to LAR Attachment A is required to demonstrate that the systems have been properly evaluated per specific requirements of NFPA 13 code.
Affected Pages Please see the mark up provided on Page E-4.
E-2
Change TVA-2
==
Description:==
Reference 3.10.9 in LAR, Attachment A, "NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements," Requirements/Guidance is revised to replace "MDQ099920100004 Rev. 0 [Supplement Section C.6.e] - NFPA-12 Code Compliance Evaluation," with "1992-1-15 [Enclosure 2 Section C.6.e] - BFN-NRC, Browns Ferry Nuclear Plant (BFN) - Fire Protection Report (FPR)."
Justification:
LAR, Attachment A incorrectly referenced MDQ099920100004, Rev. 0 [Supplement Section C.6.e] - NFPA-12 Code Compliance Evaluation (PER 860601). The specific concern of LAR, Attachment A, Requirements/Guidance 3.10.9 (i.e., thermal shock damage of gaseous fire suppression systems) is not contained in the NFPA-12 Code Compliance Evaluation. However, thermal shock damage is addressed in TVA Letter dated January 15, 1992, "Browns Ferry Nuclear Plant (BFN) - Fire Protection Report (FPR)," Enclosure 2, Section C.6.e. Therefore, the reference to MDQ099920100004 is replaced with a reference to the TVA letter dated January 15, 1992. This is a change in reference only, and does not affect compliance strategies or conclusions made in the LAR.
Affected Pages Please see the mark up provided on Page E-5.
E-3
Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Compliance NFPA 805 Ch. 3 Reference Requirements / Guidance Statement Compliance Basis 3.9.1 [Fire Suppression System 3.9.1* N/A General requirements. The requirements of this Section are addressed in Code Requirements] If an automatic or manual water-based fire suppression system is required Sections 3.9.1(1) through 3.9.1(4).
to meet the performance or deterministic requirements of Chapter 4, then the system shall be installed in accordance with the appropriate NFPA standards including the following:
3.9.1 [Fire Suppression System 3.9.1 (1) NFPA 13, Standard for the Installation of Sprinkler Systems Complies with Sprinkler systems at BFN are evaluated to be in compliance with NFPA 13 Code Requirements] (1) Use of EEEEs - 1985, 1987, 1991 and 2002 editions as shown in the referenced Code Compliance Evaluations.
Partial suppression is provided for fire zones 01-01, 01-02, 01-03, 01-04, 02-01, 02-02, 02-03, 02-04, 03-01 and 03-02 in the Reactor Building.
These sprinkler systems are adequate as evaluated in MDQ099920110009.
See Table C-2 of the LAR for required systems.
Item for Implementation:
Corrective actions were identified in the Code Compliance Evaluations.
These corrective actions are identified in Modifications 98, 99, 100 and 2 101 in Table S-2 of Attachment S and Implementation Item 20 in Table S-3 of Attachment S.
References Document ID TVA-1 MDQ099920100005 Rev. 1 [All] - NFPA-13 Code Compliance Evaluation - 1985 Edition MDQ099920110001 Rev. 1 [All] - NFPA-13 Code Compliance Evaluation - 1987 Edition MDQ099920110002 Rev. 1 [All] - NFPA-13 Code Compliance Evaluation - 1991 Edition MDQ099920110003 Rev. 1 [All] - NFPA-13 Code Compliance Evaluation - 2002 Edition MDQ099920110009 Rev. 1 - NFPA-805 Transition - Fire Area Designation 3.9.1 [Fire Suppression System 3.9.1 (2) NFPA 15, Standard for Water Spray Fixed Systems for Fire Complies with Water spray systems at BFN are evaluated to be in compliance with NFPA Code Requirements] (2) Protection Use of EEEEs 15 - 1985, or NFPA 15 - 2001 as shown in the referenced Code Compliance Evaluations.
See Table C-2 of the LAR for required systems.
Item for Implementation:
Corrective actions were identified in the Code Compliance Evaluations.
These corrective actions are identified in Modification 102 in Table S-2 of Attachment S and Implementation Item 21 in Table S-3 of Attachment S.
