ML091040214

Calculation No. OSC-9314, Rev. 0, NFPA 805 Transition Risk-Informed, Performance-Based Change Evaluation Methodology.

ML091040214
Person / Time
Site:	Oconee
Issue date:	10/28/2008
From:	Duke Energy Carolinas
To:	Office of Nuclear Reactor Regulation
Shared Package
ML091040234	List: ML091040205 ML091040206 ML091040207 ML091040208 ML091040214
References
OSC-9314, Rev 0
Download: ML091040214 (35)

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FORM 805.3 (R01-08) H CERTIFICATION OF ENGINEERING CALCULATION - REVISION LOG Station And Unit Number Oconee Units 1, 2, and 3 Title Of Evaluation NFPA 805 Transition Risk-Informed, Performance-Based Change Evaluation Methodology Evaluation Number OSC-9314 Active Evaluation / Analysis Yes L No J Evaluation Pages (Vol) Supporting Documentation Volumes Orig Chkd Vedf. App Issue

...... (Vol) Meth. __A-'r Date Rev. 1,2,3, Rec'd No. Revised Deleted Added Revised Deleted Added Deleted Added Date Date 'Other" Date Date Note 1: When approving an Evaluation revision with multiple Originators or Checkers, the Approver need sign only one block.

iii

L .....:.UnitCALCULATION IMPACT ASSESSMENT (CIA) ntion / Unit Oconee Units 1, 2, & 3 Calculation No. OSC-9314 Rev. 0 Page

_P No. (if applicable) 0-08-02520 By Ira M. e erly, Date ;oz.ifIo?

Prob. No. (stress & s/r use only) N/A Checked By, Date /67:ýýt_70 V Note: A NEDL search is NEDL reviewed to identify calculation? N required for (formally SAROS) [calculation originals (i.e.

Rev. O's)

Identify in the blocks below, the groups consulted for an Impact Assessment of this calculation originationlrevision.

Indiv. Contacted/Date Indiv. Contacted/Date

[]RES [ NGO

[Power, I & C, ERRT, [QA Tech. Services (ISI),

Reactor] Severe Accident Analysis, Elect. Sys. & Equip., Design

& Reactor Supp., Civil M

MCE Steve Jarrett* Structural, Core Mech. &

[Primary Systems, Balance Harold Lefkowitz* T/Hl Analysis, Mech. Sys. &

of Plant, Rotating Equipment, Equip., Nuclear Design, Valves & Heat Exchangers, Safety Analysis, and Civil] Matls/Metallurgy/Piping]

.[] MOD PRA Brandi Weaver*

[Mechanical Engr., Electrical E] Training Engr., Civil Engr.]

El Local IT ID Operations - OPS Support Cam Eflin* [E Regulatory Compliance C1 Maintenance - Tech. Support [] Chemistry

.] Work Control - Program. Supp. 0 Radiation Protection

[] Other Group LI No Group required to be consulted

• - These consultationsoccurred aspartof the development of the NFPA 805 TransitionLAR.

Listed below are the identified documents (ex: TECHNICAL SPECIFICATION SECTIONS, UFSAR SECTIONS, DESIGN BASIS DOCUMENTS, STATION PROCEDURES*, DRAWINGS, OTHER CALCULATIONS, ETC.) that may require revision as a result of the calculation origination or revision, the document owner/group and the change DOCUMENT GROUP CHANGE REQUIRED OSC-9292 MCE No change required OSC-9518 NGO No change required OSC-9375 NGO No change required PIP 0-08-02520 MCE No change required (Attach Additional Sheets As Required) iv

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 1 of 14 1.0 PROBLEM STATEMENT

1.1 DESCRIPTION

OF THE PROPOSED CHANGE During the transition to NFPA 805 variances from the deterministic requirements of NFPA 805 or with the current licensing basis will be evaluated using the 'Change Process' outlined in NEI 04-02. These variances are based on the regulations that were applicable to Oconee Nuclear Station (ONS) prior to the transition to a 10 CFR 50.48(c) FPP and would be limited to cases where the nuclear safety performance criteria, also referred to as fire protection program variances, of NFPA 805 are not met During the transition to NFPA 805, a change evaluation calculation will be performed for each fire area containing variances from the deterministic requirements of the current licensing basis.

Change evaluations will result in a demonstration that the variances are compliant with the acceptance criteria of NFPA 805 and therefore are in compliance with the NFPA 805 performance based approach.

The purpose of this calculation is to document the methodology to be used in the performance of the change evaluations created for the NFPA 805 Transition.

1.2 CALCULATION CLASSIFICATION This calculation provides a methodology to be used in support of the transition of the Fire Protection Safe Shutdown Program in accordance with 10CFR50.48(c). Since the transition documentation generated using this methodology is considered QA-I (see Enclosure, Section 1.2 for additional information) this calculation is classified as QA-1.

2.0 DESIGN METHOD/ANALYTICAL METHOD

2.1 BACKGROUND

The methodology for the change evaluations is based on the guidance in NFPA 805 and Section 5.3 of NEI 04-02, Guidance for Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 1, as modified by Frequently Asked Questions (FAQs) 06-0002 and 06-0003, and Regulatory Guide 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants.

Section 5.3 of NEI 04-02 provides requirements and guidance for performing risk-informed, performance-based change evaluations. Frequently Asked Questions (FAQs)'06-002 and 06-003 provided clarifications and minor modifications on the process. Nuclear Regulatory Commission (NRC) clarification of the change evaluation process is provided in Section C.3.2 of Regulatory Guide 1.205. The change evaluation process consists of the following subtasks (Figure 1 depicts the NEI 04-02 process as envisioned for post transition. Figure 2 reflects the process to be used during transition.):

'Both the NRC and the industry recognized the need for additional clarifications to the guidance provided in RG 1.205, NEI 04-02, Revision 1, and NFPA 805. The NFPA 805 FAQ process was jointly developed by NEI and NRC to facilitate timely clarifications of NRC positions.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 2 of 14

" Defining the Change

" Preliminary Risk Screening

" Risk Evaluation

" Acceptance Criteria Assessment In addition, the fire protection features and systems relied upon in the change evaluations will be assessed for inclusion into the monitoring program.

The following subsections describe the methodology used to evaluate the acceptability of this risk-informed, performance-based change evaluation.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 3 of 14 Defining the Change (NEI 04-02 Section 5.3.2) fincluding FAQ 06-002 Rev. 1]

Preliminary Risk Screening (NE 04-02 Section 5.3.3)

Risk Evaluation (NEI 04-02 Section 5.3.4)

Acceptance Criteria (NEI 04-02 Section 5.3.5)

Figure 1 - Change Evaluation Process (based on NEI 04-02 Figure 5-1)

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 4 of 14 Definition of Change Risk Evaluation Calculation of Delta CDF/LERF Review of Acceptance Criteria Evaluate Features for inclusion in Monitoring Figure 2 - Change Evaluation Process Modified for Transition

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 5 of 14 2.2 DEFINING THE CHANGE The Change Evaluation process begins by defining the variant condition to be examined and the baseline (compliant) configuration as defined by the Licensing Basis (current pre-transition licensing basis). Refer to NEI 04-02 Sections 5.3.2 and J.2 for more detail.

The Baseline Condition is defined as that plant condition or configuration that is consistent with the Licensing Basis (pre-transition licensing basis, shown as Case 2 in Figure 1). The variant condition or configuration, either 'as found' or proposed by a plant change, that is not consistent with the Licensing Basis is defined as the Variant Condition (shown as Case 1 in Figure 3).

Variant Condition Risk CASE 1 ACDF Risk Baseline Condition Risk (Compliant with CLB)

CASE 2 A CDF = CDFCASE I - CDFCASE 2 Figure 3 - Baseline versus Variant Conditions The 'changes' associated with NFPA 805 transition are those fire protection program variances that will be addressed using the NFPA 805 performance-based approach (via the change evaluation process). These changes are identified in Calculation OSC-9292, NFPA-805 Transition B-3 Table/Report, as Open Items requiring 'change evaluations'.

