ML093410007
ML093410007 | |
Person / Time | |
---|---|
Site: | Oconee |
Issue date: | 11/30/2009 |
From: | Baxter D Duke Energy Carolinas |
To: | Document Control Desk, Office of Nuclear Reactor Regulation |
References | |
LAR 2008-01 | |
Download: ML093410007 (81) | |
Text
Duke Energ DAVE BAXTER Vice President egy, Oconee Nuclear Station Duke Energy ON01 VP / 7800 Rochester Highway Seneca, SC 29672 864-873-4460 864-873-4208 fax dabaxter@dukeenergy.com November 30, 2009 U. S. Nuclear Regulatory Commission Document Control Desk Washington, D. C. 20555
Subject:
Duke Energy Carolinas, LLC Oconee Nuclear Station, Units 1, 2, and 3 Docket Numbers 50-269, 50-270, and 50-287 Request for Additional Information regarding the License Amendment Request to adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)
License Amendment Request (LAR) 2008-01 In accordance with 10 CFR 50.90, Duke Energy Carolinas, LLC (Duke) proposes to amend Renewed Facility Operating Licenses (FOLs) Nos. DPR-38, DPR-47, and DPR-55. This LAR requests Nuclear Regulatory Commission (NRC) review and approval for adoption of a new fire protection licensing basis which complies with the requirements in 10 CFR 50.48(a), 10 CFR 50.48(c), and the guidance in Regulatory Guide (RG) 1.205. The LAR was submitted to the NRC on October 31, 2008 and supplemented in letters dated January 30, 2009, February 9, 2009, February 23, 2009, May 31, 2009, August 3, 2009, and September 29, 2009.
In order to complete review of the LAR, the NRC issued a preliminary request for additional information (RAI). Duke and the NRC met August 19 & 20, 2009, to review the RAIs to ensure that there was a common understanding of the requirements to address the requests. As a result of that meeting, several RAIs were withdrawn and are noted as withdrawn in Enclosure 1. The remaining RAIs were issued formally on November 18, 2009. Enclosure 1 contains the responses to the RAIs.
Duke also provided responses to RAIs 5-15 and 5-16 (RAIs received June 18, 2009) in Enclosure 1. The responses were omitted from the RAI response dated August 3, 2009, due to ongoing discussions between the NFPA-805 Task Force and the NRC concerning FAQ 07-30, Operator Manual Action Transition to Recovery Actions.
Duke and the NRC met on November 12, 2009, to ensure that Duke is incorporating proposed RG 1.205, Revision 1, properly. As discussed during the November 12, 2009 Public Meeting, Duke will take the following actions:
www. duke-energy. com
U. S. Nuclear Regulatory Commission November 30, 2009 Page 2
- Define Primary Control Station.
- Establish a Mission Time.
- Determine recovery actions required to demonstrate availability of a success path.
- Evaluate the risk of recovery actions (Resolution of FAQ 07-0030).
Duke previously committed to supplement the LAR by January 31, 2010, for transition to NFPA 805. However, due to methodology development and continued work required to implement proposed RG 1.205, Revision 1, the anticipated commitment date for the supplement is being delayed. The new date will be communicated to the NRC when a project schedule loaded with the additional activities required to meet the new guidance has been completed. Duke will provide this date by December 17, 2009.
Duke will also post the new project schedule to the shared website for reference. The schedule will be updated to reflect progress as work is completed.
A commitment table supporting this set of RAIs is provided in Attachment 1. Duke will provide a complete listing of NFPA 805 commitments when the LAR supplement is submitted.
If there are any questions regarding this submittal, please contact Reene' Gambrell at (864) 873-3364 or David J. Goforth at (704) 382-2659.
I declare under penalty of perjury that the foregoing is true and correct. Executed on November 30, 2009.
Sincerely, Dave B er, Vice President Oconee Nuclear Station
Enclosure:
- 1. Request for Additional Information regarding the License Amendment Request to adopt NFPA 805 Performance-Based Standard for Fire Protection for Light Water Reactor Generating Plants (2001 Edition)
Attachment:
- 1. Commitments
U. S. Nuclear Regulatory Commission November 30, 2009 Page 3 cc: w/o enclosures Mr. Luis Reyes, Regional Administrator U. S. Nuclear Regulatory Commission - Region II Atlanta Federal Center 61 Forsyth St., SW, Suite 23T85 Atlanta, Georgia 30303 Mr. John Stang, Project Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-8 G9A Washington, D. C. 20555 Mr. Paul Lain, NFPA-805 Program Manager Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Mail Stop 0-10 C15 Washington, D. C. 20555 Mr. Andy Sabisch Senior Resident Inspector Oconee Nuclear Station Ms. Susan Jenkins, Manager Infectious and Radioactive Waste Management Section 2600 Bull Street Columbia, SC 29201
Enclosure I REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LICENSE AMENDMENT REQUEST TO ADOPT NFPA 805 PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT WATER REACTOR GENERATING PLANTS (2001 EDITION)
- Request For Additional Information November 30, 2009 ENCLOSURE 1 REQUEST FOR ADDITIONAL INFORMATION REGARDING THE LICENSE AMENDMENT REQUEST TO ADOPT NFPA 805 PERFORMANCE-BASED STANDARD FOR FIRE PROTECTION FOR LIGHT WATER REACTOR GENERATING PLANTS (2001 EDITION)
REQUEST FOR ADDITIONAL INFORMATION (RAI) 1-3:
The second sentence in Section 1.1.1 of the Oconee Nuclear Station (ONS) Transition Report states:
"... 10 CFR 50.48(c) endorses, with exceptions, the National Fire Protection Association's NFPA 805, Performance-Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants - 2001 Edition, as a voluntary alternative for demonstrating compliance with 10 CFR 50.48 Section (b),
Appendix R, and Section (f), Decommissioning."
This sentence is incorrect. The change to 10 CFR 50.48 incorporates the NFPA standard by reference, with exceptions, resulting in the standard actually becoming a part of the regulation.
Similarly, the footnote on page 8 of the ONS Transition Report appears to be in error since the regulations do not "endorse" NFPA 805. The NFPA 805 rule took exception to the Life Safety and Business Interruption goals, objectives and performance criteria and incorporated by reference the remaining sections of NFPA 805.
Revise the Transition Report accordingly.
RAI 1-3 RESPONSE:
Section 1.1.1 and the footnote in the Transition Report will be revised to reflect that NFPA 805 was incorporated by reference and submitted with the License Amendment Request (LAR) supplement.
RAI 1-4:
Section 3.1 of the ONS Transition Report does not include any mention of the need to complete committed modifications.
Is the completion of any required modifications meant to be included in item "(7) complete implementation of the new licensing basis"? If so, this needs to be explicitly included in the discussion since completion of required modifications is required to complete transition.
Provide a summary of proposed modifications, a schedule for implementation of those modifications, and a proposed license condition reflecting the commitment and schedule for installation of the modifications.
RAI 1-4 RESPONSE:
Section 3.1 of the Transition Report and Attachment S of the LAR will be revised to reflect the list of committed modifications in the LAR supplement. Each committed modification will include a summary of the proposed modification scope, a schedule for implementation of the 1
- Request For Additional Information November 30, 2009 modification, and a proposed License Condition reflecting the commitment and schedule for installation of the modification. In addition, Attachment C will include an overview of how the modification is to be used in the new NFPA 805 Fire Protection Program at ONS.
RAI 1-5:
Section 4.2.2.3 of the ONS Transition Report includes a bullet for "Open Items." The use of the term "Open Item" implies that further analysis is required. The staff cannot approve a request for transition to a risk-informed, performance-based fire protection program if the analyses required have not been completed.
Title 10 CFR Part 50.48(c)(3)(ii) states:
"The licensee shall complete its implementation of the methodology in Chapter 2 of NFPA 805 (including all required evaluations and analyses) and, upon completion, modify the fire protection plan required by paragraph (a) of this section to reflect the licensee's decision to comply with NFPA 805, before changing its fire protection program or nuclear power plant as permitted by NFPA 805."
The licensee should provide a statement that all required engineering analyses have been completed and that compliance with the requirements has been achieved.
RAI 1-5 RESPONSE:
As described in NEI 04-02 and its associated FAQs, the use of the term open item is an integral part of the process. The term "open item" within Section 4.2.2.3 of the ONS Transition Report was meant to identify a 'variance from the deterministic requirement' that ONS could not self-approve, and therefore implement, under the current licensing basis. This concept was discussed in Section 4.2.2.1, Overview of Evaluation Process. Items included:
" Variances from the current deterministic licensing basis for which change evaluations (now being called fire risk evaluations) were performed (Section 4.5 of the Transition Report)
" Items for which NRC concurrence of prior approval were required (Attachment T of the Transition Report).
" Items for which modifications were committed (Attachment S of the Transition Report).
The necessary analyses for the 'variance from the deterministic requirements' were contained within the change evaluations (now being called fire risk evaluations), scoping of modifications, request for specific NRC approval or in the request for prior approval clarification.
Duke will provide a statement under oath and affirmation as to the state of the required evaluations and analyses that support the transition to a risk-informed, performance-based fire protection program in the LAR supplement.
RAI 1-6:
Attachment P of the ONS Transition Report is based on the approach documented in NFPA 805 Frequently Asked Question (FAQ) 06-0008, Revision 8. Agreement was reached and a closure memo was approved based on Revision 9. To obtain the benefit of the approach documented in 2
Enclosure 1: Request For Additional Information November 30, 2009 Revision 9, the Transition Report must reference the appropriate version. Explain how the LAR, including Attachment P, would change based on Revision 9 of FAQ 06-0008.
RAI 1-6 RESPONSE:
The ONS Transition Report will be revised, as necessary, to incorporate by reference the portions of FAQ 06-0008, Revision 9 (ML0090560170) which were approved within the associated closure memo (ML073380976). Specific to Attachment P, the Oconee LAR will be revised to remove the attachment since it is no longer needed with the approved portions of FAQ 06-0008, Revision 9.
RAI 2-7:
Item (3) of NFPA 805 Section 3.2.3 requires reviews of fire protection program related performance and trends. Describe how compliance to Section 3.2.3 (3) will be achieved.
RAI 2-7 RESPONSE:
NFPA 805 Section 3.2.3(3) requires that procedures be established for reviews of the fire protection program related to performance and trends. NFPA 805, Section 2.6 requires a monitoring program that in part is to establish acceptable performance levels and a method to monitor and assess the performance of the fire protection program. The NFPA 805 requirements for reviews of programs related to performance and trending is provided under the NFPA 805 Monitoring program as described in Transition Report, Section 4.6, Monitoring, and Attachment R, UFSAR Changes. Transition Report, Section 4.6.2 describes the overall Post-Transition NFPA 805 Monitoring Program process. The Monitoring program will be implemented after the LAR approval as part of the fire protection program transition to NFPA 805. The monitoring process will be conducted in four phases.
- Phase 1 will determine the scope which includes fire protection systems & features and fire protection program elements.
- Phase 2 will establish performance criteria.
" Phase 3 will determine risk significant fire protection program and,"Defense in Depth" elements using criteria established in Phase 2.
" Phase 4 will implement the program after the scope and criteria are established.
Performance and availability monitoring criteria will be applied to the risk significant fire protection classical elements and a tracking program will be used on the "Defense in Depth" elements. A flow chart of this process is provided in the Transition Report as Figure 4-7.
This process will result in development of a program that reviews the fire protection program performance and identifies trends in performance. The reviews will be based on specific performance goals established to measure the effectiveness of the fire protection program.
Monitoring will ensure that assumptions in engineering analysis remain valid. The monitoring program will be documented in an administrative process (i.e. program manual or site directive) instructing how to perform this process and set clear guidelines to consistently measure the performance of the fire protection program.
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Enclosure 1: Request For Additional Information November 30, 2009 RAI 2-8:
NFPA 805, Section 4.1 states: "Once a determination has been made that a fire protection system or feature is required to achieve the performance criteria of Section 1.5, its design and qualification shall meet the applicable requirement of Chapter.3."
The discussion of the fire area-by-fire area review on page 45 of the ONS LAR includes requirements to answer these two questions; "Suppression Required?" and "Detection required?" Provide the following information concerning these systems.
- a. When either of these questions is answered "Yes," what requirements apply with respect to the design and qualification of that system?
- b. What quality requirements apply to a system that has been designated as required in Attachment C of the LAR?
- c. On page 45, under Suppression Required? (Yes/No):, there is a bullet that states that systems required to meet Chapter 4 performance-based compliance (including systems credited for DID) are summarized in change evaluations. This passage implies that systems credited for defense-in-depth in plant change evaluations should be considered.
If a system is required by Chapter 4 its design and qualification would need to meet the applicable requirement of Chapter 3. Is that implication correct? What design, qualification and quality controls apply to fire detection and suppression systems that are credited for defense-in-depth?
RAI 2-8a RESPONSE:
A suppression or detection system that is identified as required in the Fire Area-by-Fire Area review (Transition Report Attachment C) is required to meet the provisions of NFPA 805 Section 3.8 or 3.9 for detection or suppression, respectively. Attachment C, Table B-3 Fire Area Transition, records if suppression and/or detection is required after the Engineering Evaluation summary. The provisions of NFPA 805 Section 3.8 and 3.9 dictate the design and qualification of the system (the system shall meet the requirements of NFPA 13, 15, 72, etc).
Required suppression and detection are addressed in the following proposed revision to the LAR. Note that Table 4-C will identify the Fire Area-by-Fire Area conclusion of required suppression and detection systems. This will be redundant information in a summary table for the ease of reference.
4.2.2.1.1 Fire Protection Systems and Features Required to Achieve the Nuclear Safety Performance Criteria In accordance with 10 CFR 50.48(c) "Once a determination has been made that a fire protection system or feature is required to achieve the performance criteria of Section 1.5, its design and qualification shall meet the applicable requirement of Chapter 3" Fire protection systems or features are required for NFPA 805 Chapter 4 compliance to achieve the performance criteria of Section 1.5 if the:
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- Request For Additional Information November 30, 2009
" Fire Protection Systems and Features are required to meet NFPA 805 Section 4.2.3, Deterministic Approach, or
- Fire Protection Systems and Features are required to meet NFPA 805 Section 4.2.4, Performance-Based Approach Methodology To determine if a fire protection system or feature is required to achieve the performance criteria of NFPA 805 Section 1.5 the following process was used:
Deterministic Approach If a fire protection system or feature is required to meet one of the following deterministic compliance strategies, then it is required to meet the nuclear safety performance criteria and therefore its design and qualification shall meet the appropriate sections of NFPA 805 Chapter 3:
- 1. Fire protection systems and features required for deterministic compliance in accordance with Section 4.2.3 of NFPA 805
- a. Outside Containment
" Fire area boundary. One success path of required cables and equipment shall be located in a separate area having boundaries consisting of fire barriers with a minimum fire resistance rating of 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />. Every opening in the fire barriers forming these boundaries shall be protected with passive fire protection features having a fire-resistive rating equivalent to the fire barrier (Section 4.2.3.2).
" 3-hour rated fire barrier or Electrical Raceway Fire Barrier System. Separation of required cables and equipment of redundant success paths by a fire barrier having a 3-hour fire resistance rating. Enclosure of cable and equipment and associated non-safety circuits of a redundant success path in a fire barrier or Electrical Raceway Fire Barrier System having a 3-hour fire resistance rating (Section 4.2.3.3(a)).
" Suppression and detection. Separation of required cables and equipment of redundant success paths by a horizontal distance of more than 20 ft with no intervening combustible materials or fire hazards. In addition, automatic fire detectors and an automatic fire suppression system shall be installed throughout the fire area (Section 4.2.3.3(b)).
i-hour 1 rated fire barrier or Electrical Raceway Fire Barrier System and suppression and detection. Enclosure of cable and equipment and associated non-safety circuits of one redundant success path in a fire barrier or Electrical Raceway Fire Barrier System having a 1-hour fire resistance rating. In addition, automatic fire detectors and an automatic fire suppression system shall be installed throughout the fire area (Section 4.2.3.3(c)).
- b. Inside Non-inerted Containment Non-combustible radiant energy shield. Separation of required cables and equipment of redundant success paths by a noncombustible radiant energy shield. These assemblies shall be capable of withstanding a minimum 1/2-hour fire exposure when tested in accordance with NFPA 251, Standard Methods of 5
- Request For Additional Information November 30, 2009 Tests of Fire Endurance of Building Construction and Materials (Section 4.2.3.4(b)).
Suppression and detection. Installation of automatic fire detectors and an automatic fire suppression system throughout the fire area (Section 4.2.3.4(c)).
- 2. Required by Existing Engineering Equivalency Evaluation (EEEE)
As allowed by Section 2.2.7 of NFPA 805, "...the user shall be permitted to demonstrate compliance with specific deterministic fire protection design requirements in Chapter 4 for existing configurations with an engineering equivalency evaluation." These existing engineering equivalency evaluations include evaluations previously known as Generic Letter 86-10 evaluations, exemptions, and deviations. Fire Protection systems and features that form the bases for acceptability of these existing compliance strategies are required to meet the nuclear safety performance criteria.
Performance-Based Approach In accordance with NFPA Section 4.2.4.2, the "...use of fire risk evaluation for the performance-based approach shall consist of an integrated assessment of the acceptability of risk, defense-in-depth, and safety margins." If the fire protection system or feature is required to demonstrate the acceptability of risk or defense-in-depth, then it is required by Chapter 4. The following method is used to determine if a fire protection feature or system is required for the acceptability of risk or defense-in-depth.
- 1. Acceptability of Risk A fire protection feature may be required for the 'acceptability of risk' in one of two ways:
- a. It is explicitly credited to reduce risk in the NFPA 805 transition fire risk evaluation, or
- b. It is 'of higher significance' to the overall fire risk for the plant. Examples include:
" If the calculated risk (CDF/LERF) of system(s) in an area are above the established threshold.
" If the system is determined to have a risk achievement worth above the established threshold. This is determined during the expert panel review conducted as part of the plant monitoring scoping activities.
- 2. Defense-in-Depth In accordance with NFPA 805 Section 2.4.4, Plant Change Evaluation, "...The evaluation process shall consist of an integrated assessment of the acceptability of risk, defense-in-depth, and safety margins." NFPA 805 Section 4.2.4.2 refers to the acceptance criteria in this section. Therefore fire protection systems and features required to demonstrate an adequate balance of defense-in-depth are required by NFPA 805 Chapter 4. Criteria for determining what constitutes an adequate balance are provided in Table 4-A.
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- Request For Additional Information November 30, 2009 Table 4-A - Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Prevent fires from starting
- Combustible Control Combustible and hot work controls are fundamental elements of defense in depth and as such are always in place. The issue to be
- Hot Work Control considered during the fire risk evaluations is whether this element needs to be strengthened to offset a weakness in another echelon thereby providing a reasonable balance. Considerations include:
- Creating new Transient Free Areas
- Modifying an existing Transient Free Area The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine if additional controls should be added.
Review the remaining elements of defense in depth to ensure an over-reliance is not placed on programmatic activities for weaknesses on plant design.
Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage
" Detection system Automatic suppression and detection may or may not exist in the fire area of concern. The issue to be considered during the fire risk
- Automatic fire suppression evaluation is whether installed suppression and or detection is
" Portable fire extinguishers provided required for defense in depth or whether suppression/detection for the area needs to be strengthened to offset a weakness in another echelon
- Hose stations and hydrants provided thereby providing a reasonable balance. Considerations include:
for the area
- If a fire area contains both suppression and detection and
- Fire Pre-Fire Plan fire fighting activities would be challenging, both detection and suppression may be required
- If a fire area contains both suppression and detection and fire fighting activities would not be challenging, detection and manual fire fighting required (consider enhancing the pre-plans) 0 If a fire area contains detection and a recovery action is required, the detection system may be required.
a If a fire area contains neither suppression nor detection and a recovery action is required, consider adding detection or suppression.
The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine the types of fires and reliance on suppression probability should be evaluated in the area to best determine options for this element of defense in depth.
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- Request For Additional Information November 30, 2009 Table 4-A - Considerations for Defense-in-Depth Determination Method of Providing DID Considerations Provide adequate level of fire protection for systems and structures so that a fire will not prevent essential safety functions from being performed
" Walls, floors, ceilings and structural If fires occur and they are not rapidly detected and promptly elements are rated or have been extinguished, the third echelon of defense in depth would be relied evaluated as adequate for the upon. The issue to be considered during the fire risk evaluation is hazard. whether existing separation is adequate or whether additional measures (e.g., supplemental barriers, fire rated cable, or recovery
" Penetrations in the fire area barrier actions) are required to offset a weakness in another echelon; are rated or have been evaluated as thereby, providing a reasonable balance. Considerations include:
adequate for the hazard.
- If the variance is never affected in the same fire scenario,
- Supplemental barriers (e.g., Electrical internal fire area separation may be adequate and no Raceway Fire Barrier System, cable additional reliance on recovery actions is necessary.
tray covers, combustible liquid dikes/drains, etc.)
- If the variance is affected in the same fire scenario, internal fire area separation may not be adequate and reliance on a
" Fire rated cable recovery action may be necessary.
" Reactor coolant pump oil collection
- If the CCDP associated with the variances is high regardless system (as applicable) of whether it is in the same scenario, a recovery action and I
" Guidance provided to operations or reliance on supplemental barriers should be considered.
personnel detailing the required
- There are known modeling differences between a Fire PRA success path(s) including recovery and nuclear safety capability assessment due to different actions to achieve nuclear safety success criteria, end states, etc. Although a variance may be performance criteria. associated with a function that is not considered a significant contribution to core damage frequency, the variance may be considered important enough to the nuclear safety capability assessment to retain as a recovery action.1 The fire scenarios involved in the fire risk evaluation quantitative calculation should be reviewed to determine the fires evaluated and the CCDP in the area to best determine options for this element of defense in depth.
Summary Table 4-B summarizes the approach to determine the fire protection systems and features required to meet the performance or deterministic approaches of Chapter 4 of NFPA 805.
An example would be components in the nuclear safety capability assessment associated with maintaining natural circulation at a pressurized water reactor that are not modeled explicitly in the Fire PRA since they are not part of a core damage sequence.
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- Request For Additional Information November 30, 2009 Table 4-B Determination of NFPA 805 Chapter 4 Required FP Systems and Features Performance Based Deterministic Compliance Loss Category (NFPA 805 § 4.2.3) Compliance Prevention (NFPA 805 § 4.2.4)
Description Required for Required as a Required for Required to Not NFPA 805 basis of the maintain an Required to Chapter 4 acceptability acceptability of adequate meet separation of existing risk balance of § 4.2.3 criteria EEE determination Defense-in- or (evaluations Depth § 4.2.4 previously known as Generic Letter 86-10 evaluations, exemptions, and deviations)
Required Yes Yes Yes Yes No System I 4.8.2.2.1 Results of the Determination of Required Fire Protection Systems and Features In accordance with the methodology outlined in Section 4.2.2.1.1 fire protection systems and features in each fire area in the power block were evaluated to determine if it is:
- Required to meet NFPA 805 Section 4.2.3, Deterministic Approach, or
- Required to meet NFPA 805 Section 4.2.4, Performance-Based Approach The results of this review are provided in Table 4-C - Fire Protection Systems and Features Required for NFPA 805 Chapter 4 Compliance. Note Table 4-C is empty and provided as an example of what will be included in the LAR supplement.