References Document ID MDQ099920100007 Rev. 1 [All] - NFPA-15 Code Compliance Evaluation - 1985 Edition MDQ099920110004 Rev. 1 [All] - NFPA-15 Code Compliance Evaluation - 2001 Edition Fire Safety Analysis Data Manager (4.129) TVA Browns Ferry Run: 03/23/2013 10:31 Page: 44 of 55 BFN Units 1, 2, and 3 NFPA 805E-4 Transition Report, Page 120 of 1661
Attachment A NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements Compliance NFPA 805 Ch. 3 Reference Requirements / Guidance Statement Compliance Basis References Document ID MDQ099920100004 Rev. 0 [Attachment E; Section 4.3.3.4] - NFPA-12 Code Compliance Evaluation 3.10.9 [Gaseous Suppression 3.10.9 Complies The possibility of thermal shock was considered in the design of the CO2 System Cooling The possibility of secondary thermal shock (cooling) damage shall be fire suppression systems.
Considerations] considered during the design of any gaseous fire suppression system, but particularly with carbon dioxide.
1992-1-15 [Enclosure 2 Section C.6.e] - BFN-NRC, Browns Ferry Nuclear Plant (BFN) - Fire Protection Report (FPR)
References Document ID MDQ099920100004 Rev. 0 [Supplement Section C.6.e] - NFPA-12 Code Compliance Evaluation TVA-2 3.10.10 [Gaseous Suppression 3.10.10 N/A Carbon Dioxide is the product of decomposition and will not react with the System Decomposition Issues] Particular attention shall be given to corrosive characteristics of agent atmosphere and form corrosive products.
decomposition products on safety systems.
According to the NFPA Fire Protection Handbook, 2008 edition, CO2 does not leave residue. The lack of residue eliminates the possibilities for corrosion on equipment in areas protected by CO2 suppression systems.
References Document ID NFPA Fire Protection Handbook Rev. 2008 Ed. [Chapter 1, Section 17] -
3.11 Passive Fire Protection 3.11 Passive Fire Protection Features. N/A Section Heading.
Features. This section shall be used to determine the design and installation requirements for passive protection features. Passive fire protection features include wall, ceiling, and floor assemblies, fire doors, fire dampers, and through fire barrier penetration seals. Passive fire protection features also include electrical raceway fire barrier systems (ERFBS) that are provided to protect cables and electrical components and equipment from the effects of fire.
3.11.1 Building Separation. 3.11.1 Building Separation. Complies with All major buildings within the power block are separated from each other Each major building within the power block shall be separated from the Use of EEEEs by barriers having a fire resistance rating of 3-hours, or are evaluated to others by barriers having a designated fire resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> or be equivalent to a 3-hour rating.
by open space of at least 50 ft (15.2 m) or space that meets the requirements of NFPA 80A, Recommended Practice for Protection of BFN utilizes the exception to this section.
Buildings from Exterior Fire Exposures.
Exception: Where a performance-based analysis determines the Evaluation MDQ099920110009 documents the acceptability of the adequacy of building separation, the requirements of 3.11.1 shall not separation of the Refuel Floor from the Reactor Building, Control Building, apply. and Turbine Building, the separation of the Transformers in the Yard from adjacent structures (Turbine and Reactor Buildings), the separation of the chillers from the Unit 1 and 2 Diesel Generator Building, and the separation of the miscellaneous structures in the yard and buildings containing safe shutdown equipment.
References Document ID 0-FPR-VOLUME 1/PART 2 Rev. 14 [Section 6.0] - The Fire Protection Report, Fire Hazards Analysis MDQ099920110009 Rev. 1 - NFPA-805 Transition - Fire Area Designation Fire Safety Analysis Data Manager (4.129) TVA Browns Ferry Run: 03/23/2013 10:31 Page: 49 of 55 BFN Units 1, 2, and 3 NFPA 805E-5 Transition Report, Page 125 of 1661
ATTACHMENT 5 TO ENCLOSURE Tennessee Valley Authority Browns Ferry Nuclear Plant, Units 1, 2, and 3 NFPA 805 License Amendment Request Attachment U, Table U-1 Update Security-Related Information Change TVA-9
==
Description:==
LAR, Attachment U, Table U-1, "Internal Events PRA Peer Review - Facts and Observations," is revised to update the referenced revision number for NDN00099920070032, "HR- BFN Probabilistic Risk Assessment - Human Reliability Analysis," from Revision 2 to Revision 3.