Due to the nature and complexity of individual changes that will be addressed as part of the transition change evaluation, it is necessary to organize and group individual changes. To the maximum extent possible, variances being addressed by change evaluations should be organized by plant location (i.e., fire area). The rationale for this grouping is:

" Key analytical tools for measuring compliance (e.g., safe shutdown analyses for 10 CFR

50) are organized in this manner. This will facilitate the clear documentation of a

'compliant case' for use in the change evaluation.

" Analytical tools for measuring fire risk (i.e., Fire PRA) are primarily spatially oriented and can focus analyses on specific targets and scenarios.

" This grouping supports reporting requirements for the NFPA 805 transition License Amendment Request (LAR) in Regulatory Guide 1.205.

The definition of the change should include the following information:

" Background - A short discussion of the performance goal affected by the change evaluation.

" Description of the Variant Condition - describe the component/cabling or components/cabling that cause the variant condition.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 6 of 14 Description of how the variant condition will be modeled in the Fire PRA (FPRA) -

describe how and why the variant condition will be modeled in the FPRA, including the basic events that will be manipulated. If the variant condition is not modeled in the FPRA, the rationale for exclusion should be noted.

2.3 PRELIMINARY RISK SCREENING Once the definition of the change is established and groupings/organizations are completed, a preliminary risk screening may be performed to identify and resolve minor changes to the fire protection program. Refer to NEI 04-02 Sections 5.3.3 and J.3 for additional detail.

It is not expected that the reviews associated with change evaluations during the transition to NFPA 805 will identify trivial or editorial changes or those that have minimal risk impact (since the focus of the transition change evaluation is to address variances from the current licensing basis). However, a record of changes that were screened from detailed evaluation should be developed for traceability.

2.3.1 SCREENING OF TRIVIAL OR EDITORIAL CHANGES The Evaluator shall determine if the change is trivial, based upon guidance in NEI 04-02.

Examples include:

" Changes to titles in procedures or program documents

" Change to Fire Brigade Training facility that has no impact on established training scenarios

" Changes to the Combustible Control Form that does not affect content.

" Changes to document layout.

" Changes to document numbers.

During the initial transition to NFPA 805, no open items were screened using the trivial or editorial criteria summarized above.

2.3.2 SCREENING OF MINIMAL RISK IMPACT CHANGES If the change is determined not to be trivial, the Evaluator shall perform a preliminary risk review, using the guidance in NEI 04-02, Appendix I, Attachment 3. If any of the preliminary risk review questions have "potentially greater than minimal" impact, a detailed quantitative risk evaluation is required.

During the initial transition to NFPA 805, no open items were screened using the minimal risk criteria summarized in the referenced NEI 04-02 guidance.

2.4 RISK EVALUATION' The FPRA is adequate to support the NFPA 805 change evaluation process. In lieu of a peer review, the NRC conducted of review of the ONS FPRA March 17-21, 2008. The NRC review, which was required by Regulatory Guide 1.205, noted that there were a number of incomplete supporting requirements communicated as 'findings'. Most of the findings from the NRC review have either been addressed or deemed to have no impact on FPRA quantification. Additionally, the FPRA has been found to meet Capability Category II in most but not all cases; those areas where the Capability Category II requirement was not met were evaluated and found to have no

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 7 of 14 impact on the application., Finally, the unresolved open items and any identified uncertainties with potential impact on the FPRA results have been addressed in the sensitivity analysis. Based on the evaluations performed and the sensitivity analyses (Refer to OSC-9518, ONS NFPA 805 Application Calculation), it was concluded that the FPRA was adequate for use in the change evaluation process.

It is noted that only the Unit 3 FPRA was reviewed by the NRC. That review is considered sufficient for Units 1 and 2 since the approach applied for development of the Unit 3 FPRA was not significantlyaltered. For Unit 2, a separate fault tree model was developed and a quantification file was built using the methodology contained in the Unit 3 FPRA documentation file. For Unit 1, a comparative screening analysis was performed which demonstrated that a separate Unit 1 fault tree and FPRA quantification file are not necessary.

The Evaluator shall coordinate as necessary with the Safe Shutdown Engineer, Fire Protection Engineer and Fire PRA Engineer to performlrevise the calculations to assess the change using risk-informed, performance-based techniques (including, but not limited to fire modeling and PRA). The risk evaluation may be in the form of a limiting or bounding fire modeling/fire risk analysis or a detailed integrated analysis. Refer to NEI 04-02 Sections 5.3.4, J.4, and J.5 for more detail.

The risk evaluation, as discussed in NEI 04-02 may consist of a number o}f steps Initial, Evaluation (Sections 5.3.4.1, 5.3.4.2, J.4.1, and J of NEI 04-02) o Fire Modeling o Bounding Risk Assessment Detailed Risk Evaluation (Sections 5.3.4.3 and J.5 of NEI 04-02)

This step as outlined in Section 5.3.4.1 and 5.3.4.2 of NEI 04-02 can be used to perform bounding fire modeling and bounding risk assessments. This step can be performed in order to optimize the effort necessary to successfully document the change evaluation.

If the Initial Evaluation cannot demonstrate the acceptability of the change, a detailed Risk Evaluation shall be performed. This is discussed further in Appendix J of NEI 04-02.

If the change involves equipment/cables required only for cold shutdown or whose function is not modeled in the PRA, then a qualitative risk assessment will be performed. This qualitative assessment should include the following:

" The desired safe end state for the traditional treatment of post fire safe shutdown under the provisions of 10 CFR 50.48(b) is cold shutdown. The transition to invoke the provisions of 10 CFR 50.48(c) includes the use of a Fire PRA. The safe end state evaluated in a PRA is not cold shutdown, but is instead a condition characterized as "safe and stable"..

This is typically hot standby/shutdown conditions. The PRA treatment of the plant response to a fire event does not necessarily require or credit the use of plant systems exclusive to cold shutdown. As such, the treatment of any such systems and functions in the context of a change evaluation would generally result in no measurable impact on the calculated plant risk.

" There are however, some possible exceptions.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 8 of 14 o If the fire induced plant transient is of such a nature that in order to achieve safe and stable conditions, .cold shutdown related systems and/or functions are required, the PRA would inherently require those functions to be successful. In these cases, the calculated risk metrics for the postulated fire event includes the consideration of failures that would disable there systems and/or functions, or o If the variance would affect achievement of a key safety function during a non-power higher risk-evolution, then deterministic options should be considered in accordance with the non-power operations methodology.

The following programmatic level items have the potential to impact the results of the FPRA and are included here due to their potential for impacting the results contained in OSC-9518, ONS NFPA 805 Application Calculation.

" The breaker coordination analysis for Oconee is not complete, and therefore the extent to which this may increase conditional core damage probability (CCDP) is currently unknown. The current FPRA does not assume any breaker coordination issues.

" For risk reduction, the baseline Fire PRA includes the Protected Service Water and the 3TC/3TD cable reroute/protection modifications.

" The Unit 3 PRA model has known issues that have been identified and are being tracked in the PRA Tracker item database. These model discrepancies have been assessed and do not affect the final delta CDF or delta LERF results.

" The review of the Unit 3 PRA model also included items issues related to application of the Regulatory Guide 1.205. These items have been assessed and do not affect the final delta CDF or delta LERF results.

" As part of the ongoing pilot plant effort, various topics needing clarification or additional guidance are being addressed through the Frequently Asked Questions (FAQ) process.

Several of these items have the potential to directly or indirectly affect the FPRA and/or the change evaluation process. FAQs have been incorporated into the change process to the extent practical. A discussion of methods that may be considered beyond those contained in NUREG/CR-6850 is included in the Oconee Fire PRA Scenario Development calculation, OSC-9375, Revision 0.

2.5 ACCEPTANCE CRITERIA ASSESSMENT For the transition to NFPA 805, the overall acceptance of the transition change evaluation will be in the form of a license amendment per 10 CFR 50.90, as required by 10 CFR 50.48(c)(3)(i).

