Text

Technical Adequacy of the South Texas Project Probabilistic Risk Assessment
	ML043070448
Person / Time
Site:	South Texas
Issue date:	10/28/2004
From:	Jordan T; Southern Nuclear Operating Co
To:	; Document Control Desk, Office of Nuclear Reactor Regulation
References
	NOC-AE-04001813, TAC MC3923, TAC MC3924
	Download: ML043070448 (57)
	v • d • e

Nuclear Operating Company Ed.G South Teo/ed EkdrlC nCf1?&Staton PO9Box289AMpwidth Te77483 AA October 28, 2004 NOC-AE-04001813 U. S. Nuclear Regulatory Commission Attention: Document Control Desk One White Flint North 11555 Rockville Pike Rockville, MD 20852 South Texas Project Units 1 and 2 Docket Nos. STN 50498 and STN 50499 Technical Adequacy of the South Texas Project Probabilistic Risk Assessment

Reference:

Letter, T.J. Jordan to NRC, "Broad-Scope Risk-Informed Technical Specification Amendment Request," dated August 2, 2004 (NOC-AE-04001666)

TAC Nos. MC3923 and MC3924 The referenced letter proposed to implement a risk-informed process for determining allowed outage times for South Texas Project (STP) Technical Specifications (TS). The risk-informed process involves the application of the STP Configuration Risk Management Program (CRMP),

which is the same procedurally controlled program utilized by STP Nuclear Operating Company (STPNOC) for the implementation of 10CFR50.65(a)(4). STPNOC proposed the change as a pilot plant for the industry Risk-Informed Technical Specifications (RITS) and for evaluation of Regulatory Guide (RG) 1.200.

To support the NRC review of the referenced license amendment request, STPNOC hereby submits an analysis of the STP Probabilistic Risk Assessment (PRA) conducted in accordance with RG 1.200, "An Approach for Determining the Technical Adequacy of Probabilistic Risk Assessment Results for Risk-Informed Activities." STPNOC believes the analysis confirms that the quality of the STP PRA is sufficient to provide confidence in the results such that the PRA can be used in regulatory decision making.

04001813 (RG1200).doc STI: 31801983 IRP1

NOC-AE-04001813 Page 2 of 3 If there are any questions regarding this submittal, please contact Wayne Harrison at (361) 972-7298 or me at (361) 972-7902.

I declare under penalty of perjury that the foregoing is true and correct.

Executed on October2g, 2004 Vice President Engineering & Technical Services Awh Attachments:

1. General Description

2. Plant Changes that Have Not Been Incorporated

3. Conformance to Standards

4. Key Assumptions and Approximations

5. Resolution of Peer Review Comments

NOC-AE-04001813 Page 3 of 3 cc:

(paper copy) (electronic copy)

Bruce S. Mallett A. H. Gutterman, Esquire Regional Administrator, Region IV Morgan, Lewis & Bockius LLP U. S. Nuclear Regulatory Commission 611 Ryan Plaza Drive, Suite 400 J. J. Nesrsta Arlington, Texas 76011-8064 City Public Service U. S. Nuclear Regulatory Commission David H. Jaffe Attention: Document Control Desk U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike R. L. Balcom Rockville, MD 20852 Texas Genco, LP Richard A. Ratliff C. A. Johnson Bureau of Radiation Control AEP Texas Central Company Texas Department of State Health Services 1100 West 49th Street Jon C. Wood Austin, TX 78756-3189 Cox Smith Matthews Jeffrey Cruz C. Kirksey U. S. Nuclear Regulatory Commission City of Austin P. 0. Box 289, Mail Code: MN116 Wadsworth, TX 77483 R. K. Temple City Public Service C. M. Canady City of Austin Electric Utility Department 721 Barton Springs Road Austin, TX 78704

Attachment I NOC-AE-04001813 Attachment 1 General Description

Attachment I NOC-AE-04001813 Page I of 6 Regulatory Guide 1.200, PRA Quality Pilot Risk-Informed Technical Specifications, Initiative 4B (RITS 4B)

Whole Plant Configuration Risk Management Pilot Introduction The purpose of this document is to facilitate NRC review of the adequacy of a Probabilistic Risk Assessment (PRA) for a risk-informed Technical Specification change. South Texas Project (STP) is a pilot plant for Industry Initiative 4B for Regulatory Guide (RG) 1.200, "An Approach for Determining the Technical Adequacy of PRA Results for Risk-Informed Activities."

Specifically, STP Nuclear Operating Company (STPNOC) has developed risk management methods that allow the use of PRA technology in determining the risk associated with multiple components being removed from service concurrently. The technical approach uses a Configuration Risk Management Program (CRMP) in which the impact of equipment out of service is assessed in terms of core damage frequency (CDF) and/or large early release frequency (LERF). The integrated impact of multiple components being out of service is calculated in terms of cumulative risk to determine allowable outage times (AOTs) for a configuration within the constraints of predetermined risk thresholds. This document follows a format similar to that of RG 1.200.