Justification:
The supporting calculation (i.e., TVA Fire PRA - Task 7.5 Fire-Induced Risk Model) for LAR, Attachment U, Table U-1, contained an incorrect revision number for the HR - BFN Probabilistic Risk Assessment - Human Reliability Analysis in multiple locations. The incorrect revision number was carried forward into LAR Attachment U, Table U-1 (PER 776207). The TVA Fire PRA - Task 7.5 Fire-Induced Risk Model has been revised to correct the revision number for the HR - BFN Probabilistic Risk Assessment - Human Reliability Analysis. This change corrects the LAR to refer to the correct HR - BFN Probabilistic Risk Assessment - Human Reliability Analysis revision.
Affected Pages Please see the mark ups provided on Pages E-27 through E-37.
E-26
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations HR-C3 Closed 2-13 In Table B-1 of the HRA Basis for Identified CCF HFLs without screening values: No impact.
HR-D5 Notebook, HFL_1003_LT56A Significance: HFL_1003CCFLT0056, HFL_1003CCFLT0058, has a value of 9E-04 which is Given that the HFL_1003CCFLT0203, HFL_1068CCFPTLOPR, HR-D7 higher than the component miscalibration has a HFL_2003CCFLT0056, HFL_2003CCFLT0058, This change to the SY-A15 failure of the same level higher value than the HFL_2003CCFLT0203, HFL_2068CCFPTLOPR, internal events fault transmitter yet it is not in the mechanical failure it HFL_3003CCFLT0056, HFL_3003CCFLT0058, tree was done prior to fault tree based on the should be included HFL_3003CCFLT0203, HFL_3068CCFPTLOPR. the development of common cause failure of all 4 as a valid failure the BFN Fire PRA.
level transmitters being in the more in the tree. One The independent miscalibration events fault tree (note 1 in table). The level transmitter associated with each of these common cause independent miscalibration failing due to a failure events have been added to the fault tree.
should be included in the fault hardware issue and a Table B-1 of the TVA Calculation, Revision 3 tree. This is applicable to second due to NDN00099920070032 Revision 2, HR - BFN TVA-9 other precursor events also. miscalibration is a Probabilistic Risk Assessment - Human valid Reliability Analysis has been updated to include changes to the PRA model.
Possible Resolution:
Add the independent miscalibration events to the fault tree HR-D5 Closed 2-14 HFL_1003_CCFT0056 is Basis for The F&O relates to all of the pre-initiators that No impact.
HR-C3 Common cause miscalibration Significance: accounted for common miscalibration errors. Revision 3 of all 4 level transmitters, The pair CCFs will Fault trees have been updated and TVA inspection of the fault tree have a higher value Calculation, NDN00099920070032 Revision 2, The pre-initiators are TVA-9 shows that specific pairs of than the 4 of 4 event HR - BFN Probabilistic Risk Assessment - modeled in the same failures (AC, BD) would also thus impact the Human Reliability Analysis has been revised to manner for the fire cause a failure to initiate the results. reflect this change. HFL_1003_LT56A, PRA.
logic. These CCF pairs should HFL_1003_LT56B, HFL_1003_LT56C, and be added to the model. This Possible Resolution: HFL_1003_LT56D have been added to the will apply to other Calculate the pair model.
miscalibration CCFs also. CCFs and add to the fault tree Page U-15 BFN Units 1, 2, and 3 NFPA 805 E-27 Transition Report, Page 1356 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations LE-C2 Closed 2-38 The operator actions in the Basis for LERF HFEs have been updated in a manner No impact.
LE-C7 LERF analysis are not based Significance: consistent with the process used for Level 1 on that same type of HFE SR requires the HFEs and are documented in the TVA calculations used in the Level same level of rigor in Calculation, NDN00099920070032 Revision 2, The Fire PRA uses TVA-9 1 analysis HRA as in level 1. HR - BFN Probabilistic Risk Assessment - the same Level 2 Human Reliability Analysis. model.