Refer to NEI 04-02 Sections 5.3.5 and J.6 for more detail. Acceptance criteria for fire area change evaluations are based on ensuring:

" Quantitative Risk Acceptance Criteria

" The change in core damage frequency (ACDF) is acceptable, and

• The change in large early release frequency (ALERF) is acceptable, and

" Defense-in-depth is maintained

" Safety margins are maintained

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 9 of 14 The acceptance criteria for the change evaluation consist of two parts. One is quantitatively based and the other is qualitatively based. The quantitative figures of merit are ACDF and ALERF. The qualitative factors are defense-in-depth and safety margin. If a change meets the acceptance criteria described below, this is confirmation that a success path effectively remains free of fire damage (Ref. NEI 04-02, Sections 5.3.4.2 and 5.3.5).

QUANTITATIVE RISK ACCEPTANCE CRITERIA The transition risk evaluation shall be measured quantitatively for acceptability using the ACDF and ALERF criteria from Regulatory Guide 1.174, as clarified in Section 5.3.5 of NEI 04-02 and Regulatory Guide 1.205. The results of the acceptability determination shall be clearly documented in the calculations/analyses.

From an overall transition perspective, the acceptance criteria are provided in Regulatory Guide 1.205, Section C.2.2:

"The total change in risk associatedwith the transition to NFPA 805 should be consistent with the acceptanceguidelines in Regulatory Guide 1.174, "An Approachfor Using ProbabilisticRisk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis." I The acceptance criteria of Regulatory Guide 1.174 are summarized in the following table.

Region ACDFlyr ALERF/yr Status CommentslConditions 1 > I.OE-05 > I.OE-06 Unacceptable Proposed changes in this region are not acceptable.

11 < 1.OE-05 < 1.OE-06 Acceptable Proposed changes in this region are acceptable and > and -ý N/conditions provided the cumulative total CDF from all CDF I.OE-06 1.OE-07 initiators is less than 1.0E-04Iyr and from all LERF initiators is <1 E-5/yr. Cumulative effect of changes must be tracked and included in subsequent changes.

Ill < 1.OE-06 < 1.OE-07 Acceptable Proposed changes in this region are acceptable w/ conditions provided the cumulative total CDF from all initiators is less than I.OE-03/yr and from all LERF initiators is <1 E-4/yr. Cumulative effect of changes must be tracked and included in subsequent changes.

If the risk evaluation determines that ACDF and ALERF are acceptable and that defense-in-depth and safety margins are maintained, then the Evaluator shall document the results. This is confirmation that a success path effectively remains free of fire damage.

If the risk evaluation determines that either ACDF or ALERF are not acceptable, then the Evaluator shall document that the results are not acceptable and alternatives should be pursued until the quantitative acceptance criteria are met.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 10 of 14 OSC-95 18, ONS NFPA 805 FPRA Application Calculation, contains the methodology associated with calculating these values.

DEFENSE-IN-DEPTH A review of the impact of the change on defense-in-depth shall be performed, using the guidance below from NEI 04-02. NFPA 805 defines defense-in-depth as:

• Preventing fires from starting

• Rapidly detecting fires and controlling and extinguishing promptly those fires that do occur, thereby limiting damage

• Providing adequate level of fire protection for structures, systems and components important to safety; so that a fire that is not promptly extinguished will not prevent essential plant safety functions from being performed.

In general, the defense-in-depth requirement is satisfied if the proposed change does not result in a substantial imbalance among these elements.

The review of defense-in-depth is typically qualitative and should address each of the elements with respect to the proposed change. Defense-in-depth may be assessed at a compartment, fire scenario, or fire area basis if applicable to multiple changes.

Fire protection features and systems relied upon to ensure defense-in-depth should be clearly identified in the assessment (e.g., detection, suppression system).

Reliance on OMAs (recovery actions) in lieu of protection is considered part of the third echelon of defense-in-depth.

NOTE: Per NFPA 805, recovery actions are defined as: "Activities to achieve the nuclear safety performance criteria that take place outside of the main control room or outside of the primary control(s) station for the equipment being operated, including the replacement or modification of components."

Consistency with the defense-in-depth philosophy is maintained if the following acceptance guidelines, or their equivalent, are met:

1. A reasonable balance is preserved among 10 CFR 50.48(c) defense-in-depth elements.

2. Over-reliance and increased length of time or risk in performing programmatic activities.

to compensate for weaknesses in plant design is avoided.

3. Pre-fire nuclear safety system redundancy, independence, and diversity are preserved commensurate with the expected frequency and consequences of challenges to the system and uncertainties (e.g., no risk outliers). (This should not be construed to mean that more than one safe shutdown train must be maintained free of fire damage.)

4. Independence of defense-in-depth elements is not degraded.

5. Defenses against human errors are preserved.

6. The intent of the General Design Criteria in Appendix A to 10 CFR Part 50 is maintained. For ONS the applicable GDC are contained in UFSAR, Chapter 3.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 11 of 14 It should be noted that all elements of fire protection defense-in-depth may not exist for beyond design basis fire scenarios. For example, a conditional core damage probability of 1.0 is possible if enough fire barriers are breached. Such beyond design basis scenarios, however, should be demonstrated to be of less risk significance, with certainty. A scenario that does not lead to core damage and a CDF of 9E-08/year would be treated differently than a scenario that leads to core damage and a CDF of 9E-08/year. In the end, the balance results in consideration of all aspects of the scenario, including the risk, defense-in-depth, safety margins, uncertainty, and other relevant issues.

SAFETY MARGIN A review of the impact of the change on safety margin shall be performed. An acceptable set of guidelines for making that assessment is summarized below. Other equivalent acceptance guidelines may also be used.

a Codes and standards or their alternatives accepted for use by the NRC are met, and a Safety analysis acceptance criteria in the licensing basis (e.g., FSAR, supporting analyses) are met, or provides sufficient margin to account for analysis and data uncertainty.

The requirements related to safety margins for the change analysis is described for each of the specific analysis types used in support of the fire risk assessment.

These analyses can be grouped into three categories. These categories are:

" Fire Modeling

" Plant System Performance

" PRA Logic Model The following guidance on these topics is provided', Additional information is contained in NEI 04-02 Section 5.3.5.3.

Fire Modeling A combined analysis approach will be used during the transition change evaluations and therefore maximum expected fire scenario (MEFS)/limiting fire scenario (LFS) will not be analyzed separately. From a safety margin perspective the evaluation of the fire modeling performed should address the following topics:

" Was the fire modeling performed using codes/methods acceptable to the NRC?

" Were the heat release rates used conservative?

" If the treatment in the FPRA, resulted in no damage to the targets of concern, and an LFS is not calculated, ensure that over reliance is not placed on the third element of defense-in-depth (i.e., operator manual actions).

Plant System Performance The development of the fire risk assessment may involve the re-examination of plant system performance given the specific demands associated with the postulated fire event. The methods, input parameters, and acceptance criteria used in these analyses need to be reviewed against that used for the plant design basis events. This review would serve to establish that the Safety

CalculationNo. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 12 of 14 Margin inherent in the analyses for the plant design basis events have been preserved in the analysis for the fire event and therefore satisfy the requirements of this section.

From a safety margin perspective the evaluation of the plant system performance should address the following topics:

V Were input parameters for plant performance analyses (e.g., heat-transfer coefficients, pump performance curves) altered from those used for plant design basis events such that margin was lessened?

- Were codes and standards used to determine plant system performance acceptable to the NRC?

PRA Logic Model The quantification for fire related CDF/LERF relies upon the FPRA model. It is recognized that use of a FPRA often requires model modifications to be performed to the internal events PRA.

These modifications may include altering basic event failure probabilities, adding basic events, and logic structure changes. These changes should be evaluated against the methods and criteria for the overall FPRA model development for consistency, or confirmation of bounding treatment, to confirm that the Safety Margin inherent in the PRA model is preserved.

From a safety margin perspective the evaluation of the PRA logic model should address the following topic:

. Were the risk-informed, performance based processes utilized based upon NFPA 805, 2001 edition, endorsed by the NRC in 10 CFR 50.48(c)?

a Was the change evaluation process in accordance with NEI 04-02, Revision 1, which is endorsed by the NRC in Regulatory Guide 1.205, Revision 0?