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- Request For Additional Information November 30, 2009 Table 4-C -Fire Protection Systems and Features Required for NFPA 805 Chapter 4 Compliance Fire Area Fire Zone Description Suppression Provided? Suppression Required? Detection Provided? Detection Required? Feature2 Provided? Feature Required?
2 Features other than Suppression or Detection are those fire protection features internal to the Fire Area such as Electrical Raceway Fire Barrier Systems. Fire barrier assemblies are not optional and always required.
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Enclosure 1: Request for Additional Information November 30, 2009 RAI 2-8 b RESPONSE:
Required fire protection systems at ONS are classified in accordance with the Duke Quality Assurance topical report and NSD-307, Quality Standard Manual. NSD-307 defines the quality assurance condition for fire protection as QA Condition 3. QA Condition 3 applies uniquely to fire protection structures, systems, components, and services as defined in the respective station Fire Protection Design Basis Document as needed to protect nuclear safety-related structures, systems, or components.
RAI 2-8c RESPONSE:
Systems credited for defense in depth in plant change evaluations are considered "required-by Chapter 4" and therefore the design, qualification, and quality control of such systems would be the same as any other required system. The response to RAI 2-8a and 2-8b pertain to all required systems, including those credited for defense in depth.
RAI 2-9a:
Some NFPA 805 Chapter 3 elements are complicated or have multiple varied applications throughout the plant (e.g., 3.11.4 Through PenetrationFire Stops). This results in some elements requiring more that one compliance strategy entry to fully capture the licensee's compliance basis.
ONS's B-1 Table currently contains one compliance statement for each NFPA 805 Chapter 3 element or sub-element. Where ONS relies on more than one compliance strategy for a particular element, the B-1 Table must fully capture all of ONS's methods of compliance. Ensure that where necessary, all of ONS's compliance strategies are captured in the B-1 Table. The description of the B-1 Table methodology in the Transition Report (i.e., Section 4.1) should be updated to reflect any changes required by multiple compliance strategies. ONS must also ensure that all aspects of each particular Chapter 3 element or sub-element are addressed by the appropriate compliance statements.
For example, Table B-1 of the LAR fundamental element NFPA 805, Section 3.11.2, identifies two compliance statements; "NRC prior Approval" and "Licensee Evaluation" for fire barriers.
Both of these are described as "fire barriers between the major buildings". This description does not provide any distinction between the two. The NRC staff is unable to identify which barriers fall into which category. Please clarify the distinction. Additionally, these are the only compliance strategies for fire barriers. This would imply that no fire barriers simply meet tested/listed configurations for walls and all fire barriers are covered by only those two compliance strategies. Clarify that this is true.
RAI 2-9a RESPONSE:
Table B-1 will be reviewed to ensure compliance strategies are correctly addressed and captured for all Chapter 3 Sections. The compliance basis where multiple compliance statements are used for one NFPA 805 element or sub-element will be revised as necessary.
These revisions will ensure, in a clear manner, the multiple compliance statements are addressed by the compliance basis. The compliance basis will clearly show which aspects of the basis correspond to the appropriate compliance statements. In addition, the description for the Table B-1 methodology in Section 4.1 of the Transition Report will be reviewed and enhanced as necessary with regards to the use of multiple compliance statements. These 11
Enclosure 1: Request for Additional Information November 30, 2009 revisions will be reflected in the revised Transition Report and Table B-1 which will be submitted with the LAR supplement.
RAI 2-9b:
Evaluations of ONS's compliance with several NFPA standards are referenced in a number of B-1 Table elements. A detailed summary of the results of each of these compliance evaluations should be provided in the, LAR and referenced in the B-1 Table. These summaries may be compiled in a separate attachment to the submittal.
At a minimum, each summary should include:
- 1. a description of all evaluated conditions determined to be acceptable based on an engineering, or other type of, evaluation, including:
- a summary of each condition;
- a summary of the evaluation of each condition; and
- a summary of the resolution of each condition.
- 2. a description of all apparent code deviations, including:
- a summary of each deviation;
- a summary of the evaluation of each deviation; and 0 a summary of the resolution of each deviation.
If ONS wishes to treat these compliance evaluations like existing engineering equivalency evaluations, then this information need not be submitted but available for future review.
Unless specifically limited by the Chapter 3 element, these evaluations should be completed, at
-'a minimum, for all power block areas. (Note that certain standards, such as NFPA 600, apply plant-wide by their nature, and cannot be so limited.)
A partial list of the NFPA standards referenced is:
" NFPA 14
- NFPA 51B
' NFPA 55
- NFPA 72D
ONS is performing NFPA code compliance evaluations in order to demonstrate compliance with a number of Table B-1 elements. These evaluations are being prepared as engineering 12
Enclosure 1: Request for Additional Information November 30, 2009 evaluations. The RAI 2-6 response (Duke letter dated August 3, 2009) addressed NFPA Code Compliance as follows:
"With respect to evaluations of ONS's compliance with NFPA standards as referenced in Table B-i, ONS will review the NFPA code compliance engineering evaluations against the requirements of FAQ 07-0033 when the evaluations are completed. These evaluations will be referenced in a revised Table B-1 by calculation number and the calculations (code compliance evaluations) will be available for NRC review. Upon ONS review of the calculations, it will then be determined ifthere are evaluations outside of the bounds of FAQ 06-0008 that would require summary in the LAR and require submittal to the NRC for approval."
Since NRC approval is not required for these types of evaluations following transition, it is proposed that these evaluations do not need to be summarized/included in the LAR. The NFPA code compliance evaluations will be available for NRC review.
If there are deviations in the NFPA code compliance evaluations where the deviation/non-compliance does not affect safety of functionality of the system, the non-compliance will be submitted in Attachment L of the Transition Report for NRC approval.
ONS recognizes that unless specifically limited by the Chapter 3 element, NFPA code compliance evaluations should be completed, at a minimum, for all power block areas. The following NFPA codes are applicable plant-wide at ONS: NFPA 20, NFPA 24, NFPA 51B, and NFPA 600.
RAI 2-10:
In 10 CFR 50.48(c), the NRC has specifically not endorsed sections of the NFPA 805 standard.
Generally this text should not be part of ONS's licensing basis. Two examples in the B-1 table are the exceptions to 3.3.5.3 and the exception to 3.6.4. What is the basis for providing none endorsed text of the NFPA 805 standard in the B-1 table?
RAI 2-10 RESPONSE:
The text of the non-endorsed sections of NFPA 805 will be removed. This includes the exception to Section 3.3.5.3 and the exception to Section 3.6.4 not endorsed under 10 CFR 50.48(c), (c.2.v and vi). This will be reflected in revised Table B-1 which will be submitted with the LAR supplement.
RAI 2-11:
B-1 Table: Compliance Strateqy: Complies Via Previous Approval For Chapter 3 elements with a "Complies via Previous Approval" compliance strategy, the following details are required to be provided: (1) appropriate excerpts from submittals regarding the issue for which previous approval is being claimed, followed by (2) appropriate excerpts from the NRC documents that provide the formal approval of the fire protection system/feature.
References for both submittal and approval documents are also required. Additionally, there must be a positive statement with regard to condition(s) in place at the time of approval being still in effect for the NFPA 805 compliance strategy. For example, from Table B-1 section 3.11.3, Fire Barrier Penetrations, fire dampers (page 76 of 80) were to be upgraded as a 13
- Request for Additional Information November 30, 2009 condition for approval of the condition (Safety Evaluation (SE) Section 4.9.3). However, no positive statement delineating that all conditions and modifications have been completed or otherwise dispositioned as conditions of acceptance was provided in the Transition Report.
During the review, the NRC staff identified a number of entries where the above requirements were not met. The following matrix details the inadequacies that were identified during the review. ONS should correct these and ensure that there are no others.
Chapter 3 Element Identified Issue 3.3.5.3 Provide submittal and approval excerpts for each deviating detail.
Order the excerpts so that the submittal excerpt is followed by the approving document excerpt.
Provide a positive statement that all SE conditions for approval are met and still in effect.
3.5.1 Provide an excerpt from the submittal document.
Order the excerpts so that the submittal excerpt is followed by the approving document excerpt.
Provide a positive statement that all SE conditions for approval are met and still in effect.
3.5.3 Provide an excerpt from the submittal.
Order the excerpts so that the submittal excerpt is followed by the approving document excerpt.
Provide a positive statement that all SE conditions for approval are met and still in effect.
3.5.4 Provide an excerpt from the submittal.
Order the excerpts so that the submittal excerpt is followed by the approving document excerpt.
Provide a positive statement that all SE conditions for approval are met and still in effect.
3.5.5 Provide an excerpt from the submittal.
Order the excerpts so that the submittal excerpt is followed by the approving document excerpt.
Provide a positive statement that all SE conditions for approval are met and still in effect.
14
Enclosure 1: Request for Additional Information November 30, 2009 Chapter 3 Element Identified Issue 3.5.13 Provide a submittal document reference.
Provide an excerpt from the submittal document.
Order the excerpts so that the submittal exceprt is followed by the approving document excerpt.
Provide a positive statement that all SE conditions for approval are met and still in effect.
3.6.1 Provide a submittal document reference.
Provide an excerpt from the submittal document.
Order the excerpts so that the submittal exceprt is followed by the excerpt.
approving document Provide a positive statement that all SE conditions for approval are met and still in effect.
RAI 2-11 RESPONSE:
Table B-1 will be reviewed to ensure the compliance basis is clearly documented in all cases when using the "Comply via Previous Approval" compliance strategy. A traceable reference to the submittal document will be provided in all cases, and where necessary, the submittal excerpt will be included. The excerpt from the submittal document will be provided in cases where the NRC approval document is not explicitly clear. Many excerpts from the referenced NRC approval documents contain sufficient detail regarding the approval conditions of the fire protection system/feature or the NRC document contains the verbatim text from the submittal document. If a submittal excerpt is required, it will be placed before the approving document excerpt. A positive statement that all conditions for approval are met and still in effect will be provided for all cases where "Comply via Previous Approval" is cited for the compliance basis.
These changes will be reflected in the revised Table B-1 which will be submitted with the LAR supplement.
RAI 2-12:
During the review of the Table B-1, the NRC staff identified the following issues that are linked to specific Table elements. ONS should review the submittal and ensure that any similar conditions are resolved appropriately.
B-1 Table: Element 3.3.1.1 NFPA 805 Section 3.3.1.1 states: "The fire prevention activities shall include but not be limited to the following program elements:" Provide a compliance statement that addresses the "... but not be limited to ..." part of the requirement.
15
Enclosure 1: Request for Additional Information November 30, 2009 B-1 Table: Element 3.3.1.2 NFPA 805 Section 3.3.1.2 states, in part: "These procedures shall include but not be limited to the following program elements: .... " Provide a compliance statement that addresses the "... but not be limited to ..." part of the requirement.
B-1 Table: Element 3.3.1.2.(5)
Identify in the Table which other NFPA standards were determined to be applicable, and provide references to code compliance calculations for these other standards.
B-1 Table: Element 3.3.1.2.(6)
Identify in the Table entry which NFPA standards were determined to be applicable, and provide references to code compliance calculations for these standards.
B-1 Table: Element 3.3.1.3.1 Address ONS compliance with NFPA 241, Standard for Safeguarding Construction, Alteration, and Demolition Operations, as required by this section of NFPA 805.
B-1 Table: Element 3.4.1
- Provide point-by-point compliance statements for the subsections of this element.
B-1 Table: Element 3.4.2.1 The Compliance Basis for this element refers to another document not submitted. In the LAR, list what categories of information the ONS pre-fire plans actually contain.
B-1 Table: Element 3.4.3.(a)
Provide point-by-point compliance statements and references for the numbered subsection of this element. Compliance with all of these points is required.
B-1 Table: Element 3.4.3.(c)
Provide point-by-point compliance statements and references for the numbered subsection of this element. Compliance with all of these points is required.
B-1 Table: Element 3.5.13 Address the seismic portion of this requirement and provide appropriate compliance information.
B-1 Table: Element 3.6.2 Ensure that the Compliance Strategy for this element is correct (i.e., how does the code compliance evaluation relate to the "Comply" Compliance Strategy?).
16
- Request for Additional Information November 30, 2009 B-1 Table: Element 3.11.1 Does ONS utilize the exception? If so, provide a detailed summary of the completed performance based analysis.
B-1 Table: Element 3.11.3 ONS needs to address the compliance to NFPA 101 for this element. The NRC did not take exception to its inclusion in this element. Therefore, the provisions of NFPA 101 related to this element (i.e., the characteristics of fire barrier penetration protective devices [fire doors and dampers]) do apply. ONS must address these requirements and correct this entry.
RAI 2-12 RESPONSE:
To facilitate review, the RAI text is repeated below in italics with the ONS responses following (indented). Where the response indicates that Table B-1 will be revised, the revision will be reflected in a revised Table B-1 which will be submitted with the LAR supplement. In addition, Table B-1 will be reviewed in its entirety to ensure that any similar conditions to those identified in this RAI are resolved appropriately.
B-1 Table: Element 3.3.1.1 NFPA 805 Section 3.3.1.1 states: "The fire prevention activities shall include but not be limited to the following programelements:" Provide a compliance statement that addresses the "... but not be limited to ... " part of the requirement.
The Table B-1 response to NFPA 805, Section 3.3.1.1 will be revised to state "Upon review of the elements listed below, ONS believes that the NFPA 805 code requirements are satisfied and no additional elements were evaluated."
B-1 Table: Element 3.3.1.2 NFPA 805 Section 3.3.1.2 states, in part: "These procedures shall include but not be limited to the following program elements: .... " Provide a compliance statement that addresses the "... but not be limited to ..."part of the requirement.
NFPA 805, Section 3.3.1.2 is an introduction to six (6) sub-parts which individually address minimum programmatic elements to be included in procedures for control of general housekeeping practices and the control to transient combustibles. The Table B-1 response to NFPA 805, Section 3.3.1.2 will be revised to state "Upon review of the elements listed below, ONS believes that the NFPA 805 code requirements are satisfied and no additional elements were evaluated."
B-1 Table: Element 3.3.1.2.(5)
Identify in the Table, which other NFPA standardswere determined to be applicable, and provide references to code compliance calculationsfor these other standards.
The Table B-1 response to NFPA 805, Section 3.3.1.2(5) will be revised to indicate that no other NFPA standards were determined to be applicable based on guidance in FAQ 06-0020.
B-1 Table: Element 3.3.1.2.(6)
Identify in the Table entry which NFPA standardswere determined to be applicable, and provide references to code compliance calculations for these standards.
17
- Request for Additional Information November 30, 2009 The Table B-1 response to NFPA 805 Section, 3.3.1.2(6) will be revised to indicate that no other NFPA standards were determined to be applicable based on guidance in FAQ 06-0020.
B-1 Table: Element 3.3.1.3.1 Address ONS compliance with NFPA 241, Standardfor SafeguardingConstruction,Alteration, and Demolition Operations,as requiredby this section of NFPA 805.
Compliance with NFPA 241 is addressed through compliance with NFPA 51B. NFPA 241, 2000 edition, as referenced by NFPA 805-2001 ed., Section 5.1.1, with respect to hot work, states "Responsibility for hot work operations and fire prevention precautions, including permits and fire watches, shall be in accordance with NFPA 51B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work." The Table B-1 response to NFPA 805, Section 3.3.1.3.1 will be revised to state this position.
B-1 Table: Element 3.4.1
" Provide point-by-point compliance statements for the subsections of this element.
" Provide a positive statement concerning which NFPA standard ONS follows.
The Table B-1 response to NFPA 805, Section 3.4.1 will be reviewed to ensure point-by-point compliance statements are provided for the subsections of this element. The response to the subsections of Section 3.4.1 will be revised as necessary. In addition, a positive statement that ONS complies with NFPA 600 via engineering evaluation (NFPA 1500/1582 does not apply) will be provided.
B-1 Table: Element 3.4.2.1 The Compliance Basis for this element refers to anotherdocument not submitted. In the LAR, list what categoriesof information the ONS pre-fire plans actually contain.
The Table B-1 response to NFPA 805, Section 3.4.2.1 will be enhanced to clearly outline/list the contents of the ONS Fire Plan. Note that ONS uses the terminology 'Fire Plan' as their pre-fire plans.
B-1 Table: Element 3.4.3.(a)
Provide point-by-pointcompliance statements and references for the numbered subsection of this element. Compliance with all of these points is required.
The Table B-1 response to NFPA 805, Section 3.4.3(a) will be reviewed to ensure point-by-point compliance statements are provided for the subsections of this element. The response to the subsections of Section 3.4.3(a) will be revised as necessary.
B-1 Table: Element 3.4.3.(c)
Provide point-by-point compliance statements and references for the numbered subsection of this element. Compliance with all of these points is required.
The Table B-1 response to NFPA 805, Section 3.4.3(c) will be reviewed to ensure point-by-point compliance statements are provided for the subsections of this element. The response to the subsections of Section 3.4.3(c) will be revised as necessary.
B-1 Table: Element 3.5.13 Address the seismic portion of this requirement and provide appropriatecompliance information.
The Table B-1 response to NFPA 805, Section 3.5.13 will be revised to ensure the seismic portion of this requirement is addressed. The supporting information in the Table B-1 calculation discusses and addresses this requirement.
18
Enclosure 1: Request for Additional Information November 30, 2009 B-1 Table: Element 3.6.2 Ensure that the Compliance Strategy for this element is correct (i.e., how does the code compliance evaluation relate to the "Comply"Compliance Strategy?).
The Table B-1 Compliance Statement to NFPA 805, Section 3.6.2 will be revised to "Comply by Engineering Equivalency Evaluation".
B-I Table: Element 3.11.1 Does ONS utilize the exception? If so, provide a detailed summary of the completed performance based analysis.
ONS utilizes the exception to NFPA 805, Section 3.11.1 where a performance-based analysis is used to determine building separation. These analyses are documented in the referenced Engineering Evaluations documented within Table B-I. The Table B-1 response to NFPA 805, Section 3.11.1 will be enhanced to clarify building separation.
B-I Table: Element 3.11.3 ONS needs to address the compliance to NFPA 101 for this element. The NRC did not take exception to its inclusion in this element. Therefore, the provisions of NFPA 101 related to this element (i.e., the characteristicsof fire barrierpenetrationprotective devices [fire doors and dampers])do apply. ONS must address these requirements and correct this entry.
ONS offers the following clarification with regards to NFPA 101, 2000 edition compliance as referenced by NFPA 805. NFPA 101, Section 8.2.3.2.1(a) with regards to rated fire door assemblies refers to NFPA 80. NFPA 101 Section 9.2.1 with regards to rated fire dampers refers to NFPA 90A. ONS is performing detailed code compliance reviews for both NFPA 80 and NFPA 90A. The Table B-1 response to NFPA 805, Section 3.11.3 will be revised to clarify the position that NFPA 101 compliance is achieved through compliance with NFPA 80 and NFPA 90A.
RAI 2-13:
B-1 Table: "Document Detail" The purpose of the "Document Detail" field in the Table B-1 is to help ensure traceability for the licensee's licensing basis. During the review, the NRC staff identified that Document Detail for traceability was not used. ONS should review the table and ensure that entries are completed where applicable.
RAI 2-13 RESPONSE:
The "Document Detail" field in the Table B-1 will be completed, as necessary, for clarification/traceability. Complete and clear reference documentation and associated document details will be provided. This will be reflected in the revised Table B-1 which will be submitted in the LAR supplement.
RAI 2-14:
Attachment L of the Transition Report discusses specific deviations from NFPA 805 Chapter 3 requirements (section 3.3.1.2(1), 3.3.1.3.4, and 3.3.7) for which ONS is seeking approval.
Provide the following relative to these requested deviations:
- The regulatory basis (i.e., 10 CFR 50.48(c)(2)(vii) or 10 CFR 50.48(c)(4)) and an appropriate regulatory justification for each deviation request.
19
Enclosure 1: Request for Additional Information November 30, 2009
" The level of technical justification or basis for request and detail provided for these deviation requests is currently insufficient as a basis for an SE. For each deviation request, ONS should provide a level of detail and technical justification equivalent to that submitted for stand-alone licensing actions.
" The technical justification must show how the performance-based approach:
- 1. Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;
- 2. Maintains safety margins; and
- 3. Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).
RAI 2-14 RESPONSE:
Attachment L of the Transition Report will be revised to add specific detail regarding meeting the acceptance criteria of 10 CFR 50.48(c)(2)(vii). This will be submitted with the LAR supplement.
The technical justifications will be reviewed and revised as necessary to ensure they show how the performance-based approach:
" Satisfies the performance goals, performance objectives, and performance criteria specified in NFPA 805 related to nuclear safety and radiological release;
- Maintains safety margins; and
- Maintains fire protection defense-in-depth (fire prevention, fire detection, fire suppression, mitigation, and post-fire safe shutdown capability).
RAI 2-15:
The current licensing basis (CLB) includes fire protection commitments in the updated final safety analysis report (UFSAR) 9.5-1, elements of lessons learned incorporated into the Fire Protection Program, and UFSAR description of barriers and controls in place to assist the program in protecting the established goals. In the new NFPA 805 LAR, these commitments and lessons learned are not-being retained (e.g., not specifically identified in the Tables B-1, B-2, or B-3; not part of an exemption or evaluation carried forward in the LAR; not part of the UFSAR update provided in Attachment R). Examples of items not included are:
- Guidance in NRC Document "Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls and Quality Assurance" has been incorporated (ONS UFSAR Section 9.5.1.3, pages 9-70). This first example identifies the 10-point Fire Protection quality assurance (QA) program. This commitment controls the fire protection program procurement, testing, non-conforming actions, and many more aspects of the QA program not identified in NFPA 805.
- Reviews are conducted of work requests by the ONS planning section to determine the effects of these activities on station fire barriers or stops. This identification then alerts personnel to special precautions that must be taken (ONS FSAR Section 9.5.1.3, pages 9-71).
20
Enclosure 1: Request for Additional Information November 30, 2009
- Work involving ignition sources such as welding and burning is performed under closely controlled conditions. The Site Fire Protection Engineer audits the welding and burning program to assure its proper implementation.
- Transformers that are oil-filled and within 50 feet of a building containing safety related systems are protected with an automatic water spray system (ONS FSAR Section 9.5.1.3, pages 9-72).