Possible Resolution:
Use the same HRA Revision 3 process as Level 1 for the LERF HFE events.
LE-D7 Closed 2-39 In the documentation for CIL it Basis for TVA Calculation, NDN00099920070037 No impact.
states the fault tree is Significance: Revision 0, LE.01 - BFN Probabilistic Risk quantified and the resulting Not describing the Assessment - LERF Analysis (Appendix A) has value is used in the actual method of been revised to correctly reflect the actual model This is a comment on quantification of the node. quantifying the node and also better reflect the information in the TVA completeness of the Inspection of the fault tree can lead to errors in Calculation, NDN00006420070018 Revision 1, BFN Internal Events shows that the containment use of the PRA. SY.11 - BFN Probabilistic Risk Assessment - PRA documentation.
isolation fault tree is Primary Containment Isolation System. This has no effect on quantified with the node Possible Resolution: the structure, directly. Direct quantification Correct the CIL quantification, or of node is the appropriate writeup in the LERF results of the BFN action. notebook to correctly Fire PRA.
reflect the actual model and also better reflect the information in the Primary Containment Isolation notebook.
LE-C6 Closed 2-41 Systems models are not Basis for Systems models are now developed for LERF. No impact.
developed for LERF. Significance: The LERF Analysis documentation has been Documentation indicates split Systems models are revised to reflect the updated to include fraction values with no good needed to properly descriptions of the LERF system models. The Fire PRA uses basis for them. reflect impact of the same Level 2 specific failures. It is model.
believed that the values being used Page U-19 BFN Units 1, 2, and 3 NFPA 805 Transition E-28 Report, Page 1360 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations HEPs of less than 1E-7, and and overall results dependent failure modes that are not usually Fire PRA.
eight are less than 1E-6. may be artificially treated. The HRA Calculator currently provides Note that the HRA lowered, and the the capability to explicitly calculate the joint acknowledges these low importance of the probability of dependent and independent post-combined HEPs, but does not operator actions may initiator HFEs in the same accident enforce any lower bound. be understated. sequence/cutset: This methodology Further, it states that a improvement reduces the need for a threshold sensitivity will be performed in Possible Resolution: value. Overly conservative threshold values the Quantification Notebook, Establish a have the potential for skewing the results.
but none is performed. reasonable lower bound for combined HFE probabilities.
Perform sensitivities to determine the significance of this lower bound.
HR-G1 Closed 4-23 Several operator actions that Basis for Detailed analysis has been performed for HFAs No impact.
have RRW > 1.005 have Significance: with Risk Reduction Worth (RRW) > 1.005 and HEPs with screening values. These HFEs should results are documented in the TVA Calculation, The Fire PRA has its The HFEs are: be evaluated using a NDN00099920070032 Revision 2, HR - BFN own Human TVA-9 HFAZ0074ALIGN_DWS detailed analysis in Probabilistic Risk Assessment - Human (CDF/LERF), accordance with the Reliability Analysis. Reliability Analysis HFAZ0023IFISOL (CDF), requirements of HR- (NDN0009992012000 HFAZ0084CADALIGN (CDF), G1. Revision 3 01, 1 TVA FIRE PRA HFAZ0_SPRAYMLOCA - Task 7.12 Post-Fire (LERF), HFAZ0HCIINIT30 Possible Resolution: Human Reliability (LERF), and Perform a detailed Analysis) and does HFAZ0071CTLPOWER analysis of all HFEs not use the values (LERF) with RRW >1.005. generated in the Internal Events PRA.
Page U-37 BFN Units 1, 2, and 3 NFPA 805E-29 Transition Report, Page 1378 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations HR-F2 Closed 4-25 There are many operator Basis for Significance HFEs have been reviewed and detailed analyses No impact.