9 Was the FPRA developed in accordance with NUREG/CR-6850, which was developed jointly between the NRC and EPRI?

2.6 MONITORING OF CREDITED SYSTEMS, FEATURES AND PROGRAM ELEMENTS Section 2.6 of NFPA 805 states:

"A monitoringprogramshall be establishedto ensure that the availabilityand reliabilityof the fire protectionsystems andfeatures are maintainedand to assess theperformance of the fire protectionprogram in meeting the performancecriteria. Monitoringshall ensure that the assumptionsin the engineering analysis remain valid" The intent of the review will be to confirm (or modify as necessary) the adequacy of the existing surveillance, testing, maintenance, compensatory measures, and oversight processes for transition to NFPA 805. This review will consider the following:

1) The adequacy of the scope of systems and equipment within existing plant programs, i.e.,

the necessary fire protection systems and features and nuclear safety capability equipment (NFPA 805 Section 1.5.1) are included to ensure the assumptions and results of the change evaluation remain valid.

Calculation No- OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 13 of 14

2) The performance criteria for the availability, reliability or condition of fire protection systems and features relied on for the change evaluation.

3.0 DESIGN BASIS AND REFERENCE INFORMATION 3.1 National Fire Protection Association (NFPA) Standard 805-2001, Performance Based Standard for Fire Protection for Light Water Reactor Electric Generating Plants 3.2 Regulatory Guide 1.205, Risk-Informed, Performance-Based Fire Protection for Existing Light-Water Nuclear Power Plants, May 2006 3.3 NEI 04-02, Guidance For Implementing a Risk-Informed, Performance-Based Fire Protection Program Under 10 CFR 50.48(c), Revision 1 3.4 FAQ 06-0002, NEI 04-02 Section 5.3.3 and App. I, Order of Questions for Change Analysis Screening, Revision I c, Closure Memo ML070030276 3.5 FAQ 06-0003, Change Analysis Screening, Revision lb, Closure Memo ML070030242 3.6 Regulatory Guide 1.174, An Approach for Using Probabilistic Risk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis, Revision I 3.7 Engineering Directives Manual EDM-5 10, PRA Technical Adequacy, Rev. 0 4.0

SUMMARY

INFORMATION See Enclosure I for an outline of format and content.

4.1 CONCLUSION

The conclusion should determine that the variances identified in Section 1.1, Problem Statement are acceptable (with or without modifications) based upon:

" The measured change in CDF and LERF

" Adequate defense-in-depth and safety margins are maintained The conclusion should also state that the results of this evaluation meet the requirements of NFPA 805 change evaluation and the guidance of Regulatory Guide 1.205; therefore, the fire area is compliant using the risk-informed, performance-based approach and that a success path effectively remains free of fire damage and therefore the nuclear safety performance criteria of NFPA 805 are met.

4.2 DISPOSITION OF IDENTIFIED VARIANCES The disposition of the final compliance strategies for the identified variances should be documented in the 'Conclusion Section'. For example:

" Operator manual actions will be dispositioned as a required action, a defense-in-depth action, or the action is no longer required.

" The variance requires a modification to demonstrate compliance with either the deterministic or performance-based requirements of NFPA 805.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Page 14 of 14 4.3 DEFENSE-IN-DEPTH ENHANCEMENTS In Section 4.3 the Evaluator will discuss any enhancements to defense-in-depth elements made as a result of this change evaluation. This discussion is a summary of the information presented in Section 6.

4.4 MONITORING OF REQUIRED SYSTEMS, FEATURES AND PROGRAM ELEMENTS For each fire area the fire protection systems, features, and program elements that were relied upon should be listed for input into the monitoring program (note a comprehensive list of systems, features, and elements to be considered for monitoring are listed in the defense-in-depth table).

5.0 ASSUMPTIONS The following assumptions were used in this calculation:

0 No specific assumptions.

6.0 TECHNICAL PRESENTATION contains an outline of the format and content of the change evaluations.

Calculakion No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure I Page 1 of 18 ENCLOSURE 1 1.0 PROBLEM STATEMENT

1.1 DESCRIPTION

OF THE PROPOSED CHANGE In Section 1.1 of each change evaluation the following information should be presented:

" Open Item Identifier from Calculation OSC-9292, NFPA 805 Transition B-3 Table/Report.

" For each open item a brief description of the variant condition For fire areas that impact multiple fire areas, such as SSF, sub-divide the discussion by Unit. For example:

"1.1.1 UNIT I VARIANCES Afire in thisfire areadoes not result in an impact to Unit 1.

1.1.2 UNIT 2 VARIANCES Afire in thisfire areadoes not result in an impact to Unit 2.

1.1.3 UNIT 3 VARIANCES The changes identifiedfor evaluationfor the Unit 3 ReactorBuilding (FireArea RB3) are groupedasfollows:

Variances from Deterministic Separation Requirements - Based on the deterministic pre-transition safe shutdown analysis, Calculation OSC-9292, NFPA-805 Transition B-3 Table/Report, Revision 0, the following variances are evaluated in this calculation. See Attachment A for additional information associated with the variances.

" Open Item RB3-07-O - Potentialspurious opening ofpower operated reliefvalve (PORP) 3RC-66 with failure of the correspondingPOR V block valve (3RC-4) to close results in breach in the Reactor CoolantSystem (RCS) integrity that requires creditingan unallowed manual action to de-energize PORV block valve (3RC-4) control circuitrytofail the PORV block valve closed. This action is necessary to demonstrate compliance with the nuclearperformance criteriafor Inventory and PressureControl. (

Reference:

PIPNo.: 0-08-02520, CA Sequence No. 23) o Open Item RB3-11-OE - PotentialFailureofsteam generator (SG) Crosstie valve 3CCW-269, results in runout of 3A motor driven emergencyfeedwaterpump (MDEFDWP) and uncontrollablediversion of emergencyfeedwater (EFDW) to the non-creditedSG. This results in an unallowed manual action to remove powerfrom 3CCW-269 to precludepotentialfire effects. This action is necessary to demonstrate compliance with the nuclearperformance criteriafor Decay Heat Removal.

(

Reference:

PIPNo.: 0-08-02520, CA Sequence No. 25)."

o Open Item RB3-33 Potentialspurious closing of 3FDW-347 results in the isolation of the creditedfeedwaterflowpath to 3B steam generatorand the inability to

Calculation No. OSC.-93!4 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 2 of 18 cool down the RCS. This results in an unallowed manualaction to remove power from 3FDW-347 to precludepotentialfire effects. This action is necessary to demonstratecompliance with the nuclearperformancecriteriafor Decay Heat Removal and Inventory and Pressure Control. (

Reference:

PIPNo. 0-08-02520, CA Sequence No. 88) 1.2 CALCULATION CLASSIFICATION In section 1.2 of each change evaluation, the calculation classification should be addressed. For example,:

"This calculation is beingperformed in support of the transitionofthe Fire ProtectionSafe Shutdown Programin accordancewith IOCFR50.48(c). This calculation is considered QA-1 because it impacts systems, structures,or components (SSCs) that are QA-I (forpurposes other thanfire mitigation). Note thatper a 4/12/1995 letterfrom J. W. Hampton to the NRC, the NRC has specificallyapproved the use ofnon safety-relatedSSCs to mitigatefires.

The PRA insights associatedwith this calculationare not QA-1 basedon the guidanceprovided in Regulatory Guide 1.174, Revision 1, dated November 2002. Regulatory Guide 1.174 (An Approachfor Using ProbabilisticRisk Assessment in Risk-Informed Decisions on Plant-Specific Changes to the Licensing Basis) section 2.5 states:

"... To the extent that a licensee elects to use PRA information to enhance or modify activities affecting the safety-relatedfunctions of SSCs, thefollowing, in conjunction with the other guidance contained in this guide, describes methods acceptableto the NRC staff to ensure that the pertinentquality assurancerequirements ofAppendix B to 10 CFR Part50 are met and that the PRA is sufficient to be usedfor regulatory decisions..."

See EDM-510for how Duke Energy PRA meets the requirements of Regulatory Guide 1.174."