The STP Risk-Informed Technical Specifications (RITS) application further extends the STP CRMP relative to Technical Specifications by establishing a configuration risk basis to Technical Specifications AOTs as opposed to system-based AOTs. This concept applies the same configuration risk management principles currently used at STP for IOCFR50.65(a)(4) of the Maintenance Rule. The STP PRA has features that facilitate the ability to perform on-line configuration risk management. Additionally, to support risk-informed applications, the STP risk models employ extensive use of software macros to simulate the station's operational maintenance practices, such that combinations of equipment removed from service can be quantified and stored in a knowledge base. This knowledge base is then accessed by a special software program that provides on-shift Operations crews with the ability to assess risk from changing plant configurations. "Configurations" as used in this submittal means equipment removed from service or otherwise declared inoperable that is within the scope of the CRMP.

The model is quantified using the RISKMAN( software code that complies with station and industry software quality assurance requirements.

Description of the STP PRA The scope of a PRA is defined by the challenges included in the analysis and the level of the analysis performed. Specifically, the scope is defined in terms of:

the metrics used in characterizing the risk,

the plant operating states for which the risk is to be evaluated, and

the types of initiating events that can potentially challenge and disrupt the normal operation of the plant.

Attachment I NOC-AE-04001813 Page 2 of 6 The metrics used for risk characterization in the STP PRA are CDF and LERF. As each technical element of the PRA is performed, the sources of uncertainty are identified and analyzed such that their impacts are understood at this level and on the risk results (CDF and LERF). The risk perspective is based on the total risk connected with the operation of the reactor The STP PRA is a full-scope Level 1 /2 PRA that incorporates internal events, (fires and floods), and external events (seismic, fire, flood). STP's PRA features a seismic PRA, flood PRA (including spatial interaction analysis), human reliability analysis, and detailed common cause modeling. The PRA is maintained and updated under a PRA configuration control program in accordance with station procedures. Periodic reviews are conducted and updates are performed, if necessary, for plant changes (including performance data, procedures, and modifications). The reviews and updates are performed by qualified personnel with independent reviews and approvals.

STPNOC has used the PRA for risk-informed insights and applications since the mid-1980s.

The NRC has previously reviewed the STP PRA in support of approving the following risk-informed licensing applications:

1. Amendment Nos. 59 & 47, dated February 17, 1994, extended the AOTs for ten LCOs and the intervals for 3 surveillance tests.

2. Amendment Nos. 85 & 72, dated October 31, 1996, extended the AOT for the standby diesel generators and their associated support systems.

3. Amendment Nos. 125 & 113, dated September 26, 2000,relaxed LCO requirements for control room and fuel handling building HVAC.

4. Approval of Exemption to Special Treatment Requirements, dated August 3, 2001, relaxed regulatory requirements for various degrees of special treatment provisions for safety related components (Option 2 Pilot).

5. Amendment Nos. 135 & 124, dated January 10, 2002, extended the AOT for ECCS Accumulators consistent with WCAP-15049-A and relaxed accumulator surveillance requirements consistent with Westinghouse Improved Technical Specifications.

6. Amendment Nos. 143 & 131, dated September 17, 2002, allowed a one-time extension of the integrated leak rate test to 15 years.

7. Amendment Nos. 146 & 134, dated December 31, 2002, extended the AOT for auxiliary feedwater.

8. Amendment Nos. 158 and 146 dated December 2, 2003, eliminated the turbine missile design basis.

9. Amendment No. 149 for STP Unit 2 dated December 30,2003, permitted a one-time extension of the AOT for standby diesel generator SDG 22 to 113 days.

In addition to the risk-informed licensing applications above, STPNOC has used the STP PRA to provide additional insight to other license amendments and to respond to NRC questions.

The following references are evaluations of the STP PRA that have been performed by the NRC and others:

Attachment I NOC-AE-04001813 Page 3 of 6

1. NRC SER related to the STP Probabilistic Safety Assessment, dated January 21, 1992, documented favorable conclusions with regard to the STP PRA, including its treatment of fire (done to support the review for Amendment Nos. 59 & 47, above).

2. 2002 Peer Review In April 2002, STP's PRA underwent an industry peer review performed in accordance with NEI-00-02, "Industry PRA Peer Review Process." All technical elements within the scope of the peer review were graded as sufficient to support application requiring the capabilities of a grade 2 (e.g., risk ranking applications). Most of the elements were further graded as sufficient to support application requiring the capabilities defined for grade 3 (e.g., risk-informed applications supported by deterministic insights). The general assessment of the peer reviewers was that STP's PRA could effectively be used to support applications involving risk significance determinations supported by deterministic analyses once the items noted in the element summaries and Fact & Observations (F&O) sheets were addressed.

Using STP's Corrective Action program as a tracking mechanism, with two major exceptions, all F&O items identified by the peer team have been completed and are incorporated as appropriate into the latest revision of the STP PRA (Revision 4). The STP PRA Revision 4 model is the basis for this application of Risk-Informed Technical Specifications. The two major exceptions that are not included in the current PRA are Level 2 model update for F&O items and reevaluation of internal flood modeling. The Level 2 update for F&O items is currently being performed with contractor assistance and will be complete by the end of 2004. The internal flood reevaluation is in progress and will be finished prior to the end of 2004. No issues have been identified from the flood reevaluation to date that affect the PRA. Attachment 5 provides additional information on the Peer Review.