HR-G4 actions that use screening Without any real have been performed for many HFEs that values; see Table 8 of the timing information, it previously used screening values. In addition, HR-G5 HRA. None of these actions is not possible to timing analyses have been reviewed. Timing is The Fire PRA has its appear to use any information estimate, even at a based primarily on plant specific MAAP own Human to base the time available and screening level, the calculations, timing from BFN simulator Reliability Analysis the times to operator cues probability of exercises, or estimates from BFN operator (NDN0009992012000 and perform the actions are operator failure or interviews. In response to this comment, 01, 1 TVA FIRE PRA not documented. success. updated timing analysis has been re-reviewed by - Task 7.12 Post-Fire BFN operations staff and additional changes Human Reliability Possible Resolution: have been incorporated. Analysis) and does Provide timing All model changes are included in an update to not use the values information for all the TVA Calculation, NDN00099920070032 generated in the Internal Events PRA.
operator actions, Revision 2, HR - BFN Probabilistic Risk TVA-9 including those HEPs Assessment - Human Reliability Analysis.
estimated by using screening values. Revision 3 HR-C1 Closed 4-27 There are many "Misaligned Basis for The HFE HEP codes noted in the F&O were No impact.
HFE HEP Codes" assigned in Significance: used in the previous model and were Appendix A of the HRA that The disposition of inadvertently left in the documentation.
Revision 3 are not carried through the HFEs for non- Appendix A to the TVA Calculation, This is a comment on completeness of the rest of the HRA or present in screened potential NDN00099920070032 Revision 2, HR - BFN BFN Internal Events TVA-9 the PRA model (e.g., misalignment events Probabilistic Risk Assessment - Human HARCI1, HAREA1, HAINH1, cannot be verified as Reliability Analysis has been revised to correct PRA documentation.
and HARHR2). required by HR-C1. errors and provide traceability. This has no effect on The PRA group the structure, indicated that the quantification, or Appendix would be results of the BFN updated. Fire PRA.
Possible Resolution:
Provide traceability from Appendix A of the HRA to the remainder of the pre-initiator analysis and the PRA model.
Page U-38 BFN Units 1, 2, and 3 NFPA 805 E-30 Transition Report, Page 1379 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations procedure. review of the procedures for all three units is warranted. There should at least be a focus on procedures for systems that may be different between the units.
HR-A3 Closed 4-31 There do not appear to be Basis for Activities from HR-A1 and HR-A2 that affect No impact.
any ACTIVITIES that were Significance: redundant trains or diverse systems are found in HR-A1 and HR-A2 HR-A3 requires identified in Table B-1 of Appendix B in the TVA Revision 3 identified as affecting identification of such Calculation, NDN00099920070032 Revision 2, This is a comment on TVA-9 redundant trains or diverse activities, despite the HR - BFN Probabilistic Risk Assessment - completeness of the systems. fact that the HFEs Human Reliability Analysis under the heading BFN Internal Events may include multiple "Common cause events." These activities are all PRA documentation.
components or a result of miscalibration events. This has no effect on trains. the structure, quantification, or Possible Resolution: results of the BFN Identify and Fire PRA.
document activities from HR-A1 and HR-A2 that affect redundant trains or diverse systems.
Page U-40 BFN Units 1, 2, and 3 NFPA 805 E-31 Transition Report, Page 1381 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations HR-H3 Closed 4-40 A review of non-significant Basis for Dependency analysis has been re-performed No impact.
QU-D5 cutsets found many LOOP Significance: and results are documented in the TVA Revision 3 cutsets that have This is an example of Calculation, NDN00099920070032 Revision 2, The Fire PRA has its TVA-9 combinations of two non-significant HR - BFN Probabilistic Risk Assessment -
independent HFEs which cutsets that, had they Human Reliability Analysis. A review of non- own dependency should have some level of been reviewed, significant cutsets prior to finalizing and analysis.
dependency: would have documenting results was performed and was HFA_02114KVCRSTIE uncovered the need documented in the TVA Calculation, (Failure to cross-tie 4kV SD to perform additional NDN00099920070041 Revision 3, "QU - BFN Board) AND operator dependency Probabilistic Risk Assessment - Quantification."
HFA_0231480SDBTIE analyses.
(Failure to provide alternate power to 480V SD Board). Possible Resolution:
(1) Re-perform operator action dependency analysis.
(2) Re-perform review of non-significant cutsets prior to finalizing and documenting results.