2.0 DESIGN METHODIANALYTICAL METHOD Section 2.0 of each fire area specific change evaluation should include the following information:

"The methodologyfor the change evaluations is based on the guidance in NFPA 805 and Section 5.3 ofNAE 04-02, Guidancefor Implementing a Risk-Informed, Performance-BasedFire ProtectionProgram Under 10 CFR 50.48(c), Revision 1, as modified by FrequentlyAsked Questions (FAQs) 06-0002 and 06-0003, and Regulatory Guide 1.2 05, Risk-Informed, Performance-BasedFire Protectionfor Existing Light-Water NuclearPower Plants.

NEI 04-02 provides requirementsand guidancefor performing risk-informed,performance-based change evaluations. FAQs 06-0002 and 06-0003 modified the process slightly. NRC clarificationof the change evaluationprocess is provided in Section C.3.2 ofRegulatory Guide 1.205. The Plant Change Process consists of thefollowing subtasks:

" Defining the Change

" Preliminary Risk Screening

" Risk Evaluation

Calculh.tion No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 3 of 18 Acceptance Criteria Assessment o The change incore damagefrequency (ACDF)is acceptable, and O The change in large early releasefrequency (ALERF) is acceptable, and o Defense-in-depthand safety margins are maintained.

The Acceptance CriteriaAssessment is evaluated utilizing the risk guidance outlinedin Regulatory Guide 1.174 and the plant specific risk as determined in OSC-9518, ONS NFPA 805 Fire PRA Application Calculation. In addition, thefire protectionfeatures and systems relied upon in the change evaluations are assessedfor inclusion into the monitoringprogram.

Background and methodologyfor performingchange evaluationsare documented in Calculation OSC-9314, NFPA 805 TransitionRisk-Informed, Performance-BasedChange Evaluation Methodology."

3.0 DESIGN BASIS AND REFERENCE INFORMATION

3.1 DESCRIPTION

OF THE FIRE AREA This section provides an overall discussion of the fire protection systems and features available in the area.

In Section 3.1 of each fire area specific change evaluation the following information should be presented:

• Section 3.1 - Description of the Fire Area An overall description of the area should be provided. This description should include r." interfacing boundaries. For zones in the area, elevations, and a general description of the example, "the south wall separatesthe FireArea XFire Areafrom the Unit 3 Reactor Building" Section 3.1.1 - Construction Provide a general description of the Fire Area boundary construction and its adequacy for the hazard. For example "The fire barrierseparatingthe East and West Penetration Rooms erected near the west corner of the Spent Fuel Pools is a 3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> ratedfire barrier constructed of a tube steel/pyrocrete. The fire barriersfor FireArea Xare 3 hourfire ratedconstruction or have been evaluated to be adequatefor the hazards. Applicable evaluationsare documented in thefollowing."

o OSC-7350, Att. 31, ONS Penetration Seal Database and 86-10 Evaluations, Rev. 6"

• Section 3.1.2 - Ventilation Provide a general discussion of the smoke removal capability provided for the area if necessary for an OMA and any specific assumptions/limitations associated with the fire modeling performed for the area. For example, "FireArea Xhas dedicatedsmoke removal system which can be alignedto exhaust smoke to the exterior ofthe building."

• Section 3.1.3 - Detection

Calculation, No. CSC-9314 Revisicn No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 4 of 18 A general description of the detection provided in the area. For example, "An ionization-type smoke detection system is providedfor FireArea X Alarms for the detection system are supervised in the Unit 3 ControlRoom."

a Section 3.1.4 - Fixed Fire Suppression A general description of the automatic fire suppressionprovided in the area. For example, "FireArea X has an automaticfire suppressionprovidingpartialarea coverage."

0 Section 3.1.5 - Manual Suppression A general description of the manual fire suppression available to fight a fire in the area will be described. For example, "A hose station is located on elevation 809' adjacent to Fire Area X The hose station is suppliedfrom interiorstation loopfire main and looped yardfire main system suppliedfrom high pressureservice waterfire pumps located in lower level of the Turbine Building. Portablefire extinguishers are located throughout FireArea X" X Section 3.1.6 - Fire Scenario Include a discussion of the fire scenario. For example, "FireScenarios evaluated in the FirePRA include a transient hot workfire, a transientfire, afire associatedwith ElectricalCabinet-6."

The information included in Section 3.0 of each fire area specific change evaluation should follow the requirements of EDM-101.

3.2 REFERENCES

This section provides references used as design input during the development of the change evaluation.

4.0

SUMMARY

INFORMATION Section 4.0 of the fire area specific change evaluation calculation shall contain the conclusions of the evaluation. These conclusions shall include:

" A statement that the change(s) met the acceptance criteria

• A disposition of the individual open items, and

" Suggested input to the monitoring program.

4.1 CONCLUSION

In Section 4. 1, the Evaluator shall state how the acceptance criteria are met. Examples of conclusion statement are provided below:

"The change evaluation determined that [upon completion of modifications and/orupon completion ofprogrammaticchanges] the variances identified in Section 1.1 are acceptable based upon:

Calculation No. 0SC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure I Page 5 of 18

" The measured change in CDF and LERF

" Adequate defense-in-depth and safety margins are maintained The results of this evaluation meet the requirements ofNFPA 805 Change Evaluationand the guidance ofRegulatory Guide 1.205; therefore, the fire area is compliant with the performance-based approach."

For fire areas that affect more than one unit, ensure that the conclusions are specific to the unit as necessary.

4.2 DISPOSITION OF IDENTIFIED VARIANCES In Section 4.2 the Evaluator will disposition each of the variances identified in Section 1. An example of this treatment is provided below:

Based on the change evaluationresults, the variances identified in section 1.1 have been dispositionedasfollows:

Table 42-1 F1nal VarianceD'isposlllon, Variance Description DisposAiMon Open Item FireArea X-10-O - Potentialspurious PORV No recovery action required.

opening/ failure of block valve, resulting in creditingan Necessity of a defense-in-depth unallowed manual action to remove powet to the de-energize action will be evaluated in OSC-PORV block valve (3RC VAO004) control circuitryto fail the 9535.

PORV block valve closed.

For fire areas that affect more than one unit, ensure that the variances are specific to the unit as necessary.

4.3 DEFENSE-IN-DEPTH ENHANCEMENTS In Section 4.3 the Evaluator will discuss any enhancements to defense-in-depth elements made as a result of this change evaluation. This discussion is a summary of the information presented in Section 6.

The following table represents those elements of defense-in-depth that were enhanced as a result of the change:

Table 4,3-L Defense-in-DeplAh Enhancements Defense-ln-Dep// Elemnent Dis~position Prevent fires from starting Acceptable as is.

Rapidly detect control and extinguish promptly those fires that Acceptable as is.

do occur thereby limiting fire damage

Cd!culation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 6 of 18 Table 4 3- /

Defense-in7-DeO/t En5ancOemenOs Defen~se-in-Dep/hE/eme/ DIsposition Provide adequate level of fire protectionfor systems and No recovery action required.

structuresso that a fire will not prevent essentialsafety Necessity of a defense-in-depth functions from being performed action will be evaluated in OSC-9535. (See Section 4.2 for additional information)

For fire areas that affect more than one unit, ensure that the defense-in-depth enhancements are specific to the unit as necessary.

4.4 MONITORING OF REQUIRED SYSTEMS, FEATURES AND PROGRAM ELEMENTS Upon completion of the change evaluations the Evaluator will review the assumptions and input in the analysis to verify / identify those systems, features and/or programmatic controls that need to be included in the monitoring program. Examples include:

" Combustible Free Zones established to ensure fire damage is limited.

" Suppression Systems credited for defense in depth or credited in the FPRA to limit damage

" Detection systems credited for defense in depth or credited in the FPRA to limit damage The results of this review shall be documented in Section 4.4 of the change evaluation ýý calculation. An example of this treatment is provided below (note a comprehensive list of systems, features, and elements to be considered for monitoring are listed in the defense-in-depth table):

The followingfire protectionsystems, features, andprogram elements are requiredby the FPRA and will be included in the monitoringprogram:

" Passive boundaries o ExternalFireArea Boundaries

• Programmatic/Administrative Controls o Transient ControlProgram (if 'combustiblefree zones 'are established- they should be discussed here) o Hot Work ControlProgram 5.0 ASSUMPTIONS 5.1 METHODOLOGY ASSUMPTIONS CalculationOSC-9314, NFPA 805 TransitionRisk-Informed, Performance-BasedChange Evaluation Methodology contains the methodology, including items which have been addressed

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure I Page 7 of 18 through the methodology. These items addresstopics which are programmaticin nature and which otherwise would be treated as assumptions in the areaspecific change evaluations. Refer to OSC-9314, Section 2.4for a list of these items.