Identify the CLB fire protection features and elements that are not addressed in the ONS NFPA 805 LAR. Clarify ONS intentions regarding these identified fire protection features and elements.
RAI 2-15 RESPONSE:
If the fire protection commitment is necessary to demonstrate compliance with 10 CFR 50.48(c) then the commitment is carried forward. If the commitment is not necessary to demonstrate compliance with 10 CFR 50.48(c) then the commitment is superseded as specified in Attachment M of the LAR. Attachment M of the LAR states, in part:
"It is Duke's understanding that implicit in the revocation of this license condition, all prior FPP SERs and commitments (with the exception of the SER dated April 28, 1983) have been superseded in their entirety by the revised license condition."
Disposition of each individual commitment is not required to demonstrate compliance with 10 CFR 50.48(a), (c) and GDC 3. The requirements of 10 CFR 50.48(c) meet or exceed the requirements of 10 CFR 50.48(a) and GDC 3 (FAQ 07-0032 Closure Memo ML081400292).
The ONS methods of compliance with 10 CFR 50.48(a), (c), and GDC 3 are provided in Section 5 of the Transition Report.
RAI 3-12:
The discussion of Non-Power Operational Modes included in Sections 4.3.1, 4.3.2, and Table F-i, pages 1of 6, of the ONS Transition Report states that it is based on NFPA 805, FAQ 07-0040, without citing a revision number. Section 4.3.1 also cites FAQ 07-0040, Revision 5.
FAQ 07-0040, Revision 4, has been referenced in Table F-1 (pages 2 and 3 of 6). Identify the correct revision to which the evaluation has been completed. Table H-i, page H-6, identifies FAQ 07-0040, Revision 2, as the referenced version. The information in Section 4.3 (Table 4-1) does not appear to be entirely consistent with the closed version of FAQ 07-0040. Ensure the closed version of FAQ is used and provide justification for any deviations from the closed version of FAQ 07-0040.
RAI 3-12 RESPONSE:
The correct revision of FAQ 07-0040 based on its Closure Memo dated August 11, 2008 (ADAMS accession number ML082200528) is Revision 4. The appropriate revision of the FAQ closure memo will be referenced in sections 4.3.1 and 4.3.2 (or their successors), as well as Tables 4-1, F-1 and H-I. Any deviations from the cited reference will be detailed and justified, although none are currently expected.
21
Enclosure 1: Request for Additional Information November 30, 2009 RAI 3-13:
Attachment C and Table 4-4 of the ONS Transition Report lists both Deterministic (4.2.3) and Performance-Based (4.2.4) sections as the regulatory basis post-transition for numerous fire areas. NFPA 805, Section 4.2.2, states that either a deterministic or performance-based approach shall be selected. Listing both sections does not meet this requirement.
RAI 3-13 RESPONSE:
Both regulatory bases were listed because some nuclear safety performance criteria success paths met the deterministic approach while others met the performance-based approach.
Listing both was consistent with the guidance in NEI 04-02 and the pilot plant review conducted with the NRC. However, ONS understands that the methodology in NEI 04-02 will be revised to reflect a selection of one of the following approaches for each fire area:
0 4.2.3 Deterministic Approach 0 4.2.4.1 Performance-Based Approach - Fire Modeling 0 4.2.4.2 Performance-Based Approach - Fire Risk Evaluation As allowed by NFPA 805, Section 4.2.2, the performance-based approach shall be permitted to utilize deterministic methods for simplifying assumptions within the fire area. Therefore, if deterministic methods (engineering equivalency evaluations, exemptions/licensing actions, specific sections of 4.2.3) are employed as simplifying assumptions, one of the following regulatory basis statements will be used:
" 4.2.4.1 Performance-Based Approach - Fire Modeling with simplifying deterministic assumptions
" 4.2.4.2 Performance-Based Approach - Fire Risk Evaluation with simplifying deterministic assumptions Based on the expected revision to NEI 04-02, Attachment C and Table 4-4 will be revised in the LAR supplement.
RAI 3-14: Withdrawn RAI 3-15: Withdrawn RAI 3-16:
In Attachment F of the ONS Transition Report, page F-2, the ONS process/results description for Step 2 includes a discussion about a second expert panel that was conducted on February 16, 2006. Relative to this expert panel, provide the following:
- a general description of the composition of the expert panel, including number of licensee staff participating, disciplines and experience represented by the participants, and use of contractors/independent experts.
" an expanded description of the expert panel process that includes a description of the process used to reach consensus on the multiple spurious operations (MSOs) kept for further assessment and a description of the criteria used in the selection process.
22
- Request for Additional Information November 30, 2009
- a list of MSOs that were reviewed, any revisions subsequently made to the list, and the source of the MSOs that were reviewed (plant unique issue identified by the expert panel, generic industry MSO lists, operating experience, industry guidance documents, etc.).
RAI 3-16 RESPONSE:
During the RAI meeting on August 19-20, 2009, the originators of RAI 3-16 agreed to refer to the Fire PRA calculation, "ONS FPRA Component Selection", already submitted on the docket, for the response to this RAI. The Multiple Spurious Operations considered are discussed in Attachment D, "Multiple Spurious Considerations", and Attachment E, "Oconee MSO Expert Panel Meeting Minutes." Discussion of the general process is located in section 7.4, "Unique Fire Induced Core Damage Sequences-Expert-Panel". Information relative to the composition of the expert panel is included in Attachment E.
RAI 3-17:
The definition of "Primary Control Station" provided in Attachment G, page G-2 appears to potentially exclude what the NRC staff would consider to be valid recovery actions. The NRC staff considers the definition provided in Draft Guide (DG)-1218 to be more appropriate for NFPA 805 transition. If the definition in the DG were utilized for ONS, what additional recovery actions would be required, and what would the additional risk of their performance be?
RAI 3-17 RESPONSE:
Attachment G will be revised to adopt the definition of 'Primary Control Station" from RG 1.205, Draft Rev. 1 Regulatory Position 2.4 (October 2009, ML092881133). The existing operator manual actions will be screened against this definition to identify any additional recovery actions. The additional risk of their performance will be calculated in a fire risk evaluation per RG 1.205, Draft Rev. 1 and included in the LAR supplement.
RAI 3-18:
In Attachment G, page G-3, there is a discussion of the use of DID manual actions. Explain how the use of DID actions that are not modeled in the Fire probabilistic risk assessment (FPRA) meets the requirements of NFPA 805 Section 2.4.3.3, that requires the fire risk analysis to: "...be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant."
RAI RESPONSE 3-18:
NFPA 805 Section 2.4.3.3 states: "The PSA approach, methods, and data shall be acceptable to the AHJ. They shall be appropriate for the nature and scope of the change being evaluated, be based on the as-built and as-operated and maintained plant, and reflect the operating experience at the plant." Section 2.5.2.5 of NUREG/CR-6850 specifically advises that it is desirable to take little or no credit in the fire PRA for recovery-related manual actions identified during the safe shutdown analysis until necessary. It is understood that particular attention must be given to preemptive actions. However, in cases where the risk benefit is minimal or the risk is already low, recovery-related manual actions are not expected to be credited in the fire PRA in order to meet the standard requirements outlined in ASME/ANS RA-Sa-2009. In order to ensure that recovery-related manual actions do not have an adverse impact (if performed), a review of this potential adverse impact has been performed.
23
Enclosure 1: Request for Additional Information November 30, 2009 As discussed during the November 12, 2009 Public Meeting, Duke will revise the treatment of additional risk presented by the use of recovery actions. This treatment will generally be in accordance with the following process:
" Determine variances from the deterministic requirements of NFPA 805, Section 4.2.3
" Determine the initial population of recovery actions (activities that occur 6utside of the main control room or 'primary control stations')
o Of that population, determine recovery actions that demonstrate the availability of a success path for the nuclear safety performance criteria o Evaluate the additional risk of those recovery actions using the fire risk evaluation process
" For all other variances from the deterministic requirements (non-recovery action) perform a Fire Risk Evaluation to demonstrate compliance with NFPA 805 Chapter 4 o Based on results of the Fire Risk Evaluations, recovery actions could be required to meet acceptance criteria (change in risk or maintaining Defense In Depth/Shutdown Margin)
" All recovery actions shall be demonstrated feasible The results of the implementation of this process will be reflected in the revised Transition Report to be submitted with the LAR supplement.
RAI 3-19:
In Attachment G, page G-4, of the Transition Report states that, "OMAs [operator manual actions] that are allowed and/or have been previously reviewed and approved by the NRC (as
,documented in an approved exemption/deviation/SE) can be transitioned without using the change evaluation process."
However, use of the performance-based approach under NFPA 805, Section 4.2.4 also requires the assessment of risk. NFPA 805, Section 4.2.3, states that use of recovery actions automatically implies the use of the performance-based approach. Section 4.2.4 states, 'When the use of recovery actions has resulted in the use of this approach, the additional risk presented by their use shall be evaluated." When the risk evaluation option of the performance-based approach is selected, the calculations required by Section 4.2.4.2 are required by NFPA 805. Attachment G lists 17 OMAs being transitioned as recovery actions. Provide the evaluation of the additional risk presented by the use of these 17 recovery actions.
RAI 3-19 RESPONSE:
Duke will revise the treatment of additional risk of recovery actions. This treatment will generally be in accordance with the process documented in RAI 3-18. The results of the implementation of this process will be reflected in the revised Transition Report submitted with the LAR supplement.
24
- Request for Additional Information November 30, 2009 RAI 3-20:
In Attachment G of the ONS Transition Report, page G-6, the example cited in the third bullet states that an area without automatic detection and a high scenario conditional core damage probability may warrant a DID action. How would the operator know that he had to take the DID action if there is no detection in the area? Explain how the action could be taken if the operator does not know there is a fire in the area.
RAI 3-20 RESPONSE:
Duke will revise the treatment of additional risk of recovery actions as discussed in the response to RAI 3-18. RAI 2-8 provides current guidance on determining fire protection systems /feature and recovery actions required to meet the Defense in Depth acceptance criteria. The results of the implementation of this process will be reflected in the revised Transition Report which will be submitted with the LAR supplement.
RAI 3-21:
In Attachment G, page G-8, the ONS Transition Report states: "The alternative shutdown recovery actions are not explicitly modeled in the FPRA." NFPA 805, Section 4.2.4, requires that the additional risk of recovery actions be evaluated. Explain how this requirement is being met ifthe recovery action is not modeled in the FPRA.
RAI 3-21 RESPONSE:
Duke will revise the treatment of additional risk of recovery actions as discussed in the response to RAI 3-18. The results of the implementation of this process will be reflected in the revised Transition Report which will be submitted with the LAR supplement.
Note the recovery action does not need to be modeled in the Fire PRA to assess the additional risk of its use. The additional risk is 'evaluated' by comparing the plant condition that requires the use of the recovery action to a compliant condition.
RAI 3-22:
In Attachment G, page G-1 1, the ONS Transition Report states: "OMAs that can contribute significantly to the overall integrated decision-making process associated with the NFPA 805 transition should be identified." Was this done for ONS? If yes, what were they? What is the additional risk associated with their use?
RAI 3-22 RESPONSE:
The process, as outlined in Attachment G, will be revised such that recovery actions that demonstrate the availability of a success path will be evaluated for the additional risk presented by their use. The inclusion of pre-transition Operator Manual Actions into the population of post transition Recovery Actions will be based on the process identified in RAI 3-18.
RAI 3-23:
In Attachment G, page G-12, the ONS Transition Report states: "Due to the low risk benefit of performance of defense-in-depth actions, the additional effort per NUREG-1852 does not add 25
- Request for Additional Information November 30, 2009 measurable benefit." NFPA 805, Section 2.4.2.1, states that the availability and reliability of equipment selected shall be evaluated. The DID actions can directly impact that reliability.
Explain how this requirement is being met if the DID actions have not been evaluated for reliability.
RAI 3-23 RESPONSE:
As discussed during the November 12, 2009 Public Meeting, Duke will revise the treatment of recovery actions. See response to RAI 3-18. The results of the implementation of this process will be reflected in the revised Transition Report which will be submitted with the LAR supplement.
As discussed in Attachment G of the LAR, the reliability of recovery actions, if modeled specifically in the Fire PRA, is addressed using Fire PRA methods (i.e., Human Reliability Analysis). Although many recovery actions were not modeled in the Fire PRA, the cables that, if damaged by fire would prompt the actions, will be included as variance from the deterministic requirements. Therefore, the risk of the variances includes/bounds the risk of the actions.
Based on the acceptably low risk of the variances without the actions, further reductions possible by calculating human error probabilities and modeling the actions would not be expected to change the conclusions of the LAR.
RAI 3-24:
In Attachment G, page G-1 3 of the ONS Transition Report, Table G-1 states that for emergency lighting, tools-equipment, actions in the fire area, and time, the feasibility criterion will be performed for time critical recovery and DID actions (less than 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />).
NFPA 805, paragraph 4.2.4.1.6, requires on recovery actions that all recovery actions must be shown to be feasible including consideration of the time available to perform the action before the plant experiences a non-recoverable condition. If tools or other equipment is required to perform the recovery action, they must be on hand and available. Crediting actions in the fire area will require a performance-based analysis to demonstrate that the actions can be reliably taken when needed without causing a life-threatening condition for the operator.
Provide sufficient justification that recovery and DID actions that are required after a 2-hour time-period meet the proposed feasibility criteria in Table G-1 of the LAR.
RAI 3-24 RESPONSE:
Section 4.2.4.1.6 of NFPA 805 does not apply to ONS. All fire areas at ONS with recovery actions are compliant with NFPA 805 Section 4.2.4.2. NEI 04-02 Revision 1, as endorsed by Regulatory Guide 1.205, Revision 0 provides criteria for demonstrating feasibility.
Attachment G will be revised to include the results of the evaluation of feasibility of all NFPA 805 recovery actions. Calculation OSC-9535, Evaluation of Recovery Actions in Support of Nuclear Safety Capability Assessment, will document the evaluation of the feasibility of these recovery actions.
26
Enclosure 1: Request for Additional Information November 30, 2009 RAI 3-25:
In Attachment X of the Transition Report, page X-5, the discussion under Determination of Capability Categories makes the statement: "Internal flood (IF) is not required for the application." Actuation of fire suppression water systems, either as a result of a fire or spuriously, can have a similar impact to pipe and other equipment ruptures. Did the FPRA include consideration of internal flooding caused by fire suppression systems? If not, why not?
RAI 3-25 RESPONSE:
The potential for internal flooding is a qualitative aspect of the seismic-fire interaction assessment. The associated walkdown scope conducted for the IPEEE included review for impact of spurious operation of fire suppression systems on safety related systems. However, no flooding scenarios were identified for quantification or otherwise incorporated into the Fire PRA.
Provide a justification for your assumption that the use of armored cables, without further consideration of their current installed configuration, is adequate to prevent inter-cable faults due to fire or, alternatively, provide information that reasonably demonstrates that the as-installed configuration of the armor cable grounding scheme is consistent with the original plant design.
The LAR credits armored cables for precluding the occurrence of inter-cable shorts. As a result, only the effects of conductor-to-conductor shorts (intra-cable) within multi-conductor cables were considered. Recent (CAROLFIRE) test results demonstrate that this assumption may not be valid if the armored cables are not appropriately grounded. From the CAROLFIRE Report, Volume 1, Section 7.2.5, Grounded versus UngroundedCPTs, "Grounded versus ungrounded circuits may be a significant factor influencing the likelihood of spurious actuation for armored cables," and Section 9.2.3, Grounded Versus Un-grounded Power Supply. It appears likely that the presence of the armor itself, which is grounded in typical applications, makes it more likely that a short to ground and fuse blow failure will occur for the grounded power supply cases. In the absence of the armor, the ground plane is available only through either a grounded conductor or the raceway itself. For an un-grounded circuit, a single short to ground will not trip the circuit protection (fuse) and therefore the likelihood of spurious actuation is somewhat higher.
RAI 3-27 RESPONSE:
Armored cable is the prevalent cable utilized at ONS. The interlocked armor on the cables at ONS are terminated and grounded as per drawings 0 EE-014-04 and 0 EE-015 series. These drawing series were in effect during the plants original construction. In addition, section 6.4.1 in Engineering Design Criteria DC-4. 11, Generating Station Grounding, states that "The armor of interlocked armor cable shall be electrically continuous and grounded to equipment enclosure at each end of the cable." A similar design document exists for the Standby Shutdown Facility (OSS-0218.00-00-0010) and Radwaste (OSS-0218.00-00-009).
27
- Request for Additional Information November 30, 2009 A series of cable specifications, OSS-0316.00-00.(series), have been developed and maintained to ensure the original cable specifications are maintained. These cable specifications ensure the cables are procured with armor shielding.
A sample of plant design changes have been reviewed to ensure the original design criteria is being referenced in the change modifications with regards to grounding the cable armor. These design changes encompass all three ONS Units and pertain to safety related components that are required for fire safe shutdown. The change modifications are:
OD200342 - Unit 2 600 Volt Adequacy Motor Operated Valves Repowering The 208V Motor Operated Valves with the largest horsepower ratings will be re-powered from 600V Motor Control Centers to place the safety related 208V System back to its original design specification condition (motors above 2 horsepower are powered from the 600V Systems).
OD300343 - Unit 3 600 Volt Adequacy Motor Operated Valves Repowering The 208V Motor Operated Valves with the largest horsepower ratings will be re-powered from 600V Motor Control Centers to place the safety related 208V System back to its original design specification condition (motors above 2 horsepower are powered from the 600V Systems).
ON-13106 - Reactor Building Emergency Sump Motor Operated Valve Operator Installation 1 HP-939 & 1 HP-940 ON-1 3090 - Power Supply Reassignment and Replacement of Valve Actuators for: 1 BS-1, 1BS-2, 1HP-24, 1HP-25, 1LP-21 and 1LP-22.
ON-23106 - Reactor Building Emergency Sump Motor Operated Valve Operator Installation 2HP-939 & 2HP-940 ON-33106 - Reactor Building Emergency Sump Motor Operated Valve Operator Installation 3HP-939 & 3HP-940 ON-33092 - 600/208 AC Load Capacity Based on the above review of drawings, cable specifications, and modifications, the confidence level that the as-installed configuration of the armor cable grounding scheme is consistent with the original plant design is high and gives reasonable assurance that this configuration is adequate to prevent inter-cable faults due to fire.
RAI 3-28:
Provide a technical basis to support the validity of your assumption that fire induced circuit failures will not occur for at least 10-minutes after operators confirm the existence of a challenging active fire. As stated in NFPA 805, Section 4.2.3, the use of recovery actions automatically implies the use of the performance-based approach. Therefore, the validity of this assumption should be demonstrated by a performance-bas'ed approach consistent with NFPA 805, Section 4.2.4, which states that the additional risk presented by the use of recovery actions shall be evaluated. When the risk evaluation option of the performance-based approach is selected, the calculations required by Section 4.2.4.2 are required by NFPA 805.
28
- Request for Additional Information November 30, 2009 Attachment G of the Transition Report states that that "linking confirmation of a challenging active fire to the beginning of the 10 minute time frame before any spurious equipment operations occur is consistent with the current licensing basis." Identify the license bases documents, which support this position.
NFPA 805, paragraph 4.2.4.1.6, requires that all recovery actions must be shown to be feasible, including consideration of the time available to perform the action before the plant experiences a non-recoverable condition. Section 1.5 of NFPA 805, states that fire protection features shall be capable of providing reasonable assurance that, in the event of a fire, the plant is not placed in an unrecoverable condition. The requirements for the engineering analyses used to support the performance-based fire protection design that fulfills the goals, objectives, and criteria provided in Chapter 1 are provided in Chapter 2 of NFPA 805. With regard to the application of the performance based approach, NFPA 805, Section 2.2 (g), states: When applying a performance-based approach, perform engineering analyses to demonstrate that performance based requirements are satisfied. These analyses shall include, for example, engineering evaluations, probabilistic safety assessments, or fire-modeling calculations (see Section 2.4).
In addition, LAR, Attachment G, states: "The decision to staff the standby shutdown facility (SSF) is tied to confirmation of a challenging active fire." There is concern that the application of this approach could result in a significant delay of transfer of control to the SSF, thereby increasing the amount of time the plant is vulnerable to the effects of spurious actuations. The occurrence of fire-induced maloperations during this time could have a significant impact on the capability to achieve and maintain safe shutdown (SSD) conditions from the SSF. A similar concern was previously identified for ONS during a 2002 inspection, as described Inspection Report 50-269/02-03, 50-270/02-03,and 50-287/02-03.
RAI 3-28 RESPONSE:
Upon confirmation of an active fire in those areas of the plant where the Standby Shutdown Facility is the credited method for safe shutdown and where a fire could adversely affect the safe shutdown capability of the Standby Shutdown Facility (i.e., Standby Shutdown Facility transfer control cables) an operator will be dispatched to transfer control to the Standby Shutdown Facility. This transfer will be completed within 10 minutes of confirmation of the fire.
An example of a fire area that meets this set of conditions is the Auxiliary Building fire area. An example of a fire area that does not meet this set of conditions is the Protected Service Water Building fire area. The Standby Shutdown Facility is the credited method of safe shutdown for the Protected Service Water Building but a fire in the Protected Service Water Building will not adversely affect the safe shutdown capability of the Standby Shutdown Facility.
As noted in the ONS response to RAI 3-1, the term challenging active fire was used incorrectly in Attachment G. Attachment G should have stated ... "Embedded in the compliance strategy is the assumption that the 10 minute time frame does not start until confirmation of an active fire."
Attachment G will be corrected and submitted in the next LAR supplement.
The NRC recognized the term "confirmed active fire" in the June 22, 2005 NRC Supplemental Inspection Report 05000269/2005006, 05000270/2005006, and 05000287/2005006:
"... The SSF Emergency OperatingProcedure was revised in 2003 to include manning the SSF on a confirmed active fire in the main control room, cable room, or turbine building. A confirmed active fire was defined as a locally observed fire with smoke and eitherradiantheat or visible flarhe."
29
- Request for Additional Information November 30, 2009 RAI 3-29:
Provide documentation which demonstrates that non-essential circuits and cables which share a common enclosure (junction box, cable tray, etc.) with circuits and/or cables of required shutdown equipment during all modes of operation are provided with appropriately sized electrical protective devices (fuses, circuit breakers, or other suitable isolation device). In addition, provide documentation that (a) demonstrates that the scope of the review performed was adequate to address NFPA 805 nuclear safety capability concerns during all modes of plant operation, and (b) identifies any deficiencies identified during the review and (c) describes how the identified deficiencies have been resolved.