QU-D3 Closed 4-41 Offsite power recovery is Basis for The recovery logic/rules have been reviewed to No impact.
applied in cutsets where it Significance: ensure that recoveries are not applied to non-might not be possible. See Recoveries should recoverable failures The example cited in the U1 CDF cutset at 1.151E-08: only be applied to F&O is incorrect. If the breakers failed to open, TVA disputes the LOOP with common cause scenarios or cutsets they would still be closed and available for offsite validity of the F&O on failure of shutdown board where the recovery power recovery. the Internal Events normal feeder breakers to can be expected to PRA.
open. be successful.
Possible Resolution:
Review recovery logic/rules to ensure that recoveries are not applied to non-recoverable failures.
Page U-44 BFN Units 1, 2, and 3 NFPA 805E-32 Transition Report, Page 1385 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations SY-A8 Closed 4-42 Table 3 of the data notebook Basis for The output breakers (1818, 1822, 1812, 1816, No impact.
says that EDG boundaries Significance: 1838, 1842, 1832, and 1836) are no longer included the output breakers, Apparent explicitly modeled, but within the boundary of the but the EDG system notebook inconsistency in data EDG. TVA Calculations NDN00008220070012 The EDG logic to and the model have them as and component Revision 2, "SY.05 - BFN Probabilistic Risk start and load (close separate events. NUREG/CR- boundary definitions. Assessment - Emergency Diesel Generator output breaker) are 6928 lists breakers as System" and NDN00099920070033 Revision 4, modeled the same WITHIN the boundary of the Possible Resolution: "DA.01 - BFN Probabilistic Risk Assessment - way in both the EDG. Resolve discrepancy. Data Analysis" have been updated to reflect this internal events model change. and the fire PRA.
LE-C7 Closed 4-43 No dependency analysis is Basis for Since failure to depressurize prior to core No impact.
performed between operator Significance: damage is a failure to properly follow/execute Action IR2 (Operator fails to These two actions steps in the EOI-1 flow chart (Level 1) while depressurize after core are in the same failure to depressurize after core melt considers TVA disputes that damage) and cutset, resulting in a failure to properly follow and execute steps from there is a HFA_0001HPRVD1 (Operator combined failure the SAMG-1 flow chart (Level 2), there is no dependency between fails to initiate probability of 6.25E-8 dependency of the operator response for this actions prior to core depressurization [Level 1]). (2.5E-4*2.5E-4). action. Also, during execution of the Severe damage and those Accident Mitigation Guidelines (SAMGs), there that occur after core Possible Resolution: will be additional guidance/oversight from damage. The Fire A dependency Technical Support Center (TSC) personnel. PRA has its own analysis should be Human Reliability done between Level Analysis 1/Level 2 actions as There are no dependencies between HFEs from (NDN0009992012000 well as Level 2/Level Level 1 (EOIs) to Level 2/LERF (SAMGs and no 01, 1 TVA FIRE PRA 2 actions. dependencies among Level 2 actions. No - Task 7.12 Post-Fire dependencies are assumed among the Level 2 Human Reliability action because the emergency response Analysis) and does organization is involved in this situation. This is not use the values treated as an assumption in the analysis and generated in the documented in the assumption section of the Internal Events PRA.
TVA Calculation, NDN00099920070032 TVA-9 Revision 3 Revision 2, HR - BFN Probabilistic Risk Assessment - Human Reliability Analysis.
Page U-45 BFN Units 1, 2, and 3 NFPA 805 Transition E-33 Report, Page 1386 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations LE-C7 Closed 4-47 Split Fraction FD2 (Recover, Basis for HRAs have been quantified and are now No impact.
restore, align RHRSW or Significance: documented in the revised TVA Calculation, Revision 3 RHR (other unit) for injection No analysis (detailed NDN00099920070032 Revision 2, HR - BFN TVA-9 for containment flood) is or screening) is Probabilistic Risk Assessment - Human The Fire PRA uses based on engineering performed to Reliability Analysis. Also, discussion has been the same Level 2 judgment. HEP for DW spray determine HEPs for added to LE.01 Appendix A. model.
initiation in split fraction TD2 these split fractions.
is 'set at 1E-2.' Based on the containment event tree CET1 Possible Resolution: failure of containment, flooding does not result in Perform HRAs on a LERF sequence. Consequently, HFA_0FD2 is actions for FD2 and not a LERF contributor and need not be TD2. quantified in detail.