5.2 AREA SPECIFIC ASSUMPTIONS 5.2.1 Transients and/or hotwork activities are not normally allowed in the Reactor Building duringpower operations.

5.2.2 Sufficient intervening combustibles do not exist to allowpropagationoffire between the east and west side of the ReactorBuilding.

6.0 TECHNICAL PRESENTATION 6.1 PRELIMINARY RISK SCREENING RESULT Section 6.1 of the change evaluation calculation shall document the results of the Preliminary Risk Screening It is not expected that the Transition Change Evaluations will screen at this stage. The following statements can be utilized: (1) For trivial or editorial changes, "The open item(s) contained in section 1.1 do not include any changes evaluatedto meet the trivial or editorialchange criteria and therefore detailed change evaluationswere performed." or (2) For minimal risk impact.

changes, "The open item(s) containedin section 1. 1 do not include any changes meeting the minimal risk impact change criteriaand therefore detailed change evaluationswere performed."

6.2 RISK EVALUATION RESULTS Section 6.2 of each change evaluation will document the results of the Risk Evaluation.

6.2.1 FIRE MODELING Enter the following statement in Section 6.2.1, Fire Modeling, "At this time, no specificfire modeling has been performed to support the change evaluationprocess. The combined analysis approachis used during transition(Ref NE1 04-02, 5.3.4.3); therefore, MEFS/LFS is not analyzed separatelyfrom the FirePRA results."

6.2.2 RISK EVALUTION For each variance identified in Section 1.1 provide the results of the Risk Evaluation in Section 6.2.2 of the change evaluation calculation. An example of the results table is provided below.

"Foreach varianceidentified in Section 1.1, the detailed risk evaluation determined the change in CDF andLERF: Of the scenarios identifiedin CalculationOSC-93 75, Oconee Fire PRA Scenario Development, Rev. 0, no scenarios had a calculated CDF; above IE-08/yr.

These values are summarized in the table below. Details regardingthe change evaluation are included in Attachment A."

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 8 of 18 Table 6 2-f 01uan/1allve Change Eva/ua8/oi1 /9esul/s Varance Description Del/a CDF Delta L ERF Enter each open item from B-3 Table Enter results of calculation Enter results of calculation requiringa change evaluation Total Enter Total Enter Total Detailsregardingthe change evaluation are included in Attachment A."

For fire areas that affect more than one unit, provide a Table for each Unit.

6.2.3 DEFENSE-IN-DEPTH AND SAFETY MARGIN In Section 6.2.3, of the change evaluation calculation the Evaluator shall provide the results of the Defense-in-Depth and Safety Margin Review DEFENSE-IN-DEPTH In this Section provide the results of the Defense-in-Depth review in the following tabular format. Care should be taken to only include those fire protection features, operator manual actions that are necessary to ensure that defense-in-depth is maintained. For example:

A review of the impact of the change on defense-in-depth was performed using the guidancefrom ATEI 04-02 (See OSC-9314, TransitionRisk-Informed, Performance-BasedChange Evaluation Methodology). In general,the defense-in-depth requirementis satisfiedif the proposed change does not result in a substantialimbalanceamong these elements. The table below summarizes how adequate defense-in-depth is maintained.

NOTE: Unless specificallynoted, the information below is applicablefor all 3 units.

Table 62-2 Defense-fn-Dep/h Impac!Review le/hod of Providin'1gDID Impact on Plant Changes andNet Effec5 Conf/gura/lAo and Contro Preventfires from starting Combustible Control is implemented Required This element of defense-in-in accordancewith NSD-313, Control depth remains unchanged of Flammable and Combustible Materials Hot Work Control is implemented in Required This element of defense-in-accordance with NSD-314, Hot Work depth remains unchanged Authorization

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure I Page 9 of 18 Table 6 2-2 Defense-7n-Depth moactReview IWeIhodofProv11d,17 DID ImnpaCt on Plant Changes and 87 et Elfect Conf/gura//on and Cconfrol Rapidly detect, control and extinguish promptly those fires that do occur thereby limiting fire damage Ionization-type smoke detection Credited This element of defense-in-system is provided depth remains unchanged Automatic fire suppression is not Not Credited Not Applicable provided in the area.

Portablefire extinguishers are Credited This element of defense-in-provided in the area. depth remains unchanged Hose stationsand hydrantslocated in Credited This element of defense-in-the area(s) depth remains unchanged Fire Pre-FirePlan Not Credited Not Applicable Provide adequate level of fire protection for systems and structuresso that a fire will not prevent essential safety functions from being performed Walls, floors ceilings and structural Required This element of defense-in-elements are rated or have been depth remains unchanged evaluated as adequate for the hazard.

Penetrationsin the fire area barrier Required This element of defense-in-are rated or have been evaluatedas depth remains unchanged

ý'i adequate for the hazard.

Discuss any supplemental barriers Not Credited Not Applicable (e.g., ERFBS, cable tray covers, etc.)

Discuss fire ratedcable Not Credited Not Applicable Reactor coolant pump oil collection Not Credited Not Applicable system (as applicable)

Guidanceprovided to operations Credited No recovery action required.

personnel detailing the credited Necessity of a defense-in-success path(s) including recovery depth action will be evaluated actions to achieve nuclearsafety in OSC-9535.

performancecriteria.

LEGEND:

" Required- DIDfeature is requiredfor inclusion in the MonitoringProgramto maintain FPRA inputs and assumptions.

" Credited- DIDfeature is not requiredperFPRA, however itprovides additionallevels of nuclear safety, and will be included in a plant trackingprogram.

" Not Credited- Not requiredfor DID.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1,2, and 3 Enclosure 1 Page 10 of 18 SAFETY MARGIN In this Section the Evaluator shall document the results of the qualitative safety margin review.

The treatment should evaluate, each issue presented in Section 2.5.3 and provide an answer/discussion of these issues. An example of this treatment is provided below:

"In accordance with NEI 04-02, 5.3.5.3 guidance, the maintenanceofadequate Safety Margin is assessed by the considerationof categoriesof analyses utilizedby this change evaluation. Safety margins are consideredto be maintained if

" Codes and standardsor their alternatives acceptedfor use by the NRC are met, and

" Safety analysis acceptance criteriain the licensing basis (e.g., FSAR, supporting analyses) are met, or providessufficient margin to accountfor analysis and data uncertainty.

The maintenanceof adequate safety marginfor the change analysis is describedfor each of the specific analysis types used in support of thefire risk assessment.

FIREMODEL ING Fire modeling performed in support of the transitionhas been performed within the FPRA utilizing codes and standardsdeveloped by industry andNRC staff to provide realisticyet.

conservative results. Specifically, the heat release rates utilized in the transitionanalysis are based upon NUREG 6850, Task 8, Scoping FireModeling. These heat releaserates are conservative and representvalues used to screen outfixed ignition sources that do not pose a threatto the targets within specificfire compartments and to assign severityfactors to unscreenedfixed ignitionsources.

The combined analysis approachis used during transition;therefore, maximum expectedfire scenario/limitingfire scenariohave not been analyzed separately. The basesfor the application of these fire modeling codes andstandardswere not altered in support of this change evaluation.

PLANT S YS TEM PERFORMAANCE Plantsystem performanceparameterswere not modified as a result of this change evaluation.

These performanceparameterswere originallyestablishedto support nuclearperformance criteriacontainedin' the plant specific accident analyses. These analyses established component and system performance criterianecessary to establish safe and stable plant operation,as well as, safe shut down of the unit in the event of afire. These performanceparameterswere not modified as a resultof this change evaluation.