Section 2.4.2.2.2, NFPA 805, states that circuits that share common enclosure with circuits required to achieve nuclear safety performance criteria shall be evaluated for their impact on the ability to achieve nuclear safety performance criteria. The LAR and supporting Calculation OSC-9659, (Draft Oconee Nuclear Safety Capability Assessment), assume that adequate electrical circuit protection was included as part of ONS's plant electrical design and has been maintained as part of the design change process. Based on this assumption, an evaluation of the common enclosure concern was not performed. Although proper circuit protection and cable sizing were likely included as part of the original plant electrical design, the adequacy of protection provided in the current as built plant should be verified and demonstrated.
RAI 3-29 RESPONSE:
The ONS Breaker Coordination study, performed in support of NFPA 805, includes a review of cable sizing to support a common enclosure validation. The cable sizing review is being performed in the second phase of the breaker coordination study. The first phase, Pass 1, includes all voltage levels of "safe shutdown related" power supplies contained in AREVA Engineering Information Record (EIR) 51-5044354, Oconee Appendix R Fire Safe Shutdown Analysis, and/or the associated Appendix R Database Management System. All power supplies required by PRA and Non Power Operations are included in the breaker coordination study scope of "safe shutdown related" power supplies.
Pass 1 will determine if the protective device is coordinated, in general, for all credible levels of fault downstream taking into account the maximum available short circuit current adjacent to the bus. Therefore, Pass 1 conservatively does not take into account any specific cable length that would reduce the available fault current. If Pass 1 is successful, then coordination is validated.
If Pass 1 is not successful, additional passes will be required to justify coordination in a given plant fire area.
The second phase, Passes 2 through 6, will address all analysis beyond Pass 1 as needed to support the breaker coordination validation. If the results from Pass 1 demonstrate that coordination potentially does not actually exist on a selected power supply, then further analysis will be performed for those feeders on the selected power supply which are shown to be uncoordinated. The cables associated with these uncoordinated feeders will be identified and routed by Fire Area in order to determine the impact to the associated Fire Areas/Scenarios.
This phase will also document the cable damage curve for selected loads to provide a sampling to validate proper cable sizing to support the conclusion in the nuclear safety capability assessment with regards to common enclosure. The cables associated to the uncoordinated feeders will be documented in the Appendix R Database Management System and utilized as an input to the Nuclear Safety Capability Assessment, Fire PRA Model, and the Non Power Operations Pinch Point Analysis.
30
- Request for Additional Information November 30, 2009 Phase one of the breaker coordination study was initiated on May 22, 2009 and is scheduled to be completed by December 31, 2009. Phase two schedule will be dependent upon the scope identified during phase one of the ONS breaker coordination study. The current project completion date for all phases of the breaker coordination study is July 31, 2010.
In addition, recent modifications were reviewed with regards to the adequacy of protection provided in the current "as built" plant as compared to the original plant electrical design.
Calculations OSC-8034 and OSC-8035 are Electrical Design Input Calculations that were utilized as inputs to the modifications that installed three Motor Control Centers each at Units 2 and 3.
Based on the above review of selected cable information being performed in the new breaker coordination calculation and the review of recent plant modifications, the assumption that adequate electrical circuit protection was included as part of the ONS plant electrical design and has been maintained as part of the design change process as stated in OSC-9659, (Draft Oconee Nuclear Safety Capability Assessment provided for review during the audit), is valid.
RAI 3-30:
Provide your evaluation of spurious equipment actuations and/or mal-operations (including multiple spurious operations) during non-power operation.
NFPA 805 requires an evaluation of fire effects during all plant operating modes and conditions, including shutdown and decommissioning.
RAI 3-30 RESPONSE:
ONS calculations, OSC-9268, NFPA 805 Transition Non-Power Operations Component Selection and OSC-9313, NFPA 805 Transition Non-Power Fire Area Assessments (Pinch Points Analysis), have been revised. The calculations provide a greater level of detail in explaining how spurious equipment actuations and/or maloperations including multiple spurious operations have been analyzed in the evaluation of 'pinch points' for Non Power Operations. To summarize the calculation methodology, the component and cable selection process included all components with a potential for spurious operation including flow blockage and diversion and associated cables causing spurious operations. Any cable hit within a fire zone was considered an adverse impact on the component and any related Key Safety Function success path(s).
This resulted in a very conservative analysis.
RAI 3-31:
Identify the specific document(s) credited for demonstrating conformance to the nuclear safety capability criteria of NFPA 805, Section 2.4.2.
In multiple sections of Table B-2, AREVA Engineering Information Record (EIR) 51-5044354-003, Revision 3, October 18, 2008, is identified as containing the pre-transition license basis for compliance to the post-fire safe shutdown requirements of Appendix R. The extent to which EIR 51-5044354-003, Revision 3 conforms to the provisions of Nuclear Energy Institute (NEI) 00-01 is documented in Calculation OSC-9291, "NFPA 805 Transition B-2 Table," Revision 1.
Therefore, from Table B-2 it would appear that the ONS nuclear safety capability design basis documentation consists of EIR 51-5044354-003, Revision 3 and Calculation OSC-9291.
31
- Request for Additional Information November 30, 2009 However, during the site audit, a draft version of Calculation OSC-9659, "ONS Nuclear Safety Capability Assessment," was provided for review. This calculation states that it was performed to demonstrate how ONS meets the requirements of 10 CFR 50.48(a) and 10 CFR 50.48(c).
However, OSC-9659 is not referenced in the ONS Transition Report. As a result, it is not clear which documents (EIR 51-5044354-003 and OSC-9291 OR OSC-9659) contain the ONS basis for demonstrating conformance to the nuclear safety capability criteria of NFPA 805.
If OSC-9659 is to be cited as the ONS nuclear safety capability design basis document, provide the following:
- the final, approved version of OSC-9659,
- a summary of changes in the assumptions, methodology, criteria, and results described in OSC-9659 from those described in EIR 51-5044354-003 and OSC-9291, and
- a summary of issues identified by OSC-9659 that were not previously identified by EIR 51-5044354-003 and OSC-9291 and a description of ONS plan to reconcile each issue prior to transition.
Section 2.7.1.2 of NFPA805 requires the licensee to establish a fire protection program design basis document based on those documents, engineering evaluations, and calculations that define the fire protection design basis for the plant. As described in Standard Review Plan (SRP) Section 1.3.7, this document shall include fire hazards identification and nuclear safety capability assessment, on a fire area basis, for all fire areas that could affect the nuclear safety or radioactive release performance criteria defined in NFPA 805, Chapter 1. SRP, Section 1.3.4, further clarifies that the NRC reviewer should verify that the licensee has evaluated the existing post-fire safe shutdown analysis methodology against the requirements and criteria specified in NFPA 805, Chapters 1 and 2, regarding Intent, General Approach, Assumptions and Engineering Analyses.
RAI 3-31 RESPONSE:
ONS calculation OSC-9659, Oconee Nuclear Safety Capability Assessment for Units 1, 2, and 3, will document conformance to the nuclear safety capability criteria of NFPA.805 Section 2.4.2. The latest version of the AREVA document EIR 51-5044354, containing the pre-transition licensing basis for compliance to the post-fire safe shutdown requirements of Appendix R, is Revision 4 dated February 26, 2009. This document was used as a design input to OSC-9659.
ONS calculation OSC-9291, NFPA 805 Transition B-2 Table, which formally documents the comparison of EIR 51-5044354 to NEI 00-01, will be revised as necessary to reflect the latest version of the EIR.
A key assumption in the Nuclear Safety Capability Assessment is that adequate and accurate preliminary design details have been provided for the proposed modifications which will be listed in revised Attachment S of the LAR supplement. The Nuclear Safety Capability Assessment will evaluate the use of the added and/or modified equipment for meeting the nuclear safety performance criteria in NFPA 805.
The major changes in the methodology of OSC-9659 from those described in the EIR are as follows:
- Documenting the ability to achieve 'safe and stable' conditions instead of cold shutdown as required by Appendix R 32
Enclosure 1: Request for Additional Information November 30, 2009
- Splitting up of the Balance Of Plant fire area into 2 distinct fire areas; the Turbine building (Fire Area TB) and Auxiliary building (Fire Area AB)
" Changing the shutdown methodology of the blockhouses and turbine building from a Standby Shutdown Facility shutdown to a control room shutdown
" Changing all control room shutdowns to credit the Protected Service Water modification A summary of issues will be included in Oconee calculation OSC-9292, NFPA 805 Transition B-3 Table. A disposition of each issue will be included in the B-3 table.
Review and revise the Table B-2 alignment basis statements as necessary to ensure they appropriately bound the cited criteria.
Consistent with the guidance of NEI 04-02, Table B-2 provides a comparison of the existing safe shutdown methodology to applicable sections of NEI-00-01. In certain cases, the alignment basis statements of Table B-2 do not adequately address the specific NEI 00-01 criteria evaluated. For example, the text on Page 1, Table B-2, indicates that the following specific NEI 00-01 criteria were compared to the following existing methodology:
- NEI 00-01 criteria and assumptions and
" The requirements of Appendix R Sections III.G.1, III.G.2 and III.G.3.
Although Table B-2 states that, the ONS safe shutdown analysis (SSA) aligns with the NEI guidance, the "alignment basis" states:
"A deterministic methodology is utilized to assess conformance with Appendix R."
Simply stating that a deterministic analysis was performed does not provide sufficient.
information to conclude that the analysis conforms to the cited criteria.
Provide revised Table B-2 alignment basis statements, such as this example, to clearly indicate that the existing ONS methodology was compared to and conforms with the cited NEI 00-01 criteria.
As described in SRP, Section 1.3.4, the reviewer should verify that the licensee has evaluated the existing post-fire safe shutdown analysis methodology against the requirements and criteria specified in NFPA 805, Chapters 1 and 2, regarding intent, general approach, assumptions and Engineering Analyses. One approach acceptable to the NRC staff for determining the extent that the existing (pre-transition) SSA meets this section is described in NEI 04-02, Revision 1, which recommends a line-by-line comparison of the existing SSA against the methodology provided in Chapter 3 of NEI 00-01, Revision 1.
RAI 3-33 RESPONSE:
The Calculation OSC-9291, NFPA 805 Transition B-2 Table, is being revised to incorporate a greater level of detail on the alignment bases statements of the B-2 table to ensure reviewers 33
- Request for Additional Information November 30, 2009 can conclude conformance to the cited NEI 00-01 revision 1 sections as noted above. The revised B-2 table will be included in the revised LAR supplement.
RAI 3-34:
Describe how the impact of fire damage to fire detection and suppression systems required by NFPA 805, Section 4.2.3 or 4.2.4, were evaluated at ONS.
Recent inspections have identified instances where fire damage to unprotected cables could significantly affect the operability of fire suppression systems. At two facilities visited, fire damage to plant cables connected to manual pump start switches located in the control room could result in a failure of the fire suppression water supply system to automatically start as designed. Until at least one pump was manually started in the pump house, no water would be provided to fire suppression systems and standpipe hose connections in safety-related areas of the plant. The loss of fire suppression is of particular concern if the fire damage were to occur in areas where suppression is credited for preventing or mitigating fire damage to shutdown equipment and cables (e.g., areas that rely on automatic suppression and separation distance in lieu of 3-hour fire rated barriers). The methodology described in NEI 04-02 does not include an evaluation of fire suppression system equipment and cables.
Section 4.2.1 of NFPA 805, requires one success path necessary to achieve and maintain the nuclear safety performance criteria to be free of fire damage. An evaluation should be provided that demonstrates that the performance of the fire detection and suppression systems needed to protect the success path are not degraded by the fire.
RAI 3-34 RESPONSE:
As identified in AREVA document 51-5044354-004, Oconee Appendix R Safe Shutdown Analysis, the required Classical Fire Protection equipment and required positions were conservatively identified by the ONS Fire Protection Engineer and Communications Engineer and included in Appendix R Database Management System so that an analysis can be performed to assure that this equipment remains available on a fire area basis. This equipment includes components such as the fire pumps, deluge valves, power to Fire Detection Panels and plant communications equipment.
Those classical fire protection components that were already in the Appendix R Database Management System and part of the existing ONS Safe Shutdown Equipment List were also "flagged" as classical fire protection. Those classical fire protection components that were not already in the Appendix R Database Management System were added and "flagged" as classical fire protection only. Cable selection and cable routing were then performed for these new components in accordance with the methodology defined in 51-5044354-004, Sections 3.1 and 3.2, so that reports can be generated and an analysis performed for the classical fire protection components in the Nuclear Safety Capability Assessment.
The impact and availability of the fire detection and suppression systems needed to protect the success path required by NFPA 805 Section 4.2.3 or 4.2.4 will be included in the Nuclear Safety Capability Assessment. Any deviations to the required success path will be evaluated and included in the LAR supplement.
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Enclosure 1: Request for Additional Information November 30, 2009 RAI 3-35:
Describe how High/Low pressure interfaces were identified and evaluated. In addition, describe how FAQ 06-0006 was applied in the ONS nuclear safety performance assessment.
Attachment H of the ONS Transition Report identifies FAQ 06-0006 as one of the FAQs used in the development of the ONS transition process. However, there is no discussion of this FAQ in the Transition Report. If this FAQ was utilized as guidance in the performance of the nuclear safety performance assessment, a discussion should be included explaining how the FAQ was applied at ONS.
RAI 3-35 RESPONSE:
NEI 00-01, revision 1 defines High/Low pressure interfaces as a subset of components considered for spurious operation that involves reactor coolant pressure boundary components whose spurious operation can lead to an unacceptable loss of reactor pressure vessel/Reactor Coolant System inventory via an interfacing system loss of coolant accident. This definition has been deemed acceptable by the NRC as documented in FAQ 06-0006 Closure' Letter dated March 12, 2007 (ML070030117) and clarified in the October 18, 2007, NFPA 805 FAQ Meeting Summary dated November 28, 2007 (ML073200763).
The identification and failure analysis of high/low interface valves was performed in accordance with NEI 00-01, revision 1, in support of the Oconee Appendix R Safe Shutdown Analysis and input into the Appendix R Database Management System. The results of the analysis of the data contained in the Appendix R Database Management System will be carried forward into ONS calculation, OSC-9659, Nuclear Safety Capability Assessment.
Attachment B of the LAR, Table B-2, Nuclear Safety Capability Assessment Methodology Review, will be updated to address the methodology utilized in evaluating high/low interface valves in comparison to the guidance provided in NEI 00-01, revision 1. This will include a statement referencing the utilization of FAQ 06-0006 to determine the acceptable definition of high/low pressure interface valves. Also, calculation OSC-9659, Nuclear Safety Capability Assessment, will include a basis statement in the methodology section pertaining to high/low interface valves. This basis statement will include the use of NEI 00-01, revision 1.
RAI 5-15:
Provide information concerning which and how operator actions were evaluated for risk. Include in this discussion which actions are specifically modeled in the PRA and/or Fire PRA, which were evaluated using a quantitative assessment, which were evaluated using a qualitative bounding assessment, and which, if any, were not evaluated for their impact on risk. In addition, provide the results of the risk increase (CDF) evaluation for each operator action. Table G-2 of the TR should include some of this information (e.g., whether modeled, how evaluated) for each OMA.
Table G-2 of the TR provides a listing of OMAs and their binning, but is not clear which were evaluated using which methods. In addition, while the bounding analysis is briefly discussed, sufficient detail to judge the methodology is not provided. For example, Section G.5.2.2 notes in the first paragraph that "Qualitative methods were used to assess the additional risk presented by the use of Bin D OMAs," yet in the third paragraph states, "Bin D actions not directly associated with the SSF operation were qualitatively addressed for additional risk."
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Enclosure 1: Request for Additional Information November 30, 2009 The LAR does not provide sufficient information for the staff to review the additional risk presented by the use of recovery actions. Section 4.2.4 of NFPA 805 requires that the additional risk of relying on recovery actions be evaluated.
RAI 5-15 RESPONSE:
Duke will revise the treatment of additional risk of recovery actions as discussed in the response to RAI 3-18. The results of the implementation of this process will be reflected in the revised Transition Report which will be submitted with the LAR supplement.
RAI 5-16:
Provide an evaluation of OMAs that are transitioning to recovery actions that is consistent with the definition of a recovery action in the Draft Regulatory Guide, DG-1218. In the response, identify the OMAs that changed classification from defense-in-depth to recovery action and provide an evaluation of the additional risk presented by these recovery actions.
In Attachment G of the TR, ONS classified OMAs as recovery actions based on the assumption that the definition of an "emergency control station" in RG 1.189, rev. 1, is the same as "primary control station" in Section 1.6.52 of NFPA 805. However, DG-1218 provides additional guidance on the NRC's definition of "primary control station" and it is different then an "emergency control station" as defined by RG 1.1.89. Because of this difference, there is insufficient information in the LAR for the staff to complete its review. Section 4.2.4 of NFPA 805 requires that the additional risk of recovery actions needs to be evaluated.
RAI 5-16 RESPONSE:
As discussed during the November 12, 2009 Public Meeting, Duke will revise the treatment of recovery actions. This treatment will include the definition of recovery action and primary
-control stations. See response to RAI 3-18 for additional information. The results of the implementation of this process will be reflected in the revised Transition Report.
RAI 5-18:
The nuclear safety capability assessment is, by nature, a spatial analysis. In order to fully understand the analyses described, provide the fire protection plant layout drawings that define the:
- Fire PRA analysis areas/compartments
- Current licensing basis fire areas, fire zones, and buffer zones
- Credited fire barriers that define these areas
- Areas covered by detection and suppression systems
- No transient combustible areas
" Major fire hazards RAI 5-18 RESPONSE:
The plant layout drawings containing the requested information were provided on the shared website.
36
Enclosure 1: Request for Additional Information November 30, 2009 Due to the relatively small number of fire areas, fire zones (rather than fire areas) were chosen for convenience as the physical analysis units or fire compartments for development of the Fire PRA. Cable information was available in the Appendix R Database Management System at both the fire area and fire zone level. While partitioning the plant into a greater number of compartments based on fire zones rather than fire areas is more efficient, it has the added effect of increasing the burden for the analysis of multi-compartment fire scenarios. As described in the Fire Scenario Report, the multi-compartment analysis is addressed in part by the consideration of targets in adjacent compartments (fire zones) during the development of individual fire scenarios. If a target was within the zone of influence for a given fire scenario, it was included in the scope of equipment assumed to be damaged by the fire even if that target was located or routed in another compartment. Consequently, the use of fire zones versus fire areas had no impact on the fire risk quantification results.
RAI 5-19:
Concerning the May 31, 2009, submittal, enclosure 1, pages 4 & 5: "The following modifications will be removed from Attachment S: Re-route of 3TD cables routed over 3TC ...
[T]he MSIV modification discussed above will mitigate an overcooling event and reduce base risk to a level that will allow self approval going forward." In the February 9, 2009, response to the Acceptance Review, ONS indicated a fire core damage frequency (CDF) reduction of 1.1 7E-5/yr due to this planned re-route of the 3TD cables. Will this credit no longer be taken in the "going-forward" FPRA supporting the transition? Has the FPRA logic model been amended to remove this credit? What, if any, effect does the removal have on the total "going-forward" fire CDF and large early release frequency (LERF) and any delta-fire CDFs and LERFs associated with plant change evaluations being credited for the transition?
RAI 5-19 RESPONSE:
Credit for the 3TD modification will not be taken in the Fire PRA. The effect of only removing credit for the 3TD modification caused CDF and LERF to increase, but not enough to impact the acceptability of changes in CDF and LERF in terms of Regulatory Guide 1.174 acceptance guidelines. The impact on the delta-fire CDFs and LERFs will be reflected in the Fire Risk Evaluations which are scheduled for submittal as part of the LAR supplement.
RAI 5-20:
The May 31, 2009, submittal, attachment 1: The one remaining modification that ONS has committed for implementation is modeled in the "going-forward" FPRA supporting the transition is that for PSW. Its completion date is scheduled for December 2010. Except for the remaining commitments with due dates "to be provided in RAI response," all others have completion dates later than that for PSW, so should not affect the "going-forward" FPRA prior to completion of the PSW modification. Will any of those listed with dates "to be provided" occur prior to the PSW modification date of December 2010? If so, will they be modeled in the "going-forward" FPRA supporting the transition? If so, how will they be modeled and what will be their effect on total and delta-risk?
RAI 5-20 RESPONSE:
The following is an excerpt from the table included in Attachment 4 of the May 31 submittal concerning modification commitments and due dates. Only the commitments with due dates "to be provided in RAI response" are listed.
37
Enclosure 1: Request for Additional Information November 30, 2009 Commitment Due Date RCPs spurious start and ability to ensure pump trip - Scoping To be provided in RAI is not complete. response HPI Pressure pump spurious start and ability to ensure pump To be provided in RAI trip - Scoping is not complete. response IN 92-18 Study - In progress To be provided in RAI response The first two modifications listed above will be removed from the commitments provided in . The RCP spurious start and RCS vent valve modifications are modeled in the PRA, but they will be turned off when calculating delta risk and when calculating total risk for determining where ONS stands in regards to the RG 1.174 acceptance guidelines. Modifications related to the IN 92-18 study are still committed, but they are not credited in the Fire PRA currently. The Fire PRA does not currently credit the recovery of spurious valve actuations; therefore, these modifications are not expected to have an impact on risk. However, the Fire PRA will be reviewed after the scope of the IN 92-18 resolutions is fully evaluated.
RAI 5-21:
Page 42, Section 4.6.2, ONS LAR 2008-01, states that "Another aspect of risk criteria is establishing performance criteria. These performance criteria will be established for items within the NFPA 805 monitoring scope, regardless of their ability to be measured using risk significant criteria."
It appears that the second sentence contradicts the first, namely that "risk performance criteria" will be developed even without the ability to be measured against risk significance. Correct or explain why there is no contradiction.
RAI 5-21 RESPONSE:
As described in the response to RAI 6-1, the monitoring program will utilize an expert panel process to determine the risk significance of the Fire Protection structures, systems and components either using the quantitative risk criteria explicitly established in the Fire PRA or by utilizing the compartment fire risk contribution data to qualitatively assess the risk significance of Fire Protection features within the compartment for their contribution to fire risk reduction. The apparent contradiction was due to the poorly chosen words used to describe the process to be utilized. Risk criteria will have to be established for monitored structures, systems, and components, irrespective of whether they were explicitly modeled in the Fire PRA or not. In some cases if the monitored structures, systems, and components are explicitly modeled in the PRA, risk criteria will be tied to the assumptions for availability and reliability of the system utilized in calculating the non-suppression probability values. If the structures, systems, and components are not explicitly modeled in the Fire PRA, parameters such as the compartment CDF, LERF, CCDP and RAW values will be used to assess the potential risk significance of the Fire Protection structures, systems, and components located within that compartment on a qualitative basis using the expert panel. The process being utilized will be revised for clarity in the LAR supplement.