LE-C11 Closed 4-48 No credit is taken for Basis for LE-C11 states: No impact.
LE-C12 equipment survivability or Significance: JUSTIFY any credit given for equipment human actions following LE-C11 implies credit survivability or continued operation of equipment containment failure. be taken for No undue credit for and operator actions that could be impacted by the operation of equipment equipment failure.
survivability following equipment that is containment failure, exposed to an for Cat II/III. Section 3.1.3 of the TVA Calculation, extreme environment NDN00099920070037 Revision 0, "LE.01 - BFN resulting from core Possible Resolution: Probabilistic Risk Assessment - LERF Analysis" damage and REVIEW significant contains the following: subsequent accident progression containment breach.
sequences resulting The equipment survivability assessment, based in a large early on a review of the IDCOR Technical Report 17 release to determine (Reference 8), is documented in the TVA if engineering Calculation, NDN00099920070038 Revision 0, analyses can support "LE.02 - BFN Probabilistic Risk Assessment -
continued equipment Structural Analysis" for BFN Unit 1. As long as operation or operator the drywell and torus are intact, it is assumed actions after that the environment in the reactor and turbine containment failure buildings will not prevent the use of equipment in that could reduce those buildings. However, at the time of drywell LERF. failure, it is assumed in the Level 2 assessment that any active equipment in the torus room, adjacent corner rooms, and anywhere else in the reactor building will not be available due to Page U-47 BFN Units 1, 2, and 3 NFPA 805 E-34 Transition Report, Page 1388 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations logic under the gate associated with RHRSW and RCW pump start. Review this also for other normally running pump fault trees.
DA-C13 Closed 5-30 DA.01 does not discuss Basis for Additional discussion related to Technical No impact.
Technical Specifications of Significance: Specifications for shared systems was added to shared systems changing due Changes in T/S TVA Calculation, NDN00099920070033 to maintenance activities. requirements can Revision 4, "DA.01 - BFN Probabilistic Risk This is a comment on have an impact on Assessment - Data Analysis". Coincident completeness of the the calculation of T/M maintenance events were addressed by BFN Internal Events unavailabilities. reviewing work week assessments as described PRA documentation.
in TVA Calculation, NDN00099920070033 This has no effect on Possible Resolution: Revision 4, "DA.01 - BFN Probabilistic Risk the structure, Analyze and Assessment - Data Analysis". quantification, or document the results of the BFN impacts of T/S Fire PRA.
changes in shared systems due to test and maintenance activities.
HR-D6 Closed 6-1 HRA Method (Section 6.2.2.1) Basis for Median values have been converted to mean No impact.
applies ASEP values as Significance: values and Table 5 has been updated to add the though they are mean values. Systematic Error in mean values in the TVA Calculation, Revision 3 ASME Inquiry 08-506 on this determining the NDN00099920070032 Revision 2, HR - BFN TVA-9 says this is not acceptable, probability of HEPs Probabilistic Risk Assessment - Human The Fire PRA has its and the values should be using ASEP Reliability Analysis. own Human treated as Median Values. Reliability Analysis Possible Resolution: (NDN0009992012000 Apply ASEP method 01, 1 TVA FIRE PRA assuming the point - Task 7.12 Post-Fire estimates are Median Human Reliability values Analysis) and does not use the values Page U-53 BFN Units 1, 2, and 3 NFPA 805 E-35 Transition Report, Page 1394 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations events; resulting in no failures excessive FW events incorrect logic with this human action. No coming through for other only when applying changes are necessary.
events were FW is credited. the HFE.
HR-I2 Closed 6-26 The post-processing of HEPs Basis for The combination analysis has been revised to No impact.
HR-G7 appears not to account for all Significance: include all non-truncated combinations. Results dependencies in the HFEs. Systematic issue with documented in the notebook, HR-H3 The Fire PRA has its Numerous cutsets contain applying NDN00099920070032 Revision 2, HR - BFN own dependency TVA-9 QU-A5 Combo events as well as dependencies. Likely Probabilistic Risk Assessment - Human other events post-processed if all dependencies Reliability Analysis. analysis.