PRA Z OGICMODEL Adequate safety margin is maintainedbecause the codes and standardsused have been accepted for use by the NRC andthe acceptance criteriain NFPA 805. The specific codes and standards used in the FPRA applicationand development were 10CFR50.48(c),NFPA 805, 2001 edition, NRC Regulatory Guide 1.205, Revision 0, andNUREG 6850. These codes andstandardswere applied in a manner which would provide FPRA results which contain and complement safety

Calculation No. OSC -9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 11 of 18 margin. The basesfor the applicationof these FPRA codes and standardswere not altered in support of this change evaluation.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 11 of 18 ATTACHMENT A - PRA CALCULATION RESULTS In the Attachments for to the change evaluations the PRA calculation results should be summarized for each unit (Attachment A for Unit 1, Attachment B for Unit 2 and Attachment C for Unit 3). The following discusses the concept/purpose of each section/sub-section and provides an example of content and level of detail.

A KEY PRA INSIGHTS UNIT 1 Section A. 1 should provide an overview of the fire area as follows:

" Fire zones in area

" Key PRA Components

" General layout of targets and combustibles 0 Discussion of the fire scenarios evaluated

" Discussion of the fire scenarios that will be evaluated as part of the change evaluation See example below:

A.1 FIRE AREA RB3 - UNIT 3 REACTOR BUILDING This fire area is comprised of the Unit 3 Reactor Building (RB3). Key PRA components/functionspotentially impacted include:

" Motor driven EFW Pumps (automaticSGLC operationonly)

" CC Pumps 3A & 3B

" Miscellaneous HPRI andLPI valves

" RCP's

" PressurizerHeaters

" Turbine Bypass Control

" PORV&PORVBlock

" SSF RCM Pump

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 12 of 18 Failureof CompressedAir systems is excluded in the Reactor Building (no creditedAOVs in RB). Also, sincepower is removedfrom 3RC VA0155, 156, 157, 158, 159, & 160 duringpoweroperations (reference OP 1103/002), spurious opening of the vent valves due to a fire isprecluded.Development of west side and east sidefire scenarioson either side of the shield wall within the Reactor Building relied on cable separation,location of key ignition sources and their associatedzones of influence, and lack of intervening combustiblesfor the assumed targetdamage. There are relativelyfew ignition sources in the Reactor Building that are capable of damaging multiple target sets, primarilythe reactorcoolantpumps.

Of the scenariosidentified in Calculation0SC-93 75, Oconee Fire PRA Scenario Development, Rev. 0, only RB03 ScenariosB1, D, and E had a calculatedCDFabove IE-08/yr and a calculatedLERF above 1E-09/yr. It is noted that containmentbypass scenarios involvingfailure of the HP20, & HP21 inboard and outboardisolation valvesfor afire in RB3 arepossible.

Refinements to the Altered Events table in FRANC were identified to support the delta risk calculations. The changes included the addition of circuitfailure likelihood valuesfor valves in containment bypass pathways and elimination of circuitfailure likelihood values for spuriousRCP operation. None of these changes hada significantimpact on the overallfire CDF/LERFresults. It is recommended that these changes be capturedin afuture revision to the FPRA.

The following tableprovides summary level information on the RB03 scenariosabove the screening thresholds.

Table A. f- I FC Sce n 'r/o Sc&en erio' Tergse's CDF LERF Description Damage to trays within shield wall result in loss of PORV and PORV Block and other RB03 B1 RCP 3B1/3B2 failures; Containmentbypass scenario 1.05E-08* 7,80E-09" Severe Fire (ISLOCA pathway via seal return).

Damage to trays between containment wall and shield wall in west half of Reactor RB3RB Fire - West Half Building MD EFDW result in loss of SSF RCM Pump, pumps, Turbine Bypass 7.96E-07 7.99E-07 Control,pressurizerheaters.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 13 of 18 Table A. I-/

FC Sce arAo Sc enar Targ efs CDF L ERF Desriio Damage to trays between containment wall and shield wall in east half of Reactor RB03 E RB Fire - East Building result in loss of SSF RCM Pump, 1.60E-07 6.97E-09 Half MD EFDWpumps, Turbine Bypass Control, pressurizerheaters, 3LP VA0001, and PORV.

• It was discovered that the Altered Events table in FRANC included circuit failure likelihood values for various power operated valves for RB03 scenarios 'B' and 'C'; these scenarios should have been identified as 'Bi' and 'Cl'. A change was made to the Altered Events table to support the delta risk calculations; however, this change has a minimal impact on the overall CDF / LERF results. It is recommended that this change be captured in a future revision to the FPRA.

LEGEND:

" FC - Fire Compartment (generally corresponds to. Fire Zone; AB098 is FZ 98 in the Aux Building)

" Scenario - unique alpha-numeric designator

" Scenario Description - ignition source

" Targets - External damage based on ignition source zone of influence

• CDF - Core damage frequency per reactor year

• LERF - Large early release frequency per reactor year Section A.2 should provide an overview of each open item that requires a change evaluation as follows:

" Discuss the variance and the nuclear safety performance criteria that are not being met

" Discuss the variance and how it will be modeled in the Fire PRA

" In tabular format provide 'Basic Event(s)' that will be 'toggled' to manifest the variant condition and the results of the FPRA manipulations

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 14 of 18 See example below:

A.2 - OPEN ITEM CHANGE EVALUATIONS Open items from calculation OSC-9292, NFPA-805 TransitionB-3 Table/Reportare identified as candidatesfor change evaluation.

The following discussionsprovide a definition of the problem statement (i.e., the variant condition)followed by a table thatprovides a summary of results. The following information is providedfor the Open Items addressed in this calculation:.

A.2.1 OPEN ITEM RB3-07-O - POTENTIAL SPURIOUS PORV OPENING IFAILURE OF BLOCK VAL VE An unallowed operatormanual action to isolatepressurizerPOR V block valve control circuitryfor HSB is credited in the safe shutdown analysis due to the postulatedfire-inducedfailure of 3RC-66 and 3RC-4 in FireArea RB3. In order to ensure RCS pressure and inventory controlfor achieving and maintaininghot standby (HSB), the safe shutdown analysis requireseither that the pressurizerPORV(3RC-66) remain closed or that the PORVBlock valve (3RC-4) remains available to be closedfrom the control room. Cablesfor both 3RC-66 and 3RC-4 are routed in close proximity in FireArea RB3. Therefore, it is assumed in the safe shutdown analysis that RC-66 could spuriously open and 3RC-4 may not be able to be closed (Case 1)from the control room.

From the FPRA [CalculationOSC-93 75, Oconee FirePRA Scenario Development, Rev. 0, also see supportingfire PRA database file], the POR V and POR V Block Valve are included in RB3 ScenariosBl and D. The basic event for spurious PORV opening' (GOOPORVDEX) is more representativeof the non-compliant case (Case 1). The compliantcase (Case 2) will assume GOOPORVDEX is notfailedfor ScenariosBR andD.

Tabe A.2-I Case / Case 2 Componen/ "Wanal Basic Evenl h9s/c Event Vardann Comp/ant ACDF/

elen AA A allon Descriion/Discuss4on ofiDF/L ClErF ER)

CDAL CDF17-ERF ERE dE)c 3RC VA0066 Close GOOPORVDEX PORV 3RC-V66 Spuriously Opens 9.66E-071 8.95E-071 7.17E-081 (Pzr PORV) due to Fire 8.13E-07 7.42E-07 7.16E-08

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 15 of 18 These valuesfall within the acceptable limits of delta CDFless than IE-06 and a delta LERF less than IE-07. The delta risk values are within the acceptable range, includingconsiderationof uncertainties,as describedin CalculationOSC-9518, ONSNFPA 805 Fire PRA Application Calculation,Revision 0. (NOTE: If the delta CDFanddelta LERFfall within a decade of the acceptance criteria,additionaldiscussion should be providedto support the conclusion and it's susceptibilityto impactsfrom uncertainties.)

L EGEND."

• Component - The component of concern that is associatedwith the variant condition/changeevaluation.

, ManualAction - For variantconditions/changeevaluations involving unallowed operatormanual actions, a briefstatement of the operatormanual actionfunction.

" Basic Event - Fire PRA Basic Event ID(s) correspondingto the component(s) andfunction ofconcern.