RAI 5-22:
Regarding Assumption 19, page C-3, Attachment C, ONS LAR 2008-01, states that "The interactions of failures caused by a fire in fire area BH3 and their effects on the standby busses 38
- Request for Additional Information November 30, 2009 and the Unit 1 and Unit 2 main feeder busses are not fully analyzed. The current SSD strategy for a fire in fire area BH3 is based on a simultaneous Unit 1, Unit 2 and Unit 3 shutdown from the SSF. The SSD analysis for fire area BH3 does not support a multiple unit shutdown from the control room for Unit 1 and 2 using their respective standby busses and main feeder busses and also simultaneously shutting down Unit 3 from the SSF."
Is this assumption modeled as is in the FPRA? Estimate the effect on fire risk (CDF and LERF) if this assumption is changed (such as fully analyzing the interactions of failures), e.g., via sensitivity analysis.
RAI 5-22 RESPONSE:
This assumption is not modeled in the Fire PRA. The Fire PRA failures assumed in BH3 are taken directly from the cable data provided in the Appendix R Database Management System.
There are no cable hits in BH3 for the Unit 1 or 2 main feeder buses or the standby bus located in BH12. Interlock failures in the Appendix R Database Management System are not typically included in the Fire PRA provided the dependencies, including power supply dependencies, are properly addressed in the fault tree. If further analysis indicates BH12 components are impacted, then the CDF & LERF will increase.
RAI 5-23:
For the BOP, pages 60 and 61, Attachment C, ONS LAR, discuss three sets of variances for which plant change evaluations and delta-risk calculations are performed. Each involves an unallowed or non-feasible manual action. The following statement is made: "The change evaluation determined that upon completion of the modifications the variances will be acceptable based upon: [1] The measured change in CDF and LERF [2] Adequate defense-in-depth and safety margins are maintained."
The modifications are cited in Attachment S of the LAR as yet to be developed and not yet incorporated into the FPRA. Do the calculated fire delta-CDFs and LERFs presume the modification will be such as to retain the unallowed or non-feasible manual actions? If not, discuss how these deltas were estimated. (e.g., if the non-feasible manual action was assumed, was it assigned a failure probability = 1 for the delta calculations? If not, why not?) Also, since the fire delta-CDFs and LERFs lie in Region II of the Regulatory Guide (RG) 1.174, "An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis," numerical acceptance guidelines, discuss how ONS has assured that the total CDF and LERF, including fire, lie below 1 E-4/y and 1E-5/yr, respectively.
RAI 5-23 RESPONSE:
The uninstalled modifications credited in the Fire PRA were also credited in the calculation of delta-CDF and delta-LERF. Delta-CDF and Delta-LERF will be calculated for the additional un-allowed or non-feasible manual actions necessary for crediting the modifications in the compliance strategy. The list of new actions is currently under development and will be provided in the LAR supplement. With regard to the second part of this RAI, rather than evaluate each individual change for acceptability, it is Duke's position that the reduction in risk as a result of committed modifications more than compensates for the increase in risk associated with non-compliances. As a result of transition to NFPA 805, the overall risk at the plant level decreases and therefore meets the numerical acceptance guidelines of RG 1.174.
39
Enclosure 1: Request for Additional Information November 30, 2009 RAI 5-24:
In regard to DID actions, pages G-12 and G-13, Attachment G, to the LAR, states that "the additional effort per NUREG 1852 [Demonstrating the Feasibility and Reliability of Operator Manual Actions in Response to Fire] does not add measurable benefit." Since NUREG-1 852 addresses OMA feasibility and reliability criteria in a qualitative manner, and DID actions are addressed qualitatively by ONS, provide the basis for this statement, e.g., what constitutes the "additional effort" and why it would not add "measurable benefit." Also, does the feasibility criteria listed in Table G-1, Section G.6.2, align with those from NUREG 1852? If there is any misalignment, such that the Table G-1 criteria exclude or inadequately reflect the NUREG-1852 criteria, provide the basis and discuss how assurance is provided that an important criterion is not overlooked.
RAI 5-24 RESPONSE:
ONS has re-evaluated the variances from the deterministic requirements for all fire areas with respect to the equipment necessary to ensure safe and stable conditions in any plant mode.
The variances from the deterministic requirements are being characterized as either separation issues, previously documented Operator Manual Actions or as degraded Fire Protection systems and features. Previously documented Operator Manual Actions are being screened to ensure they do not take place in the Main Control Room or Primary Control Station before they are called Recovery Actions. Once identified as a Recovery Action, they are being further screened to determine ifthey are required to demonstrate the availability of a success path for the Nuclear Safety Performance Criteria. Only those actions required to demonstrate the availability of a success path for maintaining the plant in a safe and stable condition will be evaluated for additional risk in the Fire Risk Evaluations. Consequently, the change evaluations are being revised and titled as Fire Risk Evaluations. Any recovery actions retained in plant procedures will be evaluated for feasibility using the criteria in Table G-1 of the LAR dated October 31, 2008 (based on NFPA 805 Appendix B.5.2(e) and NEI 04-02, Revision 1). This closely mirrors the information discussed in NUREG 1852. The differences between Table G-1 and NUREG-1852 are shown in the attached table and are semantic in nature. All important criteria are considered by NEI 04-02.
As discussed in Attachment G of the LAR, the reliability of recovery actions, if modeled specifically in the Fire PRA, is addressed using Fire PRA methods (i.e., Human Reliability Analysis). Although many recovery actions were not modeled in the Fire PRA, the cables that, if damaged by fire would prompt the actions, will be included as variances from the deterministic requirements. Therefore the risk of the variances includes/bounds the risk of the actions.
Based on the acceptably low risk of the variances without the actions, further reductions possible by calculating human error probabilities and modeling the actions would not be expected to change the conclusions of the LAR.
40
- Request for Additional Information November 30, 2009 Criteria NUREG-11852 Final Report (ML ML073020676) ONS LAR Table G-1 (10/31/08)
Section 3.1 Diagnostic (Available Indications) Item 2 Instrumentation In addition to the SSCs needed to directly perform the desired actions, other supporting equipment may also be Consider availability of systems and indications essential to required, including (to the extent required for successful perform the recovery action.
performance of each operator manual action)
- indicationsnecessary to show the need for the manual actions, enable their performance, and verify their successful accomplishment (if not directly observable).
Environmental (Environmental Factors) Item 8 Considerations The analysis should show that the actions can be performed under the expected environmental factors that will be When recovery actions are necessary in the fire area under encountered. consideration or require traversing through the fire area under consideration, the analysis should demonstrate that the area is tenable and that fire or fire suppressant damage will not prevent the recovery action from being performed.
Item 4 The lighting (fixed and/or portable) should be evaluated to ensure sufficient lighting is available to perform the intended action.
Staffing The number of available personnel (staffing), exclusive of Item 7 Fire Brigade members, needed to perform the actions should be consistent with the analysis. Walk-through of operations guidance (modified, as necessary, based on the analysis) should be conducted to determine ifadequate resources are available to perform the potential recovery actions within the time constraints (before an unrecoverable condition is reached), based on the minimum shift staffing. The use of essential personnel to perform actions should not interfere with any collateral industrial fire brigade or control room duties.
41
- Request for Additional Information November 30, 2009 Criteria NUREG-1852 Final Report (ML ML073020676) ONS LAR Table G-1 (10/31/08)
Section 3.1 Communications In addition to the SSCs needed to directly perform the Item 3 desired actions, other supporting equipment may also be required, including (to the extent required for successful The communications system should be evaluated to performance of each operator manual action) determine the availability of communication, where required for coordination of recovery actions.
- necessary communications.
Special Tools (Portable Equipment) Item 5 In addition to the SSCs needed to directly perform the Any tools, equipment, or keys required for the action -
desired actions, other supporting equipment may also be should be available and accessible. This includes required, including (to the extent required for successful consideration of SCBA and personal protective equipment performance of each operator manual action) if required. (This includes staged equipment for repairs).
- necessary portable equipment.
- necessary personnelprotection equipment Training (Procedures and Training) Item 10 There should be plant -procedurescovering each operator Training should be provided on the post-fire procedures manual action required to achieve and maintain hot and implementation of the recovery actions.
shutdown and trainingfor each operator on the procedures.
Accessibility (Equipment Functionality and Accessibility) Item 8 The analysis should show that (1) the functionality of When recovery actions are necessary in the fire area under equipment and cables needed to implement operator consideration or require traversing through the fire area manual actions to achieve and maintain hot shutdown will under consideration, the analysis should demonstrate that not be adversely affected by the fire, and (2) the equipment the area is tenable and that fire or fire suppressant damage will be available and readily accessible consistent with the will not prevent the recovery action from being performed.
analysis.
42
- Request for Additional Information November 30, 2009 Criteria NUREG-1852 Final Report (ML ML073020676) ONS LAR Table G-1 (10131/08)
Section 3.1 Time (Time Available) Item 9 An analysis should be prepared to evaluate the feasibility Sufficient time to travel to each action location and perform and reliability of operator manual actions. The analysis the action should exist. The action should be capable of should determine that adequate time exists for the operator being identified and performed in the time required to to perform the required manual actions to achieve and support the associated shutdown function(s) such that an maintain hot shutdown from a single fire. The adequate unrecoverable condition does not occur. Previous action time should reasonably account for all important variables, locations should be considered when sequential actions including (1) differences between the analyzed and actual are required.
conditions, and (2) human performance uncertainties that may be encountered.
Procedures (Procedures and Training) Item 6 There should be plant procedures covering each operator Written procedures should be provided.
manual action required to achieve and maintain hot shutdown and training for each operator on the procedures.
Periodic Drills (Demonstrations) Item 1 There should be periodic demonstrations of the manual The proposed recovery actions should be verified in the actions, consisting of actual executions of the relevant field to ensure the action can be physically performed actions to the extent sufficient to show continued under the conditions expected during and after the fire proficiency in performing the actions. event.
Item 11 Periodic drills that simulate the conditions to the extent practical (e.g., communications between the control room and field actions, the use of SCBAs if credited, the appropriate use of operator aids).
43
Enclosure 1: Request for Additional Information November 30, 2009 RAI 5-25:
Page G-14, Section G.4.6.2, Attachment G, ONS LAR 2008-01, states that "Each of the criteria in Table G-1 were assessed for the recovery and DID actions listed in Table G-2. The results of the assessment are included in a calculation entitled 'Recovery and Defense-in-Depth Action Evaluation in Support of Nuclear Safety Capability Assessment.' This calculation also includes a summary of thermal hydraulic (TH) analysis used to evaluate the timing of actions."
Explain the process by which MSOs that could result in unrecoverable plant conditions based on the plant's current TH analyses were addressed. For example, was there consideration of the time available to complete the action before the plant is placed in an unrecoverable condition or unrecoverable equipment damage occurs for time critical actions? In the event a new MSO combination is identified that requires further review in the Nuclear Safety Capability Assessment, and consideration is being given to using either a DID action or a recovery action to mitigate the plant impact, will the adequacy of the existing TH analyses be reviewed? Will additional TH analysis be performed if necessary? If the new TH analysis concludes that adequate time is not available to complete the proposed action, will an alternative strategy be implemented to resolve the MSO?
RAI 5-25 RESPONSE:
Multiple Spurious Operations were not addressed in the referenced thermal hydraulic analysis.
The purpose of this analysis was to demonstrate the ability to perform a single steam generator natural circulation cool down to Low Pressure Injection entry conditions.
Multiple Spurious Operations have been identified as variances from the deterministic requirements of NFPA 805 in Attachment C, NEI 04-02 Table B-3 Fire Area Transition, of the Transition Report and will be evaluated in accordance with draft calculation OSC-9314, NFPA 805 Transition Risk Informed, Performance Based Fire Risk Evaluation Methodology.
If the conclusion of the risk informed, performance based fire risk evaluation determines that a recovery action is the solution for an identified variance then the feasibility of that recovery action will be evaluated in accordance with draft calculation OSC-9535, Evaluation of Recovery Actions in Support Of Nuclear Safety Capability Assessment.
Guidance will be added to draft calculation OSC-9535 to determine the time critical aspect of a recovery action. Existing thermal hydraulic analyses will be reviewed to determine if they address the time critical aspect of any current or future recovery action and additional analyses will be conducted if required. If the result of the thermal hydraulic analyses demonstrates that it is not feasible to perform a recovery action within the allowable time, calculation OSC-9314 will drive the development of an alternative solution, e.g., a plant modification.
The final disposition of the Multiple Spurious Operations identified in Attachment C, NEI 04-02 Table B-3 Fire Area Transition, will be documented in the Transition Report to be submitted with the LAR supplement.
RAI 5-26:
Pages G-15 and G-17, Section G.8.1, Attachment G, of the LAR, states that "Industry test data as discussed in a recent draft revision to NEI 00-01 (ADAMS Accession No. ML080310056),
while not conclusive, supports the assumption that spurious operations will not occur 44
Enclosure 1: Request for Additional Information November 30, 2009 immediately upon exposing cables to fire [e]ffects." Section G.8.2, states, in referring to the 10-minute delay, that "[i]n conclusion, allowing a reasonable diagnostic time to define the appropriate safe shutdown strategy after confirmation of a challenging active fire is an appropriate risk-informed, performance-based (RI-PB) approach."
Discuss how these assumptions address the effects of fires that may be induced by high-energy arcing faults or similarly "very fast" growing fires.
RAI 5-26 RESPONSE:
The 10-minute delay for spurious actuations was not a Fire PRA assumption. Since targets within the zone of influence of a high energy arcing fault are assumed to be instantly on fire, there is no 10-minute delay in target damage. However, since loss of power may occur instantly, it may be argued that spurious actions are less likely to occur from high energy arcing fault scenarios. The same may not be true for "very fast" growing fires; however, the 10-minute delay for spurious actuations was not a Fire PRA assumption. All targets within the zone of influence were assumed to be damaged and the affected components were subject to both spurious operation and demand failure.
RAI 5-27:
On page V-3, Attachment V, of the LAR indicates that fire brigade response times longer than 20 minutes were not considered.
Discuss why fire brigade response times longer than 20 minutes apparently were not considered, at least via sensitivity analysis. If such a sensitivity analysis would have changed the results, discuss the implications.
RAI 5-27 RESPONSE:
No sensitivity analyses were performed; however, based on response times from 100 drills, only 2 drills failed to meet the 15 minute response time requirement and only 1 of these 2 drills slightly exceeded the 20 minute assumption for minimum required volume in support of the hot gas layer evaluation. Therefore, the brigade drill data adequately supports the 20 minute assumption. The hot gas layer evaluation supports two (2) objectives: 1) the zone of influence for IEEE 383 qualified cables determined for a given heat release rate is applicable provided the formation of a hot gas layer does not influence or extend the damage distance, and 2) the presence of a damaging hot gas layer (i.e., room burnout conditions) is an input to the multi-compartment analysis.
The volume assumed for the hot gas layer evaluation was based on the minimum required volume for zone of influence applicability which is greater (more conservative) than the volume for room burnout conditions due to hot gas layer. Since the compartments screened based on the larger minimum required volume (for zone of influence impact), the hot gas layer evaluation based on the smaller (room burnout) volume will also screen (i.e., the actual compartment volumes are greater than the volume associated with zone of influence applicability which is greater than the volume associated with reaching hot gas layer room burnout conditions). The fire brigade data supports the 20 minute assumption used to obtain the minimum volume for a given heat release rate. A longer assumed response time would be overly conservative given that, in some cases, the fire brigade can open a door to alleviate hot gas layer concerns and the brigade drill times were well within 20 minutes in 99 out of 100 drills. A shorter response time 45
Enclosure 1: Request for Additional Information November 30, 2009 would not change the results since the ONS fire compartments screened from hot gas layer impacts.
Fire Brigade response times will be within scope of the Monitoring program and as such the PRA assumptions of the 20 minute response time will continue to be maintained with the current 15 minute response time criteria.
RAI 5-28:
On page V-5, Attachment V, of the LAR, states that there is "No impact on quantification of Fire PRA or Change Evaluations (seismic-fire interaction is purely qualitative per NUREG/CR-6850, Fire PRA Methodology for Nuclear Power Facilities)." The responses appear to attempt to justify leaving these findings "Open."
Provide the planned resolutions of the findings.
RAI 5-28 RESPONSE:
Supporting requirement SF-A2-1 requires ONS to assess the potential impact of system rupture or spurious operation on post-earthquake plant response including the potential for flooding relative to water-based fire suppression systems, loss of habitability for gaseous suppression systems, and the potential for diversion of suppressants from areas where they might be needed for those fire suppression systems associated with a common suppressant supply. A finding was written up for the ONS treatment of this supporting requirement because, as stated in the NRC staff review report, "A seismic induced assessment of the potential for diversion of suppressants from areas where needed for fire suppression systems associated with a common suppressant supply was not conducted." This requirement is very similar to NFPA 805 requirement 3.6.4 which says that "Provisions shall be made to supply water at least to standpipes and hose stations for manual fire suppression in all areas containing systems and components needed to perform the nuclear safety functions in the event of a safe shutdown earthquake (SSE)." According to Attachment A of the LAR, entitled, NEI 04-02 Table B-1 Transition of Fundamental Fire Protection Program & Design Elements, this requirement has been satisfied per previous NRC approval, therefore it is Duke's position that no further action is necessary to close out this particular supporting requirement.
According to the NRC staff review of the ONS Fire PRA supporting requirement SF-A4 was considered not met, with the comment that the plant seismic response procedures covers seismically induced flooding, but not seismically induced fire. Duke initially considered revising the seismic response procedures to cross-reference the fire response procedures. While pursuing this option it was determined that there would be no benefit to such a revision for the following reason: There are two entry conditions for the fire response procedure, either via the fire alarm annunciator or the report of a fire. These entry conditions apply at all times and under any plant operating conditions. When the entry condition is met the operators will take the actions required in the fire response procedure, i.e., dispatch the fire brigade, determine the need to man the Standby Shutdown Facility, etc. A reference to the fire response procedure in the seismic response procedure is unnecessary, and therefore no further action is necessary to close out this supporting requirement.
Supporting requirement SF-A5 requires a review of fire brigade training procedures and an assessment of the extent to which training has prepared fire fighting personnel to respond to potential fire alarms and fires in the wake of an earthquake. It also requires review of the 46
Enclosure 1: Request for Additional Information November 30, 2009 storage and placement of fire fighting support equipment and fire brigade access routes, and an assessment of the potential that an earthquake might compromise one or more of these features. The NRC observation, resulting in a "not met" on this requirement states that "no assessment has been conducted on the potential that an earthquake might compromise one or more of the fire brigade [features]." However, it should be noted that, according to the Fire Brigade Guide (SOG) #1, General Response Procedure, "The fire brigade equipment is staged in various locations so that one single event (fire, flood, earthquake or terrorist act) will not render the fire brigade ineffective." Therefore, the fire brigade response during the event of a seismic event has been considered and no further action is necessary to close out this particular supporting requirement.
RAI 5-29:
The disposition to SR CS-Al 0, page V-3, Attachment V, of the LAR, implies that there are still some target cables that are identified at an area, or at best, compartment level, while some are located by raceway where applicable.
This would not appear to justify a wholesale reassignment of the Capability Category to 3.
Discuss the basis for this reassignment given there are still some targets located at an area level, which is Category 1.
RAI 5-29 RESPONSE:
The disposition has been expanded in a recent update to the ONS NFPA 805 Fire PRA Application Calculation. The disposition of fire affected basic events utilizes three unique disposition codes:
Code Disposition Y1 Link Basic Event to Appendix R equipment Y2 Link Basic Event to other equipment (assemble cable data)
Y3 Fail equipment in every Fire Area or set BE probability to 1.0 Cable location information for "Y3 components" is not available at the raceway level. However, since Y3 components, which are only credited by exclusion, are confirmed to be low risk contributors, the Capability Category 3 assignment may still be considered applicable for SR CS-Al 0. The justification provided in the Cable Selection Report for crediting Y3 components by exclusion, in conjunction with the cable location information (which extends to the raceway and endpoint level) provided in the Appendix R Database Management System for the remaining (Y1 and Y2) components credited in the Fire PRA is considered closer to Capability Category 3 than Capability Category 1. Furthermore, the impact of the use of Y3 components on the Fire PRA quantification results was evaluated in the sensitivity analysis in the ONS NFPA 805 Fire PRA Application Calculation.
RAI 5-30:
On pages X-5 through X-6, Attachment X, of the LAR, the following Sections B.3, C, D and E discuss only the following items: (1) the NRC "staff" review of March 2008, which occurred pre-transition, and not any peer reviews that may take place post-transition; (2) the FPRA as it currently exists and not how ONS will determine the FPRA's scope and level of detail as per the FPRA Standard for post-transition applications; (3) the FPRA as it currently exists and not how ONS will compare its FPRA model to the FPRA Standard for post-transition applications; 47
Enclosure 1: Request for Additional Information November 30, 2009 and (4) the FPRA as it currently uses supplementary analyses/requirements (none) and not how ONS will do so for post-transition applications.
The subject of Attachment X is post-transition. Discuss how ONS plans to address the four areas listed above post-transition, not just the current status of the FPRA.
RAI 5-30 RESPONSE:
The information provided in the response to RAI 1-2 (Duke submittal dated August 3, 2009) has been incorporated into the ONS NFPA 805 Fire PRA Application Calculation.
RAI 5-31:
With regard to sensitivity analysis, pages 23 and 26, Section 5, Attachment 1, of the ONS letter of, February 9, 2009, which states that "a CDF uncertainty band of plus or minus 10 percent (5.5E-05 to 4.5E-05) is more likely than the CDF range of 7.5E-05 to 2.5E-05 due solely to ignition frequency uncertainty of plus or minus 50 percent." With regard to sensitivity analysis, it is also stated that "based on the average impact of all the uncertainty parameters, the CDF uncertainty is judged to be within plus or minus 26 percent."
What is the actual uncertainty (not sensitivity) range from the internal events probabilistic risk assessment ( PRA)? Typically, this will be at least a factor of 2 in both directions (90 percent confidence range). Discuss why the uncertainty band estimated here for the FPRA is so much narrower. Is this actually an estimate of the maximum sensitivity effect, rather than a true uncertainty?
RAI 5-31 RESPONSE:
The Fire PRA uncertainty band referenced in the RAI is actually an estimate of sensitivity rather than true uncertainty. To address this, section 5.2, "Uncertainty Analysis" was added to the latest revision of Calculation, OSC-9518, Fire PRA Application. In this new section, the parametric uncertainty in the fire CDF result is compared to that of the internal events PRA in a recent update to the Calculation, OSC-9518, FPRA Application. The ratio of the 9 5 th to 5 th percentiles for the fire PRA model is -11. By comparison the ratio of the 95th to 5 th for the internal events model is -4. This result suggests that for the set of assumptions employed in this assessment, the overall parametric uncertainty for the fire CDF is higher than for the internal events model.