RG 1.200 Required Information Identification of Parts of the PRA Used to Support RITS 4B Because the STP RITS 4B pilot application is a whole plant approach to configuration risk management, all SSCs that are within the scope of the Technical Specifications and also within the scope of the CRMP are reflected in the STP PRA. Some SSCs are explicitly modeled (safety injection pumps, standby diesel generators, etc.) while others are implicitly modeled (piping supports, snubbers, etc.). A listing of components explicitly modeled is available in the archival documentation supporting this license amendment. Thus, all technical elements of the STP PRA are used to support the RITS 4B pilot effort.

Demonstration of Technical Adequacy of the STP PRA There are two aspects to demonstrating the technical adequacy of the parts of the PRA to support an application. The first aspect is the assurance that the parts of the PRA used in the application have been performed in a technically correct manner.The second aspect is the assurance that the assumptions and approximations used in developing the PRA are appropriate.

Attachment I NOC-AE-04001813 Page 4 of 6 The technical adequacy of the STP PRA is ensured by the application of station procedural controls related to maintenance/upgrades, qualification of users, and software quality assurance.

The procedural controls delineate the requirements for the scope, frequency, and approval of PRA updates. This ensures that the as-built, as-operated station is reflected in the PRA. These PRA program procedures/processes incorporate requirements and guidance for

periodic reviews and updates, if necessary

incorporation of plant physical changes and operational performance changes (including performance data, procedures, and modifications) that impact significant accident sequences

qualification of personnel performing PRA analyses

review and approval process for PRA evaluations The PRA configuration control procedures are included in the archival documentation supporting this application.

STPNOC Submittal Documentation In accordance with Regulatory Position C.4.2 of Regulatory Guide 1.200, the information described below is being provided to demonstrate that the parts of the STP PRA are of sufficient quality to support the analyses used in the STP RITS application.

-S I ' ! ' *6 * - It' I I Identification of permanent plant changes (such as design or Attachment 2 operational practices) that have an impact on those things modeled in the PRA but have not been incorporated in the Plant Changes that Have Not baseline PRA model. Been Incorporated Documentation that the parts of the PRA required to produce Attachment 3 the results used in the decision are performed consistently with the standard as endorsed in the appendices of this regulatory Conformance to Standards guide.

Identification of the key assumptionss and approximations Attachment 4 relevant to the results used in the decision-making process.

Key Assumptions and Approximations A discussion of the resolution of the peer review comments Attachment 5 that are applicable to the parts of the PRA required for the application. Resolution of Peer Review Comments

Attachment I NOC-AE-04001813 Page 5 of 6 STPNOC Archival Documentation In accordance with Regulatory Position C.4.1 of RG 1.200, STPNOC has retained archival documentation relevant to the PRA and its application to this application. The archival documentation is not provided in this submittal but has been collected in a form that may be reviewed by the NRC at their convenience.

The archival documentation includes a detailed description of the process used to determine the adequacy of the PRA. The documentation maintained is legible, retrievable (i.e., traceable), and of sufficient detail for the staff review of the bases supporting the results used in the application.

The archival documentation associated with this specific application includes enough information to demonstrate that the scope of the base PRA is sufficient with respect to:

The plant design, configuration, and operational practices,

The acceptance guidelines and method of comparison,

The scope of the risk assessment in terms of initiating events and operating modes modeled,

The parts of the PRA required to provide the results needed to support comparison with the acceptance guidelines,

. The description of the process for maintenance, update, and control of the PRA.

A full discussion of the PRA technical elements listed below is provided in the archival documentation.

Level 1 Technical Elements

Initiating event analysis

Success criteria analysis

. Systems analysis

. Parameter estimation analysis

. Accident sequences analysis

Human reliability analysis

Quantification

Interpretation of results Level 2 Technical Elements

Plant damage state analysis

Accident progression analysis

Quantification

Interpretation of results NOC-AE-04001813 Page 6 of 6 The following archival documentation is included in the information available to the NRC staff in order to facilitate review of this risk-informed pilot application.

Initiating Events Analysis Notebook

A Systems Analysis Notebook for each system in the PRA

Event Sequence Diagrams and Descriptions

Accident Progression Analysis

Plant Level Event Tree with associated documentation

Data Analysis Notebook

Human Reliability Analysis Notebook

Level 2 Accident Sequences Notebook

Level 2 Containment Event Tree Notebook

Results and quantification information

Uncertainty Analysis

Importance Reports

Description of the process for maintenance, update, and control of the PRA The scope of risk contributors addressed by the STP PRA model for supporting the RITS 4B pilot application is provided in the archival documentation. The level of modeling that is required to support the RITS 4B application requires all initiating events applicable to the STP Level 1/2 at-power (Modes 1,2,& 3) internal/external event PRA that are contained in current requirements documents to be included.

The archival information also includes supporting information such as piping and instrumentation diagrams, electrical one-lines, logic diagrams, the STP Individual Plant Examination, etc.

NOC-AE-04001813 Attachment 2 Plant Changes that Have Not Been Incorporated NOC-AE-04001813 Page 1 of 2 The following permanent plant changes have not been incorporated into the PRA:

Instrument Air System Modification - The PRA models the instrument air system in support of "Smoke Purge" operation of the Electrical Auxiliary Building (EAB) or Control Room (CR)

HVAC systems. The Smoke Purge function is used in the event of a loss of essential chilled water to the CR or EAB HVAC systems. Smoke Purge allows once-through cooling of equipment in the CR or EAB using outside air. The STP design does not rely on the instrument air system to perform safety-related functions.