" Basic Event Description/Discussion- Text description associatedwith the Basic Event IDs.

Case I - VariantCDF/LERF- For the specific open item, these values represent the core damagefrequency (CDF)and large early releasefrequency (LERF) if the discrepantcondition exists in thefire area (e.g., the cable routed in the fire area that prompted the needfor the operatormanualaction is assumed to be unprotected).

Case 2 - Compliant CDF/LERF- For the specific open item, these values represent the core damagefrequency (CDF)and large early releasefrequency (LERF)if the discrepant condition was assumed NOT to exist in the fire area (e.g., the cable routed in thefire area thatprompted the needfor the operatormanual action is assumed to be protectedor routed outside of the fire area).

• ACDF/ALERF - The change in CDFand LERF between the variant condition and the compliantcondition (Case 1 - Case 2).

A.2.2 OPEN ITEM RB3-11-OE -POTENTIAL FAILURE OF SG CROSSTIE VALVE 3CCW-269 An unallowed operatormanual action to remove powerý to the SG crosstie valve 3CCW-269 is credited in the safe shutdown analysis for hot standby (HSB). Flow to either SG A or B is necessaryto ensure HSB decay heat removal. For afire in the west side FireArea RB3 emergencyfeedwater is isolated to the 3B SG via 3FDW-316 due to the potential loss of all 3B SG instrumentation. It is conservatively assumed that spurious opening of SG Crosstie Valve 3CCW-269 will divertflow from SG A to isolatedSG B andfail motor driven EFW Pump A due to a potentialpump runout (basedon lack of documented analysis). Therefore, based on the conservative assumptions in the safe shutdown analysis, a complete loss of EFW to both SGs could occur due to the interactions describedabove.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 16 of 18 Spurious opening of 3CCW-269 is not explicitly modeled in the FPRA. In order to address the potentialfor pump runout the spurious opening of 3CCW-269 was modeled in the FPRA asfailure ofboth motor driven EFWpumps and the turbine driven EFWpump.

3CCW-269 is only impacted in RB03 Scenarios B] and D.

From the FPRA [CalculationOSC-9375, Oconee Fire PRA Scenario Development, Rev. 0 (also see supportingfire PRA databasefile, archived with this calculation)], CCW-269 are included in RB03 Scenarios B] and D. The spurious opening of 269 was conservatively modeled and are removedfailure of EFW events, where they are linked to 269, in order to create the compliant case.

The non-compliant case (Case 1) will assumefailure of both MD EDFWpumps and the TD EFDWpump.

Table A.2-2 case I Case 2 Componentl A41017 n Bas/c Event Descr/,EIeontDis0uSson Var/ant Compllant ACERF CDF/LERF CDF/LERF 3CCWVA0269 Close FEFTDFPTPS While only FEFMDPAMPRis 9.66E-071 9.29E-071 3.73E-081 FEFMDPAMPR relevant; all 5 events (including 8.14E-07 7.77E-07 3.70E-08 failure to open, transfers closed, FEFMDPBMPR and failure of the remaining EFDW NCWO269MVT pumps) are toggled.

NCWO269MVO These valuesfall within the acceptable limits of delta CDFless than IE-06 anda delta LERF less than JE-07. The delta risk values are within the acceptable range, including considerationof uncertainties,as described in calculation OSC-9518, ONS NFPA 805 Fire PRA Application Calculation,Revision 0.. (NOTE: If the delta CDFand delta LERFfall within a decade of the acceptance criteria, additionaldiscussion should be provided to support the conclusion and it's susceptibility to impactsfrom uncertainties)

L EGEND.

• Component - The component of concern that is associatedwith the variantcondition/changeevaluation.

• ManualAction - For variantconditions/change evaluations involving unallowed operatormanual actions, a briefstatement of the operatormanual actionfunction.

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure 1 Page 17 of 18

• Basic Event - Fire PRA Basic Event ID(s) correspondingto the-component(s) andfunction of concern.

• Basic Event Description/Discussion- Text description associatedwith the Basic Event IDs.

• Case 1 - Variant CDF/LERF- Forthe specific open item, these values representthe core damagefrequency (CDF)and large early releasefrequency (LERF)if the discrepantcondition exists in the fire area (e.g., the cable routed in the fire area that prompted the needfor the operatormanualaction is assumed to be unprotected).

• Case 2 - CompliantCDF/LERF- For the specific open item, these values representthe core damagefrequency (CDF)and large early releasefrequency (LERF) if the discrepantcondition was assumed NOT to exist in the fire area (e.g., the cable routed in the fire area thatprompted the needfor the operatormanual action is assumed to be protected or routed outside of the fire area).

• ACDF/ALERF - The change in CDFand LERF between the variantcondition and the compliant condition (Case I - Case 2).

A.2.3 OPEN ITEM RB3-33 POTENTIAL FAILURE OF FEEDWA TER VALVE 3FDW-347 The safe shutdown analysis requiresthat emergencyfeedwaterflow be establishedto -bothsteam generators to ensure that the requirements of the inventory and pressurecontrol nuclear safety performance criterion are maintained during the transitionfrom hot standby to cold shutdown. The spurious closing of 3FDW-347 results in the isolation of the credited emergencyfeedwaterfiowpath to the B steam generatorand the inability to cool down the RCS. The inability to cool down the RCS results in water solid conditions in the reactor coolantsystem, and lifting of the PressurizerPORV (if available)or the Pressurizercode safety reliefvalves. The lifting of the Pressurizerreliefvalves challenges containmenthabitabilityand the ability to perform requiredcold shutdown manualactions within containment. An unallowed operatormanual action to remove power to 3FDW-347 toprecludefire effects is credited in the safe shutdown analysis.

Flow to the steam generatorB is required to allow transitionfrom hot standby conditions to cold shutdown. The FPRA does not model the risk impactsfor this transitionperiod. Hot and stable conditions can be achievedutilizing the creditedA steam generator.

Based on the FPRA success criteria,this assessment concludes that the impact of the manual action is not risk significant and no quantificationof delta risk is required. Therefore, the delta risk associatedwith the operatormanual action is negligible and is characterizedas epsilon (e).

Calculation No. OSC-9314 Revision No: 0 Applicable Units: Oconee Units 1, 2, and 3 Enclosure I Page 18 of 18 Table A. 2-?

Case I Case 2 Componeni Ma/nua Basic Evern' Basi'c' Even/'~ VaAn CVF/

ApA MAatlo Descri-,0on/Discussion VGD/laanf CompFlint ALERF MDF/L ERF CDFIL ERF 3FDW-347 Open NIA NIA ne de These valuesfall within the acceptable limits of delta CDF less than JE-06 and a delta LERF less than 1E-07. The delta risk values are within the acceptablerange, including considerationof uncertainties,as describedin calculation 0SC-9518, ONS NFPA 805 Fire )

PRA Application Calculation,Revision 0. (NOTE: If the delta CDF and delta LERY fall within a decade of the acceptance criteria, additionaldiscussion should be provided to support the conclusion and it's susceptibilityto impactsfrom uncertainties.)

Z EGEAtD.",

• Component - The component of concern that is associatedwith the variant condition/change evaluation.

• Manual Action - For variantconditions/changeevaluations involving unallowed operatormanual actions, a briefstatement of the operatormanual actionfunction.

• Basic Event - FirePRA Basic Event ID(s) correspondingto the component(s) andfunction ofconcern.

• Basic Event Description/Discussion- Text description associatedwith the Basic Event IDs.

. Case I - Variant CDF/LERF- For the specific open item, these values representthe core damagefrequency (CDF)andlarge early releasefrequency (LERF)if the discrepant condition exists in thefire area (e.g., the cable routed in thefire area that promptedthe needfor the operatormanual action is assumed to be unprotected).

• Case 2 - Compliant CDF/LE)F- Forthe specific open item, these values representthe core damagefrequency (CDF)and large early releasefrequency (LERF)if the discrepantcondition was assumedNOT to exist in the fire area (e.g., the cable routed in the fire area thatprompted the needfor the operatormanual action is assumed to be protected or routed outside of the fire area).

• ACDF/ALERF- The change in CDFand LERF between the variant condition and the compliant condition (Case I - Case 2).