RAI 5-32:
With regard to dominant risk contributors, page 3, Section 4.3, OSC-9378, ONS FPRA Summary Report (NUREG/CR-6850 Task 16), May 2008, states that "at this time identification of significant contributors beyond ignition sources (scenarios) is of little value to the FPRA."
While it is recognized that the use of FRANC vs. an integrated model makes the identification of important risk contributors via the typical importance measures difficult, provide the basis for this claim "of little value."
48
Enclosure 1: Request for Additional Information November 30, 2009 RAI 5-32 RESPONSE:
The statement has been deleted from the calculation, OSC-9378, ONS FPRA Summary Report.
RAI 5-33:
With regard to sensitive electronics, on pages 24 through 25, Section 6.2, OSC-9375, calculation Oconee FPRA Scenario Development (NUREG/CR-6850 Tasks 8 arld 11), May 2008, ONS appears to have relaxed some of the assumptions from NUREG/CR-6850 concerning damage to solid-state components in cabinets within or near the zone of influence of the cabinet serving as the fire ignition source (e.g., cable-based rather than solid-state-component-based zones of influence; 10-min delay time before exposure in "adjacent cabinets").
Provide the results from sensitivity analyses performed to determine the effects of these relaxations on the FPRA.
RAI 5-33 RESPONSE:
No sensitivity analyses related to sensitive electronics were performed. The solid-state-component-based zone of influence, based on 3KWV/mA2, was intended for screening purposes. The ONS Fire PRA applies the concept that enclosures would provide protection to the sensitive internal contents from external fire effects along with guidance from Appendix S of NUREG /CR-6850 to achieve a 'beyond screening', more realistic treatment of sensitive electronics. According to Appendix S of NUREG/CR-6850, the primary concern with respect to impact on sensitive electronics is fire propagation to adjacent cabinets which house sensitive electronics. Cabinets are not considered 'adjacent' unless they are touching or nearly touching each other. As also described in Appendix S, if the adjacent cabinets are separated by a double wall with an air gap, then damage from the exposing cabinet to the exposed cabinet containing sensitive electronics is only assumed after 10 minutes of exposure. Therefore, damage need not be assumed in areas such as the Control Room where the fire would be expected to be extinguished quickly. For scenarios outside the Control Room, adjacent cabinets would fall within the zone of influence and were therefore, assumed to be impacted by the exposing cabinet.
RAI 5-34:
With regard to ventilated/open cabinets, pages 27 and 32 through 34, Section 8.4, calculation OSC-9375, Oconee FPRA Scenario Development (NUREG/CR-6850 Tasks 8 and 11), May 2008, cite "industry guidance" for assuming an electrical cabinet fire frequency apparently different from, and presumably lower than, that in NUREG/CR-6850. With regard to general transient severity factors, ONS derived these factors to "remove inherent conservatism and to better align with industry fire experience" (presumably in contrast to the current NUREG/CR-6850 approach), including a zero-failure rate approximation approach for the containment, control/auxiliary/reactor buildings, turbine building, and plant-wide to adjust transient fire frequencies It is unclear whether that adjusted frequency for an electrical cabinet fire was employed, other than as a source of data from which to develop a zero-failure rate approximation factor to adjust fire frequency in low energy cabinets. If that adjusted frequency was used directly in the FPRA calculations, provide the results from a sensitivity analysis performed to determine the effect of 49
Enclosure 1: Request for Additional Information November 30, 2009 its use in lieu of the current corresponding frequencies in NUREG/CR-6850 on the FPRA. Also provide the results from sensitivity analyses performed to determine the effect of using the general transient severity factors on the FPRA.
RAI 5-34 RESPONSE:
While fluctuations in severity factor magnitude were addressed in the sensitivity analysis, a sensitivity analysis specific to the low energy ventilated cabinets was not done; however, the results were known to be higher until this refinement was made. This refinement is considered necessary to obtain best estimate results given the reduced likelihood of a severe fire originating from a low energy cabinet and the conservatism associated with treating closed (but not well sealed or robustly secured) cabinets as open cabinets. The severity factor for ventilated electrical cabinets was derived from a review of industry events tabulated in the EPRI Fire Events Database and applied exclusively to the low energy ventilated cabinets in the Cable Room. While the scenario results improved, only one scenario, AB101 R, was particularly sensitive to this refinement. This scenario, which involved EHC Cabinets A, B, & C, realized a significant improvement in CDF as a result of the treatment. Application of the low energy severity factor allows for retention of the insights (significant target damage) without unrealistically skewing the results.
As discussed in the Fire Scenario Report, a review of the observed industry fire events found that none involved fire propagation beyond the boundary of an electrical cabinet of the type and voltage class that exists in the Cable Room. If one of these events had in fact been challenging, then the severity factor would increase by a factor of 2. Along those lines, sensitivity to the low energy cabinet treatment may be more properly based on an increase in observed events (i.e.,
from zero to one or two events). A collective 2 orders of magnitude improvement in CDF is observed even if the severity factor is doubled or tripled. The results are not particularly sensitive to the zero events approximation technique itself but more toward the concept that non-energetic cabinet fires are detected and suppressed before they become challenging.
Duke acknowledges that this approach cannot be combined with credit for suppression since it is not known for certain whether suppression played a role in reducing the fire severity of the applicable industry events in all cases.
Additionally, there has been considerable discussion relative to the low potential for fire propagation from closed (but not well sealed) cabinets. The application of the leakage fraction argument would eliminate the external target damage and would not be limited to low energy (low voltage) cabinets. A 'leaky' cabinet (one with a maximum leakage area of five percent, whether through vents, cable penetrations, or cracks and gaps) behaves similarly to a closed cabinet - no propagation. Cabinets with larger openings are considered open and assumed to propagate. While the vented surface area was not precisely determined, cabinets EHC A, B, &
C are potential candidates for the 'leaky' classification.
In light of the above, the application of this refinement allows for retention of fire risk insights while providing a more realistic best estimate of the fire risk due to low energy ventilated cabinet fires.
The RAI also requested the results from sensitivity analyses performed to determine the effect of using the general transient severity factors on the Fire PRA. The CDF results were not particularly sensitive to this treatment primarily due to the fact that most of the general transient scenarios were low risk contributors before application of the treatment. In some cases, the entire transient frequency, including the hotwork transient contribution, were included in the 50
Enclosure 1: Request for Additional Information November 30, 2009 scenario frequency. Application of the treatment was limited to logical transient placement (such as under low trays or next to risers), so the minimal sensitivity is not entirely unexpected.
Unlike hotwork transient scenarios involving pinchpoint locations, most general transient scenarios do not produce high CCDP values.
RAI 5-35:
With regard to hot work fire scenarios, on page 39 of Section 10.1, calculation OSC-9375, Oconee FPRA Scenario Development (NUREG/CR-6850 Tasks 8 and 11), May 2008, ONS appears to be reducing the non-suppression probability from the 0.38 value derived by crediting prompt suppression in NUREG/CR-6850 to a value of 0.01 based on several qualitative arguments, one of which is the use of "precautions consistent with Nuclear System Directive 314." There is an assumption that the hot-work-related fire events in the NUREG/CR-6850 database reflected hot work performed at much earlier dates when such "precautions" were not in place.
Discuss whether information that the hot-work-related fire events in the NUREG/CR-6850 database reflected hot work performed at much earlier dates when "precautions" were not in place is directly available from the event descriptions. Does ONS operating history indicate no fires or "close-calls" during hot work scenarios? Provide the basis, including the quantitative adjustment method, for what appears to be such a large reduction in non-suppression probability (approximately one and a half orders of magnitude). Recognize that the effect of "precautions" are typically included when assigning the weighting factors for maintenance, occupancy and storage to the transient ignition frequency, such that further crediting of the effects of these "precautions" in terms on non-suppression probability may be "double-counting."
Provide the results from a sensitivity analysis performed using the 0.38 non-suppression value to determine the effect on the FPRA.
RAI 5-35 RESPONSE:
While fluctuations in the magnitude of non suppression probabilities were addressed in the sensitivity analysis, a sensitivity analysis specific to the hot-work treatment was not done; however, the results were known to be higher until this refinement was made. This refinement is considered necessary to obtain best estimate results given the conservatism in the hot-work scenario definition. Hot-work scenarios are defined at pinch points without regard to access or likelihood of hot-work occurring at that location. Note that consideration of likelihood via the assignment of influence (weighting) factors is limited to the fire compartment as a whole - not to specific areas within a room. The 0.01 factor is not a direct replacement for the prompt suppression factor of 0.38 provided in Table P-3 of Appendix P for welding and cutting. Rather, the 0.01 factor includes consideration of failure of prompt suppression and failure to protect the target(s). Hot work is performed under precautions consistent with Nuclear System Directive 314 which include a fire watch and protection of nearby targets such as cable trays.
Consequently, damage to cable trays from welding and cutting would constitute a procedural non-compliance not accounted for in the suppression failure probability value. A review of the hot work related events in the EPRI Fire Events Database found that they either did not reflect an event that would have affected a plant cable or occurred more than 20 years ago. The industry has evolved significantly in the area of Fire Protection since the date of these early fire events. None of the recent ONS fire events reviewed for the Bayesian update period involved challenging hot-work fires. While the hot-work treatment resulted in an overall improvement of less than 30% in total fire CDF, the individual hot-work scenarios were not particularly sensitive to this refinement. In summary, application of the 0.01 factor as a modified non-suppression 51
Enclosure 1: Request for Additional Information November 30, 2009 probabilities value for hot-work scenarios allows for retention of the insights (target damage) without unrealistically skewing the results.
RAI 5-36:
With regard to the assumptions, page 6, Section 5.0, calculation OSC-9313, NFPA-805 Transition Non-Power FireArea Assessments (Pinch Point Analysis), October 30, 2008, states that "Failures of systems, equipment, instrumentation, controls, or power supplies, that are not a direct consequence of the fire, do not occur before, during, or following the fire."
Presumably, random failures of such items would be equally likely whether or not a fire scenario was in progress or immediately afterward, with the likelihood dependent upon the item's failure rate and the length of time over which it may be vulnerable. Provide the basis for this seemingly non-conservative assumption and discuss how an apparently more realistic assumption might affect the results of the NPO assessment.
RAI 5-36 RESPONSE:
Calculation OSC-9313, NFPA-805 Transition Non-Power Fire Area Assessments (Pinch Point Analysis), is being revised to clarify these assumptions. The assumption now is worded, "Failures of systems, equipment, instrumentation, controls, or power supplies, that are not a direct consequence of the fire, do not occur immediately before, during, or immediately following the fire." This clarification is meant to address the time-dependency concern identified above.
The original assumption is consistent with the at-power operations Nuclear Safety Capability Assessment methodology assumptions which do not require consideration of random failures or concurrent accidents/events other than those directly related to the fire. Since fire is not an event or accident as defined in the UFSAR, single failures do not have to be postulated for the Nuclear Safety Capability Assessment analyses. The primary purpose of the calculation is to identify plant locations where additional fire prevention measures need to be taken during higher risk plant operating states. The calculation assesses the potential damage footprint and then actions are recommended to preclude occurrence of a fire which could result in a loss of one or more Key Safety Functions. The assumption is used as a starting point of reference for system and component selection and to dictate the need for controlling out of service equipment during non-power operation. This assumption is considered conservative because it allows accurate assessment of potential damage footprints by ensuring all available equipment for Key Safety Function success paths are in service and because it narrows the time of consideration to the higher risk plant operating states as described in FAQ 07-0040.
RAI 5-37:
With regard to the assumptions, pages 6 through 7, Section 5.0, calculation OSC-9313, NFPA-805, Transition Non-Power Fire Area Assessments (Pinch Point Analysis), October 30, 2008, repeat the phrase "before, during, or following the fire."
While it is implicitly clear what time period is applicable "during the fire" (the time from fire ignition to final extinguishment), it is not clear what time periods apply "before" or "following" the fire. Presumably, these periods do not extend from the beginning to end of the NPO period.
Define the time periods meant by this phrase.
52
- Request for Additional Information November 30, 2009 RAI 5-37 RESPONSE:
Calculation OSC-9313, NFPA-805 Transition Non-Power Fire Area Assessments (Pinch Point Analysis), is being revised to clarify these assumptions. The phrase now is worded, "immediately before, during, or immediately following the fire." This clarification is meant to address the time-period concern identified above, but the actual timeframe remains indeterminate. In practical application, the assumptions are not entirely necessary since the primary purpose of the calculation is to identify plant locations where additional fire prevention measures need to be taken during higher risk plant operating states. The fire event itself is only used as a tool to assess the potential damage footprint and then actions are recommended to preclude occurrence of a fire which could result in a loss of one or more Key Safety Functions.
The assumptions are meant as a starting point of reference for system and component selection and to dictate the need for controlling out of service equipment during non-power operation.
The original assumptions are consistent with the at-power operations Nuclear Safety Capability Assessment methodology assumptions and allow accurate assessment of potential damage footprints by ensuring all available equipment for Key Safety Function success paths are in service and not being utilized to mitigate accidents or other events. The likelihood of the accidents or other events occurring is severely restricted by narrowing the time of consideration to the higher risk plant operating states as described in FAQ 07-0040.
RAI 5-38:
With regard to technical presentation, page 8, Section 6.0, calculation OSC-9313, NFPA-805, Transition Non-Power Fire Area Assessments (Pinch Point Analysis), October 30, 2008, states that ."Limited recommendations are made for [Category 1 fire] ... zones." Subsequently, Table 2, page 12 shows no recommendations for Category 1 fire zones.
Discuss this discrepancy, indicating which is correct. If it is the latter, provide the basis for applying no recommendations.
RAI 5-38 RESPONSE:
OSC-9313, NFPA-805 Transition Non-Power Fire Area Assessments (Pinch Point Analysis), is being revised. The wording was revised to indicate that Category 1 fire zones will have no recommendations made for them. The basis for no recommendations beyond normal fire prevention measures is that sufficient Key Safety Function success paths are available within those areas such that defense in depth can be maintained.
RAI 5-39:
With regard to technical presentation on page 12, Section 6.0, specifically, Table 2, and in , calculation OSC-9313, NFPA-805, Transition Non-Power Fire Area Assessments (Pinch Point Analysis), October 30, 2008, there appears to be no use of FAQ 40 Recommendation #6 or #8 for any fire zone Discuss why these are listed with ONS specific recommendations if they are not used.
53
Enclosure 1: Request for Additional Information November 30, 2009 RAI 5-39 RESPONSE:
Calculation, OSC-9313, NFPA-805 Transition Non-Power Fire Area Assessments (Pinch Point Analysis), is being revised to clarify the applicability of the FAQ recommendations to ONS. New tables of recommendations have been provided and the text revised to agree with the new recommendations. It is noted that implementation of the actions associated with the recommendations may not be fully developed until the transition/implementation period.
RAI 5-40:
With regard to the conclusion, page 4, Section 4.1, calculation OSC-9317, ONS NFPA 805, Transition Change Evaluation - Fire Area RB3, October 24, 2008, states that the three operator manual actions affiliated with open items RB3-07-O, -1 1-OE and 0 are not recovery actions.
Provide the basis for this conclusion. Discuss how, if they were treated as recovery actions, the results of the plant change evaluation would be different.
RAI 5-40 RESPONSE:
ONS is re-evaluating the variances from the deterministic requirements for all fire areas with respect to the equipment necessary to ensure safe and stable conditions in any plant mode.
These variances from the deterministic requirements are being characterized as either separation issues, previously documented Operator Manual Actions or as degraded Fire Protection systems and features. For those previously documented Operator Manual Actions, they are being screened to ensure that they do not take place in the Main Control Room or Primary Control Station before they are being called Recovery Actions. Once identified as a Recovery Action, they are further screened to determine if they are required to demonstrate the availability of a success path for the Nuclear Safety Performance Criteria. Only those actions required to demonstrate the availability of a success path for maintaining the plant in a safe and stable condition will be evaluated for additional risk in the Fire Risk Evaluations. Consequently, the change evaluations are being revised and titled as Fire Risk Evaluations in accordance with the nomenclature used in NFPA 805. The revised Fire Risk Evaluations will be written in accordance with the guidance of Draft Regulatory Guide 1.205, Revision 1 and are expected to be included with the LAR supplement.
RAI 5-41:
With respect to DID and safety margin, Table 6.2.2, page 4, Section 6.2.3, calculation OSC-9317, ONS NFPA 805, Transition Change Evaluation - Fire Area RB3, October 24, 2008, indicates that an ionization-type smoke detection system, portable fire extinguishers and hose stations and hydrants located in the area(s) are credited for the second element of DID.
With regard to manual suppression, discuss whether or not hose stations installed at each of the five levels of RB3 and, if not, why not. Discuss whether or not there would be timing concerns depending upon how long would it take for an extinguisher, maintained at the reactor building (RB) hatch, to be transported to the most distant potential fire location within RB3, including change in elevation if these are not maintained at each floor level. Provide the basis for these credits. Also, discuss why is there no "required" fire protection feature among the five listed for the second defense-in-depth element.
54
- Request for Additional Information November 30, 2009 RAI 5-41 RESPONSE:
ONS is re-evaluating the variances from the deterministic requirements (VFDRs) for all fire areas with respect to the equipment necessary to ensure safe and stable conditions in any plant mode. The variances from the deterministic requirements are being characterized as either separation issues, previously documented Operator Manual Actions or as degraded Fire Protection systems and features. Only those recovery actions required to demonstrate the availability of a success path for maintaining the plant in a safe and stable condition will be evaluated for additional risk in the Fire Risk Evaluations. Consequently, the previously submitted LAR "Change Evaluations" are being revised and titled as "Fire Risk Evaluations."
Initially, the change evaluations of the previous submittal did not credit Fire Protection Features for Defense in Depth considerations. The Fire Risk Evaluations will include consideration of those fire protection elements that are required for both "risk reduction" and Defense in Depth.
As part of the Defense in Depth consideration, the effectiveness of required manual firefighting capability (e.g. fire extinguishers, fire hose stations, etc.) will be evaluated in the Fire Risk Evaluations. Fire extinguishers and fire hose stations are generally located near specific hazards or spaced per the guidance of NFPA-10 and NFPA-14. Fire extinguishers are not left in the Reactor Building (containment) duringpower operation due to radiological and compressed gas cylinder concerns in a Post Loss Of Coolant Accident type environment.
Reactor Building fire mitigation and extinguishment timing will be evaluated in the final Defense in Depth assessment. The revised Fire Risk Evaluations will be included with the LAR supplement.
RAI 5-42:
With regard to fire area RB3, Table A.1-1, page 16, Section A.1, page 4, Section A.2.1, and pages 5 through 6, Section A.2.2, calculation OSC-9317, ONS NFPA 805, Transition Change Evaluation - Fire Area RB3, October 24, 2008, indicate the following: (1) essentially equal CDF and LERF values for Scenario D, while the LERF for Scenario E is over an order of magnitude lower than the corresponding CDF; (2) with regard to Open Item RB3-1 1-OE, that "3CCW-269 is only impacted in RB03 Scenarios B1 and D.".
While Scenario B1 includes a containment bypass scenario, and therefore the near equivalence of CDF and LERF might not be unexpected, both Scenarios D and E appear not to include LERF-relevant components, such that the "typical" LERF-to-CDF ratio of less than 0.1 that would be expected is evidenced in Scenario E but not D. Explain this potential discrepancy.
Also, address the potential effect of this discrepancy on the subsequent delta-CDF and delta-LERF evaluations for Open Items RB3-07-O and RB3-1 1-OE (Table A.2-2). Also, assuming that the statement regarding 3CCW-269 above is meant to exclude Scenario E, discuss how ONS has assured that the non-retained scenarios are not impacted by 3CCW-269.
RAI 5-42 RESPONSE:
Scenarios RB03 B1 and D involve failure of LERF-relevant components resulting in containment bypass. Therefore, there is no discrepancy. The near equivalence of CDF and LERF values for RB03 B1 and D is as expected; scenario RB03 E while involving failure of LERF-relevant components, requires a random failure in order to get a containment bypass, and therefore the LERF value remains significantly lower than CDF. Regarding 3CCW-269 specifically, it is only impacted in RB03 scenarios B1 & D. The CDF/LERF values for RB03 scenarios B1, D, & E are all above the screening threshold and were therefore retained for the delta risk calculation. Delta risk calculations are deemed unnecessary for non-retained 55
Enclosure 1: Request for Additional Information November 30, 2009 scenarios (i.e., those with CDF/LERF values below the screening threshold) given that the delta risk will be as low or lower than the base scenario value representing the variant or in-situ condition. In this case, 3CCW-269 was not impacted in scenario RB03 E or in any of the non-retained scenarios for RB03.
RAI 5-43:
With regard to Open Item RB3-33-O, page 7, Section A.2.3, calculation OSC-9317, ONS NFPA 805, Transition Change Evaluation - Fire Area RB3, October 24, 2008, indicates that the FPRA apparently does not model the risk impacts associated with potential failure of feedwater valve 3FDW-347. Therefore, there is no quantification of delta-CDF or delta-LERF, each of which is assigned a value of epsilon in Table A.2-3.
Typically, if a component is not modeled in a PRA, its potential contribution is considered to be bounded by the truncation limit, such that a delta calculation involving that component would be bounded by the truncation limit as well. Explain why the delta-CDF and delta-LERF are not assigned bounding values based on the truncation limit for the FPRA.
RAI 5-43 RESPONSE:
ONS is re-evaluating the variances from the deterministic requirements for all fire areas with respect to the equipment necessary to ensure safe and stable conditions in any plant mode.
The change evaluations are being revised and titled as Fire Risk Evaluations. The Fire Risk Evaluations will employ essentially the same methodology for the calculation of additional risk in terms of delta-CDF and delta-LERF; however, the scope of variances from the deterministic requirements will be different for this fire area. ONS has identified several areas where the use of epsilon may not be appropriate in the delta risk calculations. Use of the truncation limit as a bounding estimate for delta risk rather than epsilon may be appropriate in some cases such as when the unmodeled function is known to have negligible impact. But if the function is not modeled because it has zero impact, then the truncation limit is not applicable. In cases where the base scenario is below the screening threshold and no delta risk calculation is performed, use of "below the screening threshold" may be more appropriate in characterizing the delta risk for that scenario. Use of zero, epsilon, or the truncation limit may be considered a suitable estimate of the risk for variances that would result in no change in the calculated risk. The revised Fire Risk Evaluations will be included with the LAR supplement.
RAI 5-44:
With regard to methodology, page 1, Section 2.0, calculation OSC-9268, NFPA 805, Transition Non-Power Operations Component Selection, October 23, 2008, states that some plant operating states (POSs) were identified as inherently lower risk.
In identifying these as such, discuss whether or not consideration was given to the possibility of unique cable routing or placing of combustibles or ignition sources in atypical locations during these POSs before apparently dismissing them as candidates for high risk evolutions (HREs).