The previous instrument air system contained two reciprocating air compressors in each unit with a manually operated crosstie to the service air system for each unit. The service air system also contained two reciprocating air compressors. In the event of low service air header pressure, the crosstie automatically closed to maintain the instrument air function. One of the two air compressors in each unit was backed up by the associated unit's Balance of Plant diesel generator. In the event of loss of offsite power, operator action to start the diesel-backed air compressor allowed use of the instrument air for non-safety loads.

The modification to the instrument air system replaces all four reciprocating air compressors with centrifugal compressors. Each compressor is tied into a filtration, drier, and receiver distribution network. The discharge of the receiver splits into the old instrument air/service air headers. The service air system is still isolated on low header pressure. One compressor is capable of supplying normal instrument air loads. One of the four compressors is air-cooled rather than water-cooled and is supplied power from the Balance of Plant diesel generator if offsite power is lost.

The reliability of the new instrument air system after the modification is equal to or somewhat better than the previous instrument air system, with the additional benefit of removing a cooling water dependency for the diesel-backed air compressor. Core damage frequency is expected to remain unaffected or to decrease slightly when this modification is incorporated in the next PRA model update.

Energize to Actuate Modification - STP has installed a modification to the feedwater isolation valves that changes the operation of the valves from "de-energize to actuate" to "energize to actuate." This modification reduces the likelihood of inadvertent operation of a feedwater isolation valve from ancillary equipment failures (e.g., solenoid valves, actuation relays). This feedwater isolation function is not currently modeled in the PRA, but will be included in the next model update.

An extension of this modification is planned for the next refueling outage in each unit that will change the operation of the main steam isolation valves (MSIVs) from "de-energize to actuate" to "energize to actuate". Currently, each MSIV receives a steam line isolation signal from the solid state protection system (SSPS) actuation trains A and B. There are two safety-related solenoid isolation valves in series in the air supply to each MSIV and two safety-related solenoid air dump valves in parallel for each MSIV. With the previous design, failure of a single solenoid valve could result in MSIV closure.Loss of power to Train A or Train B Class 1E DC would result in closure of all MSIVs. The effect of this modification has been analyzed using the PRA NOC-AE-04001813 Page 2 of 2 model and resulted in no significant change in CDF or LERF. This modification will be included in the next PRA model update.

SSPS Bypass Modification - A modification has been installed in both units that allows bypassing individual instrument channels and logic channels for testing. Previously, during testing of the SSPS, input test signals resulted in making up one of four (usually) actuation logic signals. This resulted in the input logic shifting from two-out-of-four to one-out-of-three. The SSPS bypass modification allows the input signals to be bypassed rather than tripped. This results in the input logic becoming two-out-of-three. This modification has been analyzed using the STP PRA and resulted in no significant change in CDF or LERF. It will be included in the next PRA model update.

NOC-AE-04001813 Attachment 3 Conformance to Standards NOC-AE-04001813 Page 1 of 21 Documentation of Conformance to ASME Standard The STP PRA has undergone self-assessments and industry peer reviews to determine its level of compliance with existing industry standards and guidance. Each technical element has been assessed based on guidance in the ASME Standard for Probabilistic Risk Assessment for Nuclear Power Plant Applications, ASME RA-S-2002; the Westinghouse Owner's Group (WOG) Peer Review process; and the guidance contained in RG 1.200. The results of the peer review are available for inspection in the archival documentation and other assessments are included in electronic format as attachments to this submittal.

The STP PRA has been evaluated against the ASME Standard and RG 1.200. The NEI self-assessment items and the results of the review against the ASME Standard are contained in Table

1. The following items are outstanding based upon the results of this review.

1. Internal Flooding Analysis - A reanalysis of internal flooding at STP is being performed as a result of WOG peer review findings and the requirements of the ASME Standard as modified by RG 1.200. The flood walk-downs are complete, flood scenarios are being developed, and flood initiating event frequencies are being developed. The results of this analysis are expected to be available for review during the NRC site visit in November. To date, no internal flood scenarios have been identified that would exceed a CDF screening criteria of IE-07.

2. Human Reliability Analysis (HRA) Sensitivity Studies - The WOG peer review and the ASME Standard identified the need for performing HRA sensitivity analysis on core damage sequences looking for possible dependent HRA actions. The results of this analysis are expected to be available for review during the NRC site visit in November.

3. HRA Update - The HRA task is being updated using new methodology available in the HRA calculator from EPRI and will be complete in support of the next model revision. The task schedule will be available for review during the team visit in November 2004.

4. Peer Review Level 2 Findings and Observation (F&O) Items - . The Level 2 update for F&O items is currently being performed with contractor assistance and are scheduled to be complete by the end of 2004.

Table 1 depicts the results of STP's self-assessment performed using RG 1.200 Tables A-1 and B4. It also includes references to the related Peer Review F&O items and remarks.

Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 2 of 21 PRA ASME Included In NEI-00-02. -STP Evaluation -:.> Peer Review Technical . SR NEI 00-02 ELEMENTS 2., - -

Element- . - -: .; . - - : - -

Initiating IE-Al Yes IE-07, IE-08, Response: Covered under peer review IE R-2, 3, 8, 9 Events IE-09, IE-10 and F&O IE-04 Initiating IE-A2 Yes IE-05, IE-07, Response: Listed Initiators were Included except Internal IE-R5, R2, R3, Events IE-09, IE-10 flood initiators - which were screened out. Note that the R9 F&O IE-01, LOCA Outside Containment and ISLOCA initiators are IE-04 combined in the STP PRA - refer to VSEOS Initiator top event. The RTRIP general transient initiator Includes operator manual reactor trips. VSEOS Initiator Includes human error basic events for failure to close MOV to isolate leak path. Loss of support system Initiators, e.g., LOECW, LOCCW, include operator failure to start standby train human error basic event based on plant Abnormal Procedures.

Initiating IE-A3 Yes IE-08, IE-09 Response: Covered under peer review IE-R8, R9 Events Initiating IE-A4 Partial IE-05, IE-07, Response: Loss of a single train of class 1E DC power (A IE-R5, R2, R3, Events IE-09, IE-10 or B) Is Included as an initiating event. Also, the CCW R9 F&O IE-01, support system Initiator Is quantified with one train In IE-04 maintenance, one train running, and the potential for failure of the standby train. Loss of a single channel of Class 1E AC power will be evaluated as a potential Initiating event based on recent plant experience.

Initiating IE-A5 Yes IE-08 Response: Table 5-2 in the Initiating Events Notebook IE-R8 Events Rev.4 shows examples of several part power trips used In the data update. However, section 3.2.1 of the notebook states that Initiating events at shutdown are not included in the at-power scope.

Initiating IE-A6 Yes IE-16 Response: Input from industry reports, other PRAs,,and IE-R7 Events knowledgeable risk personnel have ensured a complete set of initiators. In addition, extensive plant operating experience Is used to update the current set of Initiators.

Recent plant operating experience Is used to evaluate addition or removal of Initiating events, e.g., loss of vital 120VAC, energize-to-actuate modification effect on loss of 1E DC. Specific operations personnel Interviews have not been used to identify potential Initiators.

Initiating IE-A7 Yes IE-16,JE-10 Response: Master logic diagram category MLD-17 IE-R7 F&O IE-Events 'General Indirect Initiators" provides for an evaluation of 04 precursor events. In addition, the support system FMEA was used to help Identify support system precursor failures.

Reference Rev.4 IE notebook.

Initiating IE-AB Yes IE-10 Response: Covered under peer review F&O IE-04 Events Initiating IE-A9 Yes IE-05, IE-10 Response: Covered under peer review IE-R5 F&O IE-Events 01, IE-04 Initiating IE-A10 Yes IE-06 Response: N/A IE-R12 Events Initiating IE-B1 Yes IE-04, AS-04 Response: Covered under peer review IE-R3, AS-R1 Events F&O AS-01 Initiating IE-B2 Yes IE-04, IE-07 Response: Covered under peer review IE-R3, R2 Events Initiating IE-B3 Yes IE-04, IE-12 Response: N/A IE-R3, R4 Events Initiating IE-B4 Yes IE-04 Response: Covered under peer review IE-R3 Events Initiating IE-Cl Yes IE-13, IE-15, Response: Each of the STP support system Initiators IE-R10, R13, Events IE-16, IE-17 (LOEAB, LOCR, LOECW, LOCCW, L1DC) credit an R7, R14, F&O operator action In the Initiating event frequency calculation. IE-04 The associated System Notebooks do not contain specific justification of this credit but reference the appropriate Abnormal Plant Response procedure.

Initiating IE-C2 Yes IE-13, IE-16 Response: Covered under peer review IE-R1O, R7 Events I I I Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 3 of 21 PRA--- ASME -Included In NEI-00-02 -. STP Evaluation Peer Review,

-Technical SR . NEI 0-02ELEMENTS,,

' L .

,Ele rn t,;.;-.;

Initiating IE-C3 No Response: STP Initiating event frequencies contained In the Events PRA model are based on per calendar year. The historical plant availability factor defined In top event GENST Is used by the PMET event tree to ensure the quantification accounts for the fraction of time the plant is at-power. Refer to section 5.0 of the IE notebook Rev.4.

Initiating IE-C4 No Response: Initiating event screening basis is provided In Events Table 3.4-1 IE Notebook Rev.4. Although the specific criteria listed In ASME IE-C4 requirement Is not used in the STP PRA screening documentation, the documented basis In STPs PRA Is correct and meets the Intent of this requirement. Most screened Initiating events are subsumed Ina different quantified IE category.

Initiating IE-C5 No req. for N/A Response: N/A Events Cat II Initiating IE-C6 Yes IE-15, IE-17 Response: The support system Initiator fault tree analyses IE-R13, R14 Events have been developed similar to the mitigating system top F&O IE-04 event fault trees, except for the appropriate change In mission time and meet the appropriate systems analysis requirements.

Initiating IE-C7 No Response: Initiator fault tree models use an appropriate Events mission time of 8760 hours0.101 days 2.433 hours 0.0145 weeks 0.00333 months to establish an annual event frequency.

Initiating IE-C8 No Response: The fault tree Initiators meet this requirement.

Events I Initiating IE-C9 Yes IE-15, IE-16 Response: The HEPs used In the support system Initiator IE-R13, R7 Events fault trees have been developed consistent with the HRA.

Initiating IE-ClO Yes IE-13 Response: Covered under peer review IE-RIO Events Initiating IE-Cll Yes IE-12, IE-13, Response: The Excessive LOCA 1E frequency (IELOCA) IE-R4, R10, Events IE-15 was based on expert judgement developed In the 1980s, R13 although documentation could not be found that provides a basis for the value. The value should be compared to generic data sources (if available) and a basis documented.

[CR 04-13754-1-11 Initiating IE-C12 Yes IE-14 Response: The ISLOCA - VSEQS notebook contains the IE-R6 F&O IE-Events plant features used to determine the frequency as 02IE-03 described In the ASME standard Initiating IE-D1 Partial IE-18,JE-19 Response: STP documentation meeting these requirements lE-R11F&O IE-Events are contained In the lE notebook Rev.4 02 Initiating 1E-D2 Partial IE-09, IE-20 Response: STP documentation meeting these requirements IE-R9 Events are contained In the IE notebook Rev.4 Initiating IE-D3 Partial IE-09, IE-18, Response: STP documentation meeting these requirements IE-R9, Ri 1 Events IE-19 are contained in the IE notebook Rev.4 F&O IE-02 Initiating IE-D4 Partial AS-04, DE-05, Response: N/A. 1E-D4 does not exist In ASME-RA-Sa- AS-Rl, DE-R3, Events SY-21 2003. SY-R21 F&O I___AS-01, DE-05 Accident AS-Al Yes AS-04, AS-08 Response: The STP PRA Is based on the linked event tree AS-R1, R3, R13 Sequence methodology via the use of . The event trees are built from F&O AS-01, Analysis Event Sequence Diagrams (ESDs) which are based, In AS-06 part, on emergency operating procedures. The STP PRA represents the as-built, as-operated power plant.

Peer Certification Comment (R13): Documentation of the accident sequence model Including guidance, Is detailed and fairly extensive, Including the ESDs and the event trees.

Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 4 of 21

PRA ASME Included In, NEI-OP02 -'; . S, Evaluation = -Peer Review Technical SR - NEI 00-02 ELEMENTS *, , ,

Elem ent - ' W . :. * - *- K Accident AS-A2 Yes AS-O6, AS-07, Response: Each of the 50+ initiating events are grouped AS-R3, R7 Sequence AS-08, AS-09, into response event trees representing each of the following F&O AS-03, Analysis AS-17 events: AS-09, AS-10,

- General Transients SY-06, AS-06,

- Steam Generator Tube Ruptures AS-04, TH-04

- Small LOCAs

- Medium LOCAs

- Large LOCAs The key safety functions are defined In the appropriate Event Tree Notebook. All functions necessary to successfully mitigate the accident/transient are questioned.

Accident AS-A3 Yes AS-07, AS-17, Response: The system function necessary to mitigate the AS-R7, SY-R17 Sequence SY-17 Initiating event are contained within each Event Tree or F&OAS-10, Analysis System Notebook, the Success Criteria Notebook and the SY-02, SY-08

._ . model Accident AS-A4 Yes AS-19, SY-05 Response: Operator actions are defined In the Event AS-Ri2,SY-R5 Sequence Sequence Diagrams and section 3.3.4 of the Individual F&O AS-07, Analysis Plant Examination. Human reliability data was updated In SY-05 the late 90s (PRA-99-010) and this currently being updated via the HRA Calculator Accident AS-A5 Yes AS-05, AS-18, Response: The accident sequence model is based on the AS-R2., R8, Sequence AS-19, SY-05 Event Sequence Diagrams as outlined In the IPE. R12, SY-R5 Analysis F&O AS-07, l SY-O5 Accident AS-A6 Yes AS-08, AS-13, Response: Covered under peer review AS-R3, R4, RI Sequence AS.04 F&O AS-06, Analysis IAS-01 Accident AS-A7 Yes AS-04, AS-05, Response: The software creates accident sequence AS-Ri, R2, R3 Sequence AS-06, AS-07, reports. These reports can be generated for Individual F&O AS-41, Analysis AS-08, AS-O9 initiators, groups of initiators, and all Initiators. Accident AS-03, AS-09, sequences are reviewed at the end of each model update AS-10, AS-06, to verify sequences make logical sense. AS404, SY-06, TH-04 Accident AS-A8 Partial AS-20, AS-21, Response: End states In the STP PRA model for CDF are AS-R9, R10, Sequence AS-22, AS-23 defined as either successful or melt (i.e., core damage). R 1I F&O TH-Analysis End states for level 2 are defined as type of release or 01 successful containment performance (e.g., large early, late.).

Accident AS-A9 Yes AS-1 8, TH-04 Response: Success criteria are based on the UFSAR, AS-R8, TH-R5 Sequence MAAP analyses, or other special analysis (i.e., room heat-Analysis _up calculations).