QU-C2 into the cutsets. A questions were accounted for, QU-D5 was submitted to the Analyst, the CDF would but the independence of all significantly increase. Revision 3 combinations in the cutsets was not documented in the HRA notebook. Possible Resolution:
Recommend revising combination analysis to include additional combinations that appear in the cutset results.
HR-G5 Closed 6-28 Basis for operator action time Basis for HFA_0085ALIGNCST is used in fault trees for No impact.
(30 min) for Significance: sequences where the source of inventory from HFA_0085ALIGNCST Event provides over the CST is required for 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. A MAAP case appears to be roughly 5% of CDF. documented in the TVA Calculation The fire PRA uses estimated, as is the time NDN00099920080006 Revision 2, "SC.02 - BFN the same time period available (7 hours8.101852e-5 days <br />0.00194 hours <br />1.157407e-5 weeks <br />2.6635e-6 months <br />). Possible Resolution: Probabilistic Risk Assessment - PRA MAAP to calculate the Provide more a more Thermal Hydraulics Calculation" shows that a corresponding human accurate assessment single CST will provide adequate inventory for error probability (new for the timing for 10 hours1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br />. Case 3F used an initial level in the action in Fire PRA is HFA_0085ALIGNCS Condensate Storage Tank (CST) of 15 feet HFA0085ALIGNCST T. (180,000 gallons or 24,060 ft3). The purpose of versus this case was to allow for a more realistic HFA_0085ALIGNCS analysis of the time to core damage following a T).
loss of feedwater with one stuck open safety relief valve. Plant data indicates that the level of Page U-65 BFN Units 1, 2, and 3 NFPA 805 E-36 Transition Report, Page 1406 of 1661
TVA BFN Attachment U - Internal Events PRA Quality Table U-1 Internal Events PRA Peer Review - Facts and Observations SR Status F&O Finding F&O Resolution Impact to Fire PRA ID Recommendations the CSTs for all three units is an average of approximately 19 feet and operator interviews reveal that it is plant practice to keep the levels of the CSTs above 15 feet during corresponding unit operation. The HRA for HFA_0085ALIGNCST has been revised using the 10 hour1.157407e-4 days <br />0.00278 hours <br />1.653439e-5 weeks <br />3.805e-6 months <br /> time period.
HR-G7 Closed 6-30 Dependencies between Basis for In general, dependencies between operator No impact.
QU-C2 operator actions appear to be Significance: actions have been derived within the rules non-conservatively applied. Systematic error outlined in the HRA Calculator. In one case, the Mainly, the Zero Dependence affecting around 1/2 dependency rules have been over-ridden by a A separate (ZD) between actions is of the combo events, user defined rule. In this particular case, a note dependency analysis commonly applied, simply including combo 18. was added stating the reason for the over-ride, has been done for the when one of the actions takes which is document in the TVA Calculation, Fire PRA.
longer than 60 minutes. What Possible Resolution: NDN00099920070032 Revision 2, HR - BFN TVA-9 appears to be the mistake is Correct dependency Probabilistic Risk Assessment - Human Revision 3 applying the last event tree analysis in the HRA. Reliability Analysis.
node in the Dependency Event Tree. In this tree, if the Need to depressurize would arise no less than 2 stress of either HFE is hr after ability to initiate SPC would no longer moderate or high, the upper permit use of HPCI/RCIC after CST depletion.
leg of the event tree is used. This statement is under the dependency event SO for combo 2, the HRA tree and occurs for combinations of assumes ZD, while the event HFA_0074HPSPC1, Failure to align RHR for tree would designate Low suppression pool cooling (non-ATWS/IORV) and Dependency. HFA_0001HPRVD1, Failure to initiate reactor-vessel depressurization (transient or ATWS).
The timing for the cues implies that there should be a complete dependence, however the timing for HFA_0074HPSPC1 occurs over 5.4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> and therefore there is no time dependence. The cue comes in, but the required action has such a long time in which to be accomplished, there is no dependence, hence zero dependence was manually chosen. The note in the calculator is sufficient to address the issue and the TVA Calculation, NDN00099920070032 Revision 2, Revision 3 TVA-9 HR - BFN Probabilistic Risk Assessment -
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