RAI 5-44 RESPONSE:
The Plant Operating States identified as inherently low risk, were identified as such in FAQ 07-0040 based on the numerous outage risk analysis documents referenced in the FAQ. No consideration of unique cable routing, combustible placement or atypical ignition source 56
Enclosure 1: Request for Additional Information November 30, 2009 locations was given in this evaluation, since the risk evaluation was based on the margins to core safety provided in the referenced documents for equipment in service during those Plant Operating States without regard to any plant specific or unique configurations. The Plant Operating States were used as a screening tool to provide the list of appropriate Key Safety Functions to be considered. From the required Key Safety Functions, component and cable selection was performed. Cable locations were then considered in determining potential pinch points where the Key Safety Functions could be impacted. The methodology utilized in calculation OSC 9313, NFPA 805 Transition Non-Power Fire Area Assessments (Pinch Points Analysis), also did not consider any unique target/fuel package/ ignition source interactions, since all potential fire scenarios within the zone were encompassed by the zone boundaries used to assess the potential damage footprint.
RAI 5-45:
With regard to methodology, page 6, Section 2.0, calculation OSC-9268, NFPA 805, Transition Non-Power OperationsComponent Selection, October 23, 2008, guideline (a) to determine whether or not "front-line" system components should be included for KSF support appears to limit the number of "electrically-supervised valves constituting system boundaries" to two. It is subsequently stated in guideline (c) that "valves in the flow path whose spurious operation could adversely affect system operation were included."
Provide the basis for the limitation in guideline (a). For guideline (c), discuss whether or not its inclusion was subject to the "two-valve" limitation of the guideline (a), and whether or not guideline (a) was limited only to "diversion" paths.
RAI 5-45 RESPONSE:
The concept for limiting the number to two was based on the need for providing some level of redundancy in Key Safety Function success paths to potentially reduce the number of pinch points identified. Calculation OSC-9268, NFPA 805 Transition Non-Power Operations Component Selection, states that front line components for system boundaries would be taken out to two electrically supervised valves. In many cases only a single valve was available for system boundaries for a given Key Safety Function success path. Potential flow diversion path valves subject to spurious operation and system boundary valves are essentially the same.
Potential flow blockage valves subject to spurious operation were included within the component selection for the main flow path such that the Key Safety Function success path could be jeopardized if fire damage occurs. Only a single valve could potentially cut off flow of a given Key Safety Function success path. The result is that a comprehensive list of valves needed to support the Key Safety Function success paths was developed including potential spurious operations that could prevent achievement of the Key Safety Function.
RAI 5-46:
With respect to DID and safety margin, Table 3.2-2, pages 4 through 5, Section 3.2.3, Enclosure
'2, calculation OSC-9321, ONS NFPA 805, Transition Change Evaluation - Fire Area BOP, October 28, 2008, indicates that an ionization-type smoke detection system, portable fire extinguishers and hose stations and hydrants located in the area(s) are credited for the second element of DID.
Discuss why is there no "required" fire protection feature among the five listed for the second DID element.
57
- Request for Additional Information November 30, 2009 RAI 5-46 RESPONSE:
ONS is re-evaluating the variances from the deterministic requirements for all fire areas with respect to the equipment necessary to ensure safe and stable conditions in any plant mode.
The variances from the deterministic requirements are being characterized as either separation issues, previously documented Operator Manual Actions or as degraded Fire Protection systems and features. Only those recovery actions required to demonstrate the availability of a success path for maintaining the plant in a safe and stable condition will be evaluated for additional risk in the Fire Risk Evaluations. Consequently, the change evaluations are being revised and titled as Fire Risk Evaluations. Initially, the change evaluations of the previous submittal did not credit Fire Protection Features for Defense in Depth considerations. The Fire Risk Evaluations will include consideration of those fire protection elements that are required for both "risk reduction" and Defense in Depth. As delineated in the response to RAI 2-8, those Fire Protection Systems and Features defined as satisfying the needs of Defense in Depth and safety margin will become "required" systems and be within scope of the Monitoring program.
The revised Fire Risk Evaluations will be included with the LAR supplement.
RAI 5-47:
With regard to the delta-CDF and delta-LERF calculations in Table A.2-2 for open item BOP OP, pages 4 through 5, Section A.2.2, Enclosure 2, calculation OSC-9321, ONS NFPA 805, Transition Change Evaluation - FireArea BOP, October 28, 2008, state that "consideration of uncertainty (plus or minus 26 percent) will not change this conclusion that the values are not above the acceptance limits." A similar statement appears for Open Item BOP-69-OP in Section A.2.3.
Discuss how this conclusion might be different with respect to RAI-1 1. (See also RAI-1 1, to which this is related.)
RAI 5-47 RESPONSE:
As discussed in response to RAI 5-31, uncertainty and sensitivity have been revisited. The Fire PRA results represent our best estimate and as such do not require quantitative consideration of uncertainty. Accordingly, the delta risk calculations do not require adjustment for uncertainty.
However, specific sensitivities may need to be considered. For a calculated delta risk value that is above the acceptance limit for self-approval or a decade below the acceptance limit for self-approval, no further analysis is required. Otherwise, the sensitivity analysis should be considered to determine if the calculated delta risk could be significantly impacted as to alter the conclusion. These insights can be used to identify compensatory action or Defense in Depth measures that can help ensure that the base case is truly representative of the best estimate case.
ONS is re-evaluating the variances from the deterministic requirements for all fire areas. The change evaluations are being revised and titled as Fire Risk Evaluations. The revised Fire Risk Evaluations will be included with the LAR supplement.
RAI 5-48:
Discuss whether or not the presence of concealed cables, routed through the control room which were not considered originally for compliance with 10 CFR 50.48 have been addressed in 58
Enclosure 1: Request for Additional Information November 30, 2009 the FPRA. If not, provide justification for the exclusion. If addressed, discuss the implications on the FPRA-related results.
RAI 5-48 RESPONSE:
The Fire PRA relies on cable routing information provided in the Appendix R Database Management System. No cables credited in the safe shutdown analysis were included in the scope of "concealed" cables in the Control Room based on the results from the cable routing effort performed for the Appendix R reconstitution effort and documented in the Appendix R Database Management System. Low voltage power cable 3KRA1 5 from Static Inverter 3KOAC to panel board 3KRA, is not a credited safe shutdown cable nor was either 3KOAC or 3KRA identified for inclusion in the Appendix R Database Management System based on Fire PRA input (applicable to Y2 components). 3KRA, located in the Unit 3 Cable Room (AB1 01) is credited in the Fire PRA as a Y3 component (credited by exclusion only); 3KRA provides an alternate power supply to vital instrumentation power panels (3KVIA, 3KVIB, 3KVIC, & 3KVID).
3KOAC is not credited in the Fire PRA. There are no scenarios in either the Unit 1 & 2 Control Room or the Unit 3 Control Room where cables above the false ceiling would be impacted.
Therefore, cable 3KRA15, routed above the false ceiling in the Unit 3 Control Room, would not be impacted in any Unit 3 Control Room scenario. 3KRA is failed in Unit 3 Cable Room scenarios since its routing is unknown within the Fire RPA in that location. Accordingly, the "concealed" cables in the Control Room have no impact on the Fire PRA results.
RAI 5-49:
The ONS PRA model does not appear to contain a failure-to-start basic event for the A high pressure injection (HPI) pump, i.e., the model assumes this pump is always running. While it is true that an HPI pump (combined charging pump) is usually running, this condition may not be true for a LOSP initiator. Discuss this apparent modeling idiosyncrasy and any effect it may have on the FPRA results, especially when a fire-induced LOSP is triggered.
RAI 5-49 RESPONSE:
The 'A' High Pressure Injection pump is presumed to be always running in the PRA ('A' or 'B' is normally operating); only B & C have failure to start and spurious start basic events. The 'A' pump breaker failing to close does not contribute to a pump failure to restart. The 'A' pump has a failure to run basic event which is used to address fire induced failure resulting from cable damage. Therefore, this asymmetry has minimal impact on the Fire PRA results. The fault tree addresses power supply dependencies; the High Pressure Injection pumps are not credited for fire scenarios impacting loss of their power supply. Note that at ONS, unlike most plants with Diesel Generators, the High Pressure Injection pump breaker is not tripped on loss of offsite power as part of a load shed prior to loading the emergency power source. Keowee has the capacity to start up with these loads connected. Additionally, the Loss Of Offsite Power initiator
(%T5) is not applied for any Fire PRA scenarios; credit for Loss Of Offsite Power recoveries are not taken in the Fire PRA.
RAI 5-50:
Page 35, Section 4.5.1.1, in the LAR, states that "In addition, 24 of the SRs are not applicable to the ONS PRA, either because the referenced techniques are not utilized in the PRA or because the SR is not required for capability category (CC) I1."
59
Enclosure 1: Request for Additional Information November 30, 2009 The reference to CC II seems to imply this constitutes some level of acceptability for the SRs.
Is this the intent of the statement? If so, provide the basis for assuming CC II constitutes an acceptable level for SRs in general.
RAI 5-50 RESPONSE:
Regulatory Guide 1.205 (Draft Rev. 1, September 2009, ML092450118) indicates that "For PRA Standard "supporting requirements" important to the NFPA 805 risk assessments, the NRC position is that Capability Category II is generally acceptable." There are cases, depending on the plant, the analysis approach taken with respect to subsequent related supporting requirements, and the specific fire risk evaluation, where Capability Category I is sufficient and perhaps cases where Capability Category III is required. Refer to responses to RAIs 5-4 and 5-8.
RAI 5-51:
Page 36, Section 4.5.1.2, in the LAR, states that "For the limited number of cases where the Unit 2 results were not considered to be bounding, a method for adjustment of the Unit 2 results for application to Unit 1 was provided. This is documented in a licensee calculation entitled "Unit 1 Comparative Screening Analysis."
Provide a summary of the results from this calculation.
RAI 5-51 RESPONSE:
The comparative screening analysis indicated that a separate Unit 1 fault tree and Fire PRA quantification file are not necessary. The analysis involved a failure comparison and a frequency comparison. The comparative screening analysis concluded that the fire CDF/LERF results for the comparable Unit 2 compartment may be used for determining the Unit 1 fire CDF/LERF in all but 5 cases. In these cases, an adjustment of the Unit 2 model must be made to determine the Unit 1 result.
- Unit 1 Surrogate Compartment Unit 2 Adjustment Compartment AB085 ABO81 Calculate CCDP/CLERP increase assuming failure of CC Pumps 2A & 2B and LPI pumps 2A, 2B, & 2C.
5AB092 -Calculate CCDP/CLERP increase assuming failure of CT4 and HPSW Pumps A & B.
UlTB23 U2TB14 Calculate CCDP/CLERP increase assuming failure of HPSW Pumps A & B and LPSW Pump B 60
- Request for Additional Information November 30, 2009 Unit 1 Surrogate Compartment Unit 2 Adjustment Compartment U1TB34 U2TB33 Calculate
[5 CCDP/CLERP increase assuming failure of upB LPSW Pump B.
U 1TB36 U2TB30
& & Increase compartment frequency by 38%.
U 1TB44 U2TB31 RAI 5-52:
Reference Table S-1, pages S-3 through S-9, Attachment S, to the LAR.
Describe how changes/modifications planned and to which committed will be addressed in the FPRA. Also, are the listed compensatory measures included in whatever version of the FPRA supports this LAR for transition to NFPA 805 and, if so, how?
RAI 5-52 RESPONSE:
Currently, only the Protected Service Water modification will be credited in the Fire PRA when generating risk results for use in the LAR supplement. The compensatory measures listed in table S-1 are not modeled in the Fire PRA.
RAI 5-53:
Page U-3, Attachment U, to the LAR, lists SR-AS-B3 as a documentation issue.
Discuss how the expected impact considers the possible effects of non-fire environmental effects that may be triggered by fire inducing an internal events initiator such that these non-fire environmental effects could affect the response to the fire itself (e.g., access/egress for fire fighting or local operator actions, spurious signals due to steam).
RAI 5-53 RESPONSE:
Attachment U provided an assessment of the impact of the open supporting requirements for ONS on the risk-informed fire protection PRA application. SR AS-B3, which dealt with identification of the phenomenological conditions created by each accident sequence, was dispositioned as a documentation issue. The examples cited in the assessment comment are not expected to have a material impact on the PRA results for the following reasons:
LOCA inside containment with failure of the Reactor Building Cooling system Loss of all Reactor Building Cooling Units is a low probability event. Steam Generator level instrumentation is not required for all sequences leading to the harsh environment in containment. The combination of circumstances that lead to the harsh environment in containment and also would require the operators to have Steam Generator level 61
- Request for Additional Information November 30, 2009 instrumentation in order to prevent core damage is expected to be an insignificant contributor to the PRA results.
Steam line breaks in the Turbine Buildinq It is possible that other equipment could be affected by a steam line break in the Turbine Building; however, it would not be expected to contribute to CDF. If all 4 KV power is lost and the turbine-driven emergency feedwater pump is lost, the station would rely on the Standby Shutdown Facility.
Clogging of the Reactor Building Emergency Sump Historically, all of the available design information has suggested that the probability.of sump blockage was expected to be insignificant. It was neglected on that basis. In recent years this conclusion has been questioned but at the same time led to a requirement to modify sumps to again reduce the probability of clogging to an acceptably low level. Other failure modes that have the equivalent consequences to clogging exist in the PRA model.
Unless clogging has a probability that is significant when compared to these other failure modes its omission has no material impact on the PRA results.
Accordingly, Attachment U concluded that the phenomenological impacts are already considered in the model; therefore, the open supporting requirement has no impact on the NFPA 805 application. Some of the aforementioned phenomenological conditions, such as those arising from a steam line break, are not capable of being fire induced. Other non-fire environmental effects triggered by fire induced failures are subsumed by the effects of the fire.
For example, operator actions taken outside the control room are failed in the areas affected by the fire, including consideration of operator access.
RAI 5-54:
For SR HR-G6, page U-3, Attachment U, to the LAR, states that "No impact is expected for documentation issues."
Discuss how the expected impact considers the effect of fires for the cues that alert operators, relevant performance shaping factors (PSFs), and availability of staff.
RAI 5-54 RESPONSE:
This was addressed during Simulator Review as documented in calculation, OSC-8978; most actions relied on redundant and diverse instrumentation. A few actions which rely on a more limited set of instruments for cues and/or execution were linked directly to instrument failure.
RAI 5-55:
For SRs LE-C10 and LE-G5, pages U-23 and U-26, Attachment U, to the LAR states that "No impact is expected for documentation issues."
Discuss how the expected impact for SR LE-C10 considers that fire effects could shift the dominance between the bypass and non-bypass events, such that the crediting for scrubbing on the bypass events could reduce their dominance. Discuss how the expected impact for SR LE-G-5 considers that fire effects could invalidate some of the limitations from the internal events LERF analysis, e.g., if the internal events analysis dismissed containment bypass due to 62
Enclosure 1: Request for Additional Information November 30, 2009 spurious opening of penetrations being very unlikely in non-fire scenarios that would be more likely given fire.
RAI 5-55 RESPONSE:
With respect to SR LE-C1 0, the LERF analysis does not take credit for scrubbing. Credit for scrubbing would reduce the LERF contribution from bypass events. However, the conservative treatment does not mask the contribution of non-bypass events, because even if some credit were given to scrubbing, the unscrubbed bypasses would still dominate LERF over the non-bypass events. Thus no adjustments are needed to support the Fire PRA.
With respect to SR LE-G5, previously screened containment penetrations and Interfacing System Loss of Coolant Accident pathways were reviewed against updated fire criteria during Task 2 (Component Selection) of the Fire PRA development. This is related to RAI 5-56.
RAI 5-56:
For SR SY-A14, page U-35, Attachment U, to the LAR, states that "No impact is expected for documentation issues."
Discuss how the expected impact considers that failures screened out for internal events PRA due to low probabilities may have higher probabilities given fire, such that they would need to be included in the FPRA? Discuss how the review of failure modes includes reviewing those previously screened out to ensure such screening remains valid for the FPRA.
RAI 5-56 RESPONSE:
In addition to Multiple Spurious Operations reviews, flow diversion paths were identified during reconstitution; these previously "low probability" component failures (e.g., normally closed motor operated valves transferring open) were reviewed and included, as necessary, during Fire PRA development (Task 2 Fire PRA Component Selection). Fire PRA also revisited previously screened Interfacing System Loss of Coolant Accident and containment penetration pathways to assess the need for inclusion in the Fire PRA.
RAI 5-57:
On page V-2, Attachment V, to the LAR, the response for SR CS-B1-1 states that "Breaker coordination impacts are not a contributing factor for top risk contributing scenarios which involve loss of 4KV power or CR abandonment."
Discuss the basis for this conclusion, including whether or not the breaker coordination impacts could have an effect on other risk-contributing scenarios such that they could become important contributors.
RAI 5-57 RESPONSE:
The top 50% risk contributing scenarios involve loss of 4KV power and reliance on Standby Shutdown Facility mitigation which are not significantly impacted by additional failures due to improper breaker coordination. If the breaker coordination study concludes that additional failures need to be considered for certain loads impacted by fire, then the risk of less significant scenarios would increase. However, per the draft breaker coordination study, the impacts are 63
- Request for Additional Information November 30, 2009 not expected to be significant since coordination at the higher voltage levels has been confirmed. Refer to RAI 3-3 response for the schedule of completion of the breaker coordination study. The Fire PRA will be updated after the results of the breaker coordination study become available.
RAI 5-58:
As indicated in the response to SR FQ-CI-1, on page V-2, Attachment V, to the LAR, the maximum human error probability for initial solve is limited to 0.1, Current NUREG/CR-6850 screening guidance for Human Error Probabilities given fire conditions includes setting values at 1.0, and the proposed update to this approach by the pilot plants includes screening as high as 1.0, as does the proposed RES/EPRI Fire Human Reliability Analysis methodology. Provide the basis for limiting the maximum human error probability for initial solve to 0.1, as indicated in the response to SR FQ-Cl-1. Discuss the possibility that some human errors may have been screened out prematurely.
RAI 5-58 RESPONSE:
The ONS PRAs (both the fire and internal events) generate cutsets with human error probabilities set to 0.1 or the nominal value, whichever is greater. The objective in elevating the human error probabilities during cutset generation is to provide confidence that important combinations of operator actions are not truncated because of small nominal values. Multiple operator actions occur frequently in ONS cutsets. Using 0.1 provided hundreds of joint human error probabilities to evaluate. Operator actions for all of the risk significant systems (Emergency Feedwater, High Pressure Injection, Standby Shutdown Facility, Low Pressure Injection, and Protected Service Water) all occur in many combinations. It is judged that using 1.0 as a quantification value would not introduce new important combinations but would increase the original cutset file size and consequently the time required for quantification.
RAI 5-59:
On pages W-2 through W-4, Attachment W, of the LAR, the collective fire CDF, fire CDF credit for non-fire-related modifications, total fire CDF decrease associated with transition, and total baseline fire CDF have been reported; the corresponding values for fire LERF have not. Also, results are reported for the top 50 percent of the fire CDF contributors; those for the corresponding contributors to fire LERF are not.
Provide the collective fire LERF, fire LERDF credit for non-fire-related modifications, total fire LERF decrease associated with transition, and total baseline fire LERF. Also provide the results for the top 50% of the fire LERF contributors, (NOTE: The February 9, 2009, response letter to the acceptance review provides the values for the top LERF contributors, but their corresponding fractional contributions to the LERF are not provided, as they were for the top CDF contributors.)
RAI 5-59 RESPONSE:
Calculation OSC-9518, NFPA 805 FPRA Application Calculation, forms the basis for Attachment W. While LERF was addressed in the calculation, few additional insights were derived from the LERF results. The calculation has been updated to address a greater percentage of the risk contributing scenarios consistent with the definition of "significant" in the 64
Enclosure 1: Request for Additional Information November 30, 2009 ASME Standard. Following the update to include the additional scenarios, only one of the risk significant scenarios for CDF involved containment bypass and none were within the top 50% of CDF contributors, therefore, most of the insights from the CDF results are directly applicable to the LERF scenarios.
RAI 5-60:
With regard to the delta risk calculation, page 5, Section 2, Attachment 1, to the ONS, February 9, 2009, letter, states that "Baseline risk is generally associated with a compliant (vs. a non-compliant) configuration. Accordingly, another way to characterize the change in risk is the difference between the compliant case and the non-compliance. In most but not all cases, the non-compliant or variant case represents the in-situ (base case) condition." This is then followed by a table indicating that the delta-risk is always calculated as Case 2 minus Case 1, where Case 2 is always the variant and Case 1 the compliant, with one or the other being the base.
This seems to imply there are only two situations: (1) current configuration is compliant yet a variation is being proposed and (2) current case is non-compliant, and the non-compliant variation is being proposed for NFPA 805 and compared against what would have been compliant. Discuss whether or not there is a third case, where the current condition is non-compliant and not being proposed for NFPA 805, but rather it is a new variant that is being proposed, and this will be measured against what would have been compliant. There may be, in effect, a third example, where neither Case 1 nor Case 2 is base.
RAI 5-60 RESPONSE:
While the delta risk technique in calculation OSC-9518 may be used to determine the net risk impact to the Fire PRA, if it is not a variance from the deterministic requirements, then no fire risk evaluation (referred to as change evaluation in the previous LAR submittal) would be initiated. Changes in compliance strategies (relying on one operator manual action versus a new one) are not considered "new" variance from the deterministic requirements. The variance from the deterministic requirements is the non-compliant condition that resulted in the need for an operator manual action.
RAI 5-61:
With regard to the delta risk calculation, page 6, Section 2, in the ONS, February 9, 2009, letter, states that "most of the operator actions are not explicitly modeled in the FPRA."
Explain the apparent exclusion of "most of the operator actions." Even if they would typically be beneficial, is it possible that they could make things worse if performed incorrectly, out of order, etc.? Provide the basis of this seemingly a priori dismissal from the FPRA.
RAI 5-61 RESPONSE:
A detrimental risk impact review was performed, but the Appendix R operator manual actions tended to recover lost equipment functions which would have had minimal risk benefit, if restoration of the function were modeled in the Fire PRA. Detrimental risk impacts will be revisited after finalization of the updated operator manual action list as part of the revised LAR supplement.
65
- Request for Additional Information November 30, 2009 RAI 5-62:
With regard to LERF considerations, page 14, Section 3.2, Attachment 1, to the ONS February 9, 2009, letter, states that "in the case of HP-21, there is the potential to lose power resulting in the failure of the active function to close even for a fire in containment. Therefore, neither of these paths represent[s] an MSO scenario (one is not real and the other is failure of the active function to close)."
Discuss why, if on the seal return line motor operated valve HP-20 fails open (i.e., is de-energized so it cannot close when it should) and air operated valve HP-21 fails to close (when it should do so) due to loss of power, this not an MSO scenario. Would there not be two valves failing to close when needed due to spurious de-energization (i.e., failure to energize so as to be able to close)?