Accident AS- Yes AS-04, AS-05, Response: System and operator response for each Initiator AS-RI, R2, R3, Sequence A10 AS-06, AS-07, Is explicitly modeled In the STP PRA event trees or system R12, SY-R5, Analysis AS-08, AS-49, analysis. HR-R6 F&O AS-19, SY-05, AS-01, AS403, SY-08, HR-23 AS-09, AS-1 0, AS-06, AS-04, SY-06, SY-05, TH-04 Accident AS- Yes AS-08, AS-10, Response: In the software, event trees are linked for each AS-R3, DE-R4 Sequence All AS-15, DE-O6, initiator. Status of the previous event tree top events Is F&O AS-06, Analysis AS Checklist maintained within the software. In addition, macros are AS-05, DE-06 Note 8 used to simplify the split fraction rules. No information Is lost by transferring from one event tree to another.

Accident AS-Bi Yes IE-44, IE-45, Response: Initiators that affect mitigating systems or IE-R3, R5, AS-Sequence IE-10, AS-04, functions are explicitly modeled within the STP PRA model. Ri, R2, R3, Analysis AS-05, AS-O6, This Is accomplished via top event boundary conditions DE-R3 F&O IE-AS-07, AS-48, and/or split fraction rules. 01, IE-04, AS-AS-49, AS-1 0, 01, AS-43, AS-AS-1i, DE-05 09, AS-10, AS-06, AS-04, AS-05, AS-02, SY-06, TH-04, DE-05 Table I - STP Self-Assessment Results NOC-AE-04001813 Page 5 of21 PRA ASME Included In NEI-00-02 SW Evaluation

.PeerReview A.

Technical SR* NEI 00 ELEMENTS. -. -

El ment* . ...

Accident AS-B2 Yes AS-10, AS-11, Response: These dependencies are documented In the DE-R2, R3, R4 Sequence DE-04, DE-05, Event Sequence Diagrams and handled in the event trees. F&O AS-05, Analysis DE-06 For example, success for Small LOCA requires high head, AS-06, AS-02, depressurization through heat removal, and low head. See DE-05, DE-06 SUCC macro In PDSSL.

Accident AS-B3 Yes AS-10, DE-10, Response: Covered under peer review DE-R9, SY-Sequence SY-1 1, TH-08 R11, TH-R2 Analysis F&O AS-05, AS-06, DE-06, SY-09, TH-02 Accident AS-B4 Yes AS-08, AS-09, Response: In the STP PRA model, all train dependent top AS-R3 F&O Sequence AS-10, AS-11 events are ordered from A to B to C. In addition, all AS-06, AS-04, Analysis conditional split fractions are calculated In the same AS-05, AS-02, manner. TH-04 Accident AS-B5 Yes AS-10, AS-11, Response: Split fraction logic rules In the STP PRA model DE-R2, R3, R4 Sequence DE-04, DE-05, accounts for the train specific dependencies. This Is F&O AS-05, Analysis DE-06, QU-25 documented In the event tree notebooks. AS-06, AS-02, DE-05, DE-06 Accident AS-B6 Yes AS-13 Response: The STP PRA model does Include time-phased AS-R4 Sequence dependencies. For example, Diesel Generator recovery Is Analysis modeled In top events OM and RE; DC battery depletion is modeled In Top Events DA, DB, DC, and DD; and Electrical Auxiliary Building room cooling Is explicitly modeled in top events FA, FB, and FC for EAB.

Accident AS-Cl Yes AS-24, AS-25 Response: A review of the top rank sequences Is F&O SY-08, Sequence performed and documented in Level 1 results notebook. TH-04 Analysis The top sequences are reviewed against the Event Sequence Diagrams to ensure the split fraction logic rules are correctly modeling the event In addition, an Informal review of all accident sequences Is performed at the end of the update process to ensure logical modeling Accident AS-C2 Yes AS-24, AS-25, Response: The treatment for each Initiator and event tree F&O SY-08, Sequence AS-26 Is documented In the Initiating Event and Event Tree TH-04 Analysis Notebooks. Specifically, the Initiator Isdefined In the former and the rules for each event tree In the later.

Accident AS-C3 Partial AS-11 , AS-17, Response: There Is no one notebook that documents all AS-R7, R9, TH-Sequence AS-20, AS-24, the Items within the check list. RS, DE-R4 Analysis DE-06, TH-05 (a) The link between Initiating event and accident sequence F&O AS-02, analysis Is contained within the STP PRA model, I.e., In the AS-06, SY-08, Initiating event dialog box of the event tree module. This TH-04, TH-05, dialog box contains a list of all the linked event trees used HR-07, SY-08, in quantifying the Initiating event DE-06 (b) The definition of Core Damage Is , the STP PRA assumes that any scenario In which the loss of core heat removal progressed beyond the point of core uncovery, and core exit temperatures exceeded 1,200F, Is a core damage scenario (documented In the Level 1 Results notebook). See response to Success Criteria Task (para.4.5.3) for more Information on the relationship of success criteria to core damage (c) See Human Reliability section for more Information on traceability of HRA (d) The STP PRA models sequences to success, any sequence not mapped to success is mapped to melt. The event tree notebooks contain more Information on how success Is defined (via the macro SUCC In the PDS event trees)

(e) Documentation for Integrated treatment In various notebooks and within the model Table I - STP Self-Assessment Results NOC-AE-04001813 Page 6 of 21

-PRA ASME Included In NEI-00-02 STP Evaluation .:. Peer Review TechnicalI SR 'NEI 00-02 ELEMENTS :,- . -

Accident AS-C4 Partial AS-11, AS-24 Response: There Is no one notebook that documents all F&O AS-02, Sequence the items within the check list AS-06, SY-08, Analysis (a) success criteria Is contained within various documents, TH-04 Including the system and event tree notebooks (b) there Is only one model, which can quantify both level 1 and 2 results. All initiating events are included within this model (c) the event sequence diagrams documented in the IPE describes the progression of each class of initiators (e.g.,

small break LOCA)

(d) the event sequence diagrams contain assumptions however, the impact of these assumptions are not specifically described In the ESDs (e) analysis/calculations are contained within the system notebooks (e.g., reference to room heat up calculations),

level 2 accident progression notebook, and event sequence diagrams within the IPE.

(Q)operation Information Is contained within the event sequence diagram and system notebooks.

(g) see system notebooks for equipment operation (e.g.,

PDP operation within the CVCS system notebook)

(h) for the most part, the STP model does not model systems under a single top event. There are some exception like RHR pump (OC) and heat exchanger (RX) and these are documented within the

_ _____ _ _ _ _ _ _ system notebook_ _ _ _ _ _ _

Success SC-Al Yes AS-20, AS-22, Response: CDF is defined In the Level 1 Quantification AS-R9, R10 Criteria AS Footnote 4 Notebook, along with reference to its basis. (F&O TH-01 F&O TH-01 Peer Review) Definition: The PRA assumes that any scenario In which the loss of core heat removal progressed beyond the point of core uncovery, and core exit temperatures exceeded 1,200 0F, Is a core damage scenario.

Success SC-A2 Yes AS-22, TH-04, Response: See Level 1 Quantification Notebook definition AS-1i0, TH-R5, Criteria TH-05, TH-07, of CDF. Additional information resides in the Level 1 R6,R3 F&O AS Footnote 4 Thermohydraulic Analysis Notebook. There Is not a single TH-O1, TH-05, location for this Information. (Known Issues from Peer TH-03, HR-07, Review TH-01, TH-04, TH-05, TH-06) SY-O8 Success SC-A3 Yes AS-06, AS-07, Response: See Event Sequence Diagrams and PRA AS-R7, R9 Criteria AS-17, AS-20 model results for significant accident sequences. See F&O AS-03, definition of SUCCESS In the PDS event tree macros. See AS-09, AS-10, Initiating Events Notebook, associated Event Tree SY-06 Notebooks (PDS), and Success Criteria Notebook. Table A-1no Impact Success SC-A4 Yes AS-07,AS-17, Response: This is spread throughout PRA documentation. AS-R7, R8, SY-Criteria AS-18, SY-08, Defined primarily in the event trees and ESDs, the PRA R8, R17,TH-SY-17, TH-09, event tree notebooks describe in detail the event and R8, 1E-R12, IE-06, DE-05 criteria for each of the mitigating functions, the event DE-R3 F&O sequence diagrams, and what systems are required to AS-10, SY-06, mitigate the event. The Thermohydraulic Analysis Notebook SY-02, SY-08, describes certain analyzed scenarios, which support the TH-04, TH-05, basis for the system success criteria. The System and DE-05 Success Criteria Notebooks detail what each system mitigating function is and their success criteria. The current PRA model does not share capabilities between units other than standby transformers because procedures did not exist at the time to perform such tasks. However, future model updates will incorporate these capabilities. Standby transformers are shared between units during planned maintenance, see PMET, and OFFSITE event trees and 4.16KV Electrical Power System Notebook. (Known Issues from Peer Review TH-03, TH-04, TH-05, TH-06, TH-07)

Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 7 of 21

. PRA *ASME Included In NEI-0O STP Evaluation . .f Peer Review Technical :SR .-NEI 00.02 ELEMENTS- . ' - . -

Element : - - . - - -

Success SC-A5 Partial AS-21, AS-23, Response: Mission times for systems are discussed AS-R9, Ri 1 Criteria AS-20 throughout the System Notebooks, Success Criteria Notebooks, and Level 1&2 Quantification and Results Notebooks. Mission time for most systems Is set at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . Some exceptions - batteries with no chargers (4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months based on calculations); Level 2 analysis power recovery following station blackout of 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months top event CV (table 2.2-1).( Known Issues from Peer Review TH-01, TH-02, TH-05, TH-06)

Success SC-A6 Yes AS-05, AS-18, Response: See Thermohydraulic Analysis Notebook and AS-R2, R8, Criteria AS-19. TH-04, supporting documentation for model. (Known Issues from R12, TH-R5, TH-05, TH-06, Peer Review TH-03, TH-04, TH-07) R6, R2, ST-R2, TH-08, ST-04, R3, SY-R5 ST-05, ST-07, F&0 AS-07, ST-09, SY-05 AS-04, TH-05, TH-07, TH-03, TH-02, HR-07, SY-08, SY-05, ST-01, IE-03 Success SC-B1 Yes AS-18, SY-