RAI 5-62 RESPONSE:
The scenario referenced in the RAI represents a loss of power supply to a Motor Operated Valve, which would prevent the ability of the valve to close on demand. Demand failures are not characterized as Multiple Spurious Operations scenarios, although they are evaluated for impact as part of the integrated Fire PRA.
RAI 5-63:
With regard to the PSW modification, pages 9 through 12, Section 3.4, in OSC-9377, ONS FPRA Model Development (NUREG/CR-6850 Task 5), May 2008,_state that "the planned PSW modification was credited in the FPRA on a limited basis."
In the FPRA model, six basic events have been introduced to address failures related to the PSW system: (1) Three new human failures, UPSWSHRDHE, BSFAPWRDHE and BHPOASWDHE, for human action failure to effect the PSW function for ASW, SSF and HPI, respectively, with final failures probabilities of 0.5, 0.15 and 0.059, respectively; (2) PSWMOD for electrical hardware failures when the PSW function is supporting HPI or SSF (potentially a surrogate for failures that will be associated with the PSW supply and support equipment that will be installed in the PSW Blockhouse and between the Blockhouse and destination, with an aggregate failure probability of 0.00151; (3) Two 'house-like' events, PSWHPI3B and PSWVOTSI, which assume values of zero or one based on the location of the fire to represent whether or not the PSW function is being credited (for a fire in the Auxiliary Building, PSWHPI3B is set = 1, otherwise it has a value of zero; for a fire in the SSF, PSWOTS1 is set
= 1, otherwise it has a value of zero). All but the basic events BSFAPWRDHE and BHPOASWDHE are discussed. Discuss these two basic events as well. Is the characterization of the function of all six events within the FPRA correct? Discuss any other model effects to represent the impact of the PSW modification in the FPRA (e.g., changes in fire ignition frequencies or combustible loadings).
RAI 5-63 RESPONSE:
The characterizations of the functions are correct. The two basic events have been added to the Protected Service Water discussion in an update to the Fire PRA Model Development report. No credit relative to Protected Service Water is taken for operator action BHPOASWDHE to align Protected Service Water power to an High Pressure Injection pump when surrogate event PSWHPI3B is failed; no credit relative to Protected Service Water is 66
- Request for Additional Information November 30, 2009 taken for operator action BSFAPWRDHE to align Protected Service Water power to the Standby Shutdown Facility switchgear when surrogate event PSWOTS1 is failed. Protected Service Water is not expected to have a significant impact on combustible loading since armored cables or their functional equivalent will be used for power and control. The impact on fire ignition frequency will be minimal overall with most of the significant ignition sources to be housed in the Protected Service Water Blockhouse. As discussed in the Fire PRA Model Development report, the Fire PRA does not currently have any scenarios for the Protected Service Water Blockhouse itself.
RAI 5-64:
With regard to control room abandonment scenarios, Scenarios E and F on page 15 of Section 3.2.2 in OSC-9375, Oconee FPRA Scenario Development (NUREG/CR-6850 Tasks 8 and 11),
May 2008, both cite minimum abandonment times (8.5 min and 12.4 min, respectively) less than the threshold assumed for the Control Room (20 min). Severity factors are then estimated based on assuming that 20 min are available for suppression prior to abandonment.
Assuming these factors would be higher if only the minimum abandonment times were assumed available for suppression, provide the basis for crediting 20 min for suppression rather than the cited minimum abandonment times. Also, based on the Regulatory Audit review of "Summary of Control Room Abandonment Times at the Oconee Nuclear Power Station Unit 3" by Hughes Associates, Inc., October 8, 2007, discuss why the abandonment times were chosen based on the assumption that both the ventilation supply and smoke purge fan were on, rather than one or both were off, which would have yielded shorter abandonment times.
RAI 5-64 RESPONSE:
Severity factors for the Control Room abandonment scenarios are not based on assuming that 20 minutes are available for suppression prior to abandonment. Rather, ifthe abandonment times are calculated to be 20 minutes or greater, then abandonment is assumed to not be required since 20 minutes provides ample time for suppression. Given that the Control Room is continuously occupied, the 20 minute time cutoff is not expected to be contentious; moreover, the results would not appreciably change if the calculated non suppression probability values considered time frames greater than 20 minutes. The Control Room abandonment discussion was updated to address an open item resulting from an internal review. Abandonment times from cases E & F were overly conservative and identified as being not applicable to the observed configuration. When using the appropriate cases (B & D) with reduced credit for the purge fan (accounting for non-fire related unavailability) and application of the probability distribution, the revised abandonment parameters (combination of non suppression and severity factor) were very comparable to the Revision 0 parameters. The best estimate of the risk for the abandonment scenarios is not determined by assuming failure of Heating, Ventilation, and Air Conditioning or other available systems. As indicated by the abandonment results for the Unit 1
& 2 Control Room, which has no smoke purge fan, the final results do not vary significantly with or without Heating, Ventilation, and Air Conditioning.
RAI 5-65:
In OSC-9375, Oconee FPRA Scenario Development (NUREG/CR-6850 Tasks 8 and 11), May 2008, these is an apparent omission.
67
Enclosure 1: Request for Additional Information November 30, 2009 It appears that the analysis for MCR abandonment addresses only degradation/loss of functions due to fire effects within the MCR. If so, discuss how degradation/loss of functions in the MCR due to fire outside the MCR (even if there are no resulting environmental fire effects within the MCR) has been addressed, including abandonment of the MCR under these condition. While it appears that a pre-existing Oconee assumption that no spurious actuation occurs within 10 minutes of a fire has not been carried in the FPRA (and this is the correct approach), it may have implicitly been retained in the MCR abandonment study due to its continued retention in the plant operating procedures planned for NFPA-805 transition. Discuss whether this is the case and, if so, how the MCR abandonment analysis for the FPRA compensates for this. If enough spurious actuations occurred within the first 10 minutes to cause degradation/loss of functions within the MCR, would non-procedural MCR abandonment be considered?
RAI 5-65 RESPONSE:
Main Control Room abandonment is only initiated for environmental considerations even if considerable fire induced failures render loss of Main Control Room functionality. Main Control Room abandonment cases intended to rely only on Standby Shutdown Facility mitigation; use of specific initiator to accomplish this objective was replaced by Control Room equipment failures following NRC staff review of Fire PRA. The assumption that no spurious actuations occur within 10 minutes of a fire has not been included in the Fire PRA.
RAI 5-66:
With regard to technical presentation, page 7 of Section 6.0 in OSC-9313, NFPA-805 Transition Non-Power Fire Area Assessments (Pinch Point Analysis), October 30, 2008, states that "[a]n
'Impact' upon a KSF success path is defined as a component or any of its associated cables being located within the fire zone such that, following a fire which conservatively assumes the total loss of all compartment contents, the component can no longer be assured of being functional ... In some cases where ... a redundant pump, valve, or instrument was ... not affected by the fire, the KSF success path was shown as not impacted."
This appears to imply that "impact" means only a complete loss of a KSF, not a decrease in redundant pathways. Verify whether or not this is correct and, if so, discuss how the distinction affects the NPO assessment (e.g., as per Section 1.3.1, "Nuclear Safety Goal," of NFPA 805,
"[t]he nuclear safety goal shall be to provide reasonable assurance that a fire during any operational mode and plant configuration will not prevent the plant from achieving and maintaining the fuel in a safe and stable condition").
RAI 5-66 RESPONSE:
Calculation, OSC-9313, NFPA-805 Transition Non-Power Fire Area Assessments (Pinch Point Analysis), is being revised. The calculation describes an "impact" as a loss of a Key Safety Function success path. In some cases, not all permutations of the Key Safety Function success paths were specifically identified due to the numerous support component options available from the basic redundancy and diversity design philosophy of nuclear plant construction. In these cases, some support components from a different Key Safety Function success path not specifically tied to the particular Key Safety Function success path under consideration would be available and could be credited to ensure that the Key Safety Function success path under consideration would still be available as well and not "impacted." Redundant pathways would then be preserved thus fulfilling the nuclear safety goal. Fire zones where all Key Safety Function success paths would be unavailable for whatever reason (common components or 68
Enclosure 1: Request for Additional Information November 30, 2009 cable routing) are identified as a "pinch point" and recommendations are made to preclude the occurrence of a fire in those areas as well as to take pre-emptive actions to mitigate the potential impact of a fire on the Key Safety Function. The revision of calculation, OSC-9313, clarifies the wording to reflect this.
RAI 5-67:
With regard to acceptance criteria assessment, defense-in-depth, page 11 of Section 2.5 in OSC-9314, NFPA-805 Transition Risk-Informed, Performance-BasedChange Evaluation Methodology, October 28, 2008, states that a scenario that does not lead to core damage, but has a CDF of 9E-08/yr would be treated differently from one that leads to core damage with the same frequency.
Discuss how a scenario that does not lead to core damage can have a non-zero CDF.
RAI 5-67 RESPONSE:
The previous Change Evaluation methodology calculation (OSC-9314) used an incorrect example in the defense in depth acceptance criteria discussion section. The cited example will be deleted from the Defense in Depth discussion section in the new revision of the calculation which is being developed in response to Regulatory Guide 1.205 draft revision 1 for performing Fire Risk Evaluations. The revised Fire Risk Evaluation methodology will be discussed in the LAR supplement.
For Open Items BOP-12-OP, BOP-28-OP, BOP-46-O and BOP-70-O, on pages 1 through 2 of Section 1.1 in OSC-9321, Oconee NFPA 805 Transition Change Evaluation- Fire Area BOP, October 28, 2008, the VFDs are characterized as involving an "unallowed" operator manual action. For Open Item BOP-69-O, the VFD is characterized as involving a manual action that is "not feasible." Subsequently, in the disposition of each Open Item in Section 4.2, Table 4.2-1 (page 5), all these Open Items are characterized as involving "no feasible recovery actions."
Why is there no "required" fire protection feature among the five listed for the second defense-in-depth element? (Note that this same question applies to Table 3.2-2 in Section 3.2.3 of Enc.
3 and Table 3.2-2 in Section 3.2.3 of Enc. 4.)
RAI 5-68 RESPONSE:
The Balance Of Plant Fire Area will be separated into two separate fire areas, Turbine Building and Auxiliary Building. Using the methods described in Draft Regulatory Guide 1.205, these two new fire areas will be re-analyzed and the requirements for NFPA 805 re-established. ONS is re-evaluating the variances from the deterministic requirements for all fire areas with respect to the equipment necessary to ensure safe and stable conditions in any plant mode. The variances from the deterministic requirements are being characterized as either separation issues, previously documented Operator Manual Actions or as degraded Fire Protection systems and features. For those previously documented Operator Manual Actions, they are being screened to ensure that they do not take place in the Main Control Room or Primary Control Station before they are being called Recovery Actions. Once identified as a Recovery Action, they are being further screened to determine if they are required to demonstrate the availability of a success path for the Nuclear Safety Performance Criteria. Only those actions 69
- Request for Additional Information November 30, 2009 required to demonstrate the availability of a success path for maintaining the plant in a safe and stable condition will be evaluated for additional risk in the Fire Risk Evaluations. Consequently, the change evaluations are being revised and titled as Fire Risk Evaluations. The Fire Risk Evaluations reconsider Defense in Depth and Safety Margins. As part of that consideration, Fire Protection Systems and Features will be identified to satisfy the needs of Defense in Depth and Safety Margin. As delineated in the response to RAI 2-8 those Fire Protection Systems and Features defined as satisfying the needs of defense in depth and safety margin will become "required" systems and be within scope of the Monitoring program. The revised Fire Risk Evaluations will be included with the LAR supplement.
Appendix C of NEI 04-02 states, "The existing FP Quality program should be transitioned as-is into the new NFPA 805 FP Program. Provide a description of your current Fire Protection Quality Assurance Program (FPQA). Describe any changes being made to the FPQA Program as part of the NFPA 805 transition.
RAI 7-4 RESPONSE:
At this time, ONS does not foresee any substantive changes to the existing Fire Protection Quality Assurance Program. ONS will make a positive statement regarding moving the current Quality Assurance Program and/or program changes by revising section 4.7 of the Transition Report. This will be reflected in the LAR supplement.
70
ATTACHMENT 1 REGULATORY COMMITMENTS
Attachment 1 Regulatory Commitments The following table identifies the regulatory commitments in this document. Any other statements in this submittal represent intended or planned actions. They are provided for information purposes and are not considered to be regulatory commitments.
Commitment Due Date Due to methodology development and continued work required to December 17, 2009 implement proposed RG 1.205, Revision 1, the anticipated commitment date for the supplement is being delayed. The new date will be communicated to the NRC when a project schedule loaded with the additional activities required to meet the new guidance has been completed.
In the Cover Letter, Duke will provide a statement under oath and TBD affirmation as to the state of the required evaluations and analyses that support the transition to a risk-informed, performance-based fire protection program. (RAI 1-5)
Revise the Transition Report submitted October 31, 2008 to reflect the following: I Incorporate by reference the portions of FAQ 06-0008, Revision 9 (ML0090560170) which were approved within the associated closure memo (ML073380976). (RAI 1-6)
Section 1.1.1 and the footnote in the Transition Report will be revised to reflect that NFPA 805 was incorporated by reference.
(RAI 1-3)
Section 3.1 will be revised to reflect the list of committed modifications. (RAI 1-4)
Section 4.1 of the Transition Report will be reviewed and enhanced as necessary with regards to the use of multiple compliance statements. (RAI 2-9a)
Add sections 4.2.2.1.1 and 4.8.2.2.1 and associated tables to the Transition Report. (RAI 2-8a)
The appropriate revision of the FAQ 07-0040 closure memo will be referenced in Transition report sections 4.3.1 and 4.3.2 (or their successors), as well as Tables 4-1, F-1 and H-1. (RAI 3-12)
Revise section 4.6.2 to more clearly explain the monitoring process being utilized. (5-21) 1 Regulatory Commitments Commitment Due Date Section 4.7 of the Transition Report will be revised to make a positive statement regarding moving the current Quality Assurance Program and/or program changes. (RAI 7-4)
Rewrite of Attachment A, Table B-1 to include the following:
- Review to ensure compliance strategies are correctly addressed and captured for all Chapter 3 Sections. (RAI 2-9a)
- The text of the non-endorsed sections of NFPA 805 will be removed. Specifically this includes the exception of Section 3.3.5.3 and the exception of Section 3.6.4 not endorsed under 10 CFR 50.48(c), (c.2.v and vi). (RAI 2-10)
- A positive statement that all conditions for approval are met and still in effect will be provided for all cases where "Comply via Previous Approval" is cited for the compliance basis.
These changes will be reflected in the revised Table B-I.
(RAI 2-11)
- The Table B-1 response to NFPA 805, Section 3.3.1.1 will be revised to state "Upon review of the elements listed below, ONS believes that the NFPA 805 code requirements are satisfied and no additional elements were evaluated."
(RAI 2-12)
- The Table B-1 response to NFPA 805, Section 3.3.1.2 will be revised to state "Upon review of the elements listed below, ONS believes that the NFPA 805 code requirements are satisfied and no additional elements were evaluated."
(RAI 2-12)
- The Table B-1 response to NFPA 805, Section 3.3.1.2(5) will be revised to indicate that no other NFPA standards were determined to be applicable based on guidance in FAQ 06-0020. (RAI 2-12)
- The Table B-1 response to NFPA 805 Section, 3.3.1.2(6) will be revised to indicate that no other NFPA standards were determined to be applicable based on guidance in FAQ 06-0020. (RAI 2-12)
- Compliance with NFPA 241 is addressed through compliance with NFPA 51B. NFPA 241, 2000 edition, as referenced by NFPA 805-2001 ed., Section 5.1.1, with respect to hot work, states "Responsibility for hot work operations and fire prevention precautions, including permits and fire watches, shall be in accordance with NFPA 51 B, Standard for Fire Prevention During Welding, Cutting, and Other Hot Work." The Table B-1 response to NFPA 805, Section 3.3.1.3.1 will be revised to state this position. (RAI 2-12)
- _The Table B-1 response to NFPA 805, Section 3.4.1 will be 2
Regulatory Commitments Commitment Due Date reviewed to ensure point-by-point compliance statements are provided for the subsections of this element. The response to the subsections of Section 3.4.1 will be revised as necessary.
In addition, a positive statement that ONS complies with NFPA 600 via engineering evaluation (NFPA 1500/1582 does not apply) will be provided. (RAI 2-12)
" The Table B-1 response to NFPA 805, Section 3.4.2.1 will be enhanced to clearly outline/list the contents of the ONS Fire Plan. (RAI 2-12)
- The Table B-1 response to NFPA 805, Section 3.4.3(a) will be reviewed to ensure point-by-point compliance statements are provided for the subsections of this element. The response to the subsections of Section 3.4.3(a) will be revised as necessary. (RAI 2-12)
- The Table B-1 response to NFPA 805, Section 3.4.3(c) will be reviewed to ensure point-by-point compliance statements are provided for the subsections of this element. The response to the subsections of Section 3.4.3(c) will be revised as necessary. (RAI 2-12)
- The Table B-1 response to NFPA 805, Section 3.5.13 will be revised to ensure the seismic portion of this requirement is addressed. (RAI 2-12)
- The Table B-1 Compliance Statement to NFPA 805, Section 3.6.2 will be revised to "Comply by Engineering Equivalency Evaluation". (RAI 2-12)
" ONS utilizes the exception to NFPA 805, Section 3.11.1 where a performance-based analysis is used to determine building separation. These analyses are documented in the referenced Engineering Evaluations. The Table B-1 response to NFPA 805, Section 3.11.1 will be enhanced to clarify building separation. (RAI 2-12)
- ONS complies with clarification with regards to NFPA 101, 2000 edition as referenced by NFPA 805. NFPA 101, Section 8.2.3.2. 1(a) with regards to rated fire door assemblies refers to NFPA 80. NFPA 101 Section 9.2.1 with regards to rated fire dampers refers to NFPA 90A. ONS is performing detailed code compliance reviews for both NFPA 80 and NFPA 90A.
The Table B-1 response to NFPA 805, Section 3.11.3 will be revised to clarify the position that NFPA 101 compliance is achieved through compliance with NFPA 80 and NFPA 90A.
(RAI 2-12)
" The "Document Detail" field in the Table B-1 will be completed, as necessary, for clarification/traceability. Complete and clear reference documentation and associated document details will be Drovided. (RAI 2-13) 3 Regulatory Commitments Commitment Due Date Rewrite of Attachment B, Table B-2 to include the following:
Revise to reflect Calculation OSC-9291, NFPA 805 Transition B-2 Table, which is being revised to incorporate a greater level of detail on the alignment bases statements of the B-2 table to ensure reviewers can conclude conformance to the cited NEI 00-01 revision 1 sections. (RAI 3-33)
- Nuclear Safety Capability Assessment Methodology Review will be updated to address the methodology utilized in evaluating high/low interface valves in comparison to the guidance provided in NEI 00-01, revision 1. This will include a statement referencing the utilization of FAQ 06-0006 to determine the acceptable definition of high/low pressure interface valves.
Also, calculation OSC-9659, Nuclear Safety Capability Assessment, will include a basis statement in the methodology section pertaining to high/low interface valves. This basis statement will include the use of NEI 00-01, revision 1. (RAI 3-35)
The impact and availability of the fire detection and suppression systems needed to protect the success path required by NFPA 805 Section 4.2.3 or 4.2.4 will be included in the Nuclear Safety Capability Assessment. Any deviations to the required success path will be evaluated and included in the LAR supplement (RAI 3-34)
Rewrite of Attachment C, Table B-3 to include the following:
- An overview of how the committed modifications are to be used in the new NFPA 805 Fire Protection Program. (RAI 1-4)
" Attachment C and Table 4-4 will be revised based on the expected revision to NEI 04-02. (RAI 3-13)
- Include disposition of issues identified by calculation OSC-9659, Oconee Nuclear Safety Capability Assessment for Units 1, 2, and 3. (RAI 3-31)
" The final disposition of the Multiple Spurious Operations identified in Attachment C, NEI 04-02 Table B-3 Fire Area Transition, will be documented. (RAI 5-25)
Attachment G - Operator Manual Actions will be revised as follows:
The term challenging active fire was used incorrectly. It should have stated ... "Embedded in the compliance strategy is the assumption that the 10 minute time frame does not start until 4
Regulatory Commitments Commitment Due Date confirmation of an active fire." (RAI 3-1 and 3-28)
- A revised list of recovery actions for each fire area will also be included. The completion date for this document is also contingent upon the pending outcome of FAQ 07-30 and its impact on change evaluations, recovery actions, and required plant modifications. (RAI 3-10)
- Revise to adopt the definition of 'Primary Control Station" from RG 1.205, Draft Rev. 1 Regulatory Position 2.4 (October 2009, ML092881133). The existing operator manual actions will be screened against this definition to identify any additional recovery actions. The additional risk of their performance will be calculated in a fire risk evaluation and included. (RAI 3-17)
" Revise such that recovery actions that demonstrate the availability of a success path will be evaluated for the additional risk presented by their use. The inclusion of pre-transition Operator Manual Actions into the population of post transition Recovery Actions will be based on the process identified in RAI 3-18. (RAI 3-22 and 3-24)
Attachment L will be revised to reflect the following:
- Attachment L will be revised to reflect completion of the NFPA code compliance reviews. (RAI 2-4 and RAI 2-9b)
- Attachment L of the Transition Report will be revised to add specific detail regarding meeting the acceptance criteria of 10 CFR 50.48(c)(2)(vii). (RAI 2-14)
Attachment P will be deleted, along with reference to P-2.
(RAI 1-6)
Attachment S will be revised to reflect correct modifications. Each committed modification will include a summary of the proposed modification scope, a schedule for implementation of the modification, and a proposed License Condition reflecting the commitment and schedule for installation of the modification. (RAI 1-4, 3-31, and 5-23)
The Fire PRA will be updated after the breaker coordination study TBD is completed. (RAI 5-57)
The Fire PRA will be evaluated after the scope of the IN 92-18 TBD resolutions is fully evaluated. ,(RAI 5-20)
Complete breaker coordination study. (RAI 3-29) July 31, 2010 The revised Fire Risk Evaluations will be written in accordance with TBD the-guidance of Draft Regulatory Guide 1.205, Revision 1 and are expected to be included with the LAR supplement (RAI 5-40, 41, 43, 46, 47, 68) 5 Regulatory Commitments Commitment Due Date Duke will revise the treatment of additional risk of recovery actions. TBD This treatment will generally be in accordance with the process documented in RAI 3-18. (RAI 3-18, 3-19, 3-20, 3-21, 3-23, 5-15 and 5-16)
Detrimental risk impacts will be revisited after finalization of the TBD updated operator manual action list as part of the revised LAR.
(RAI 5-61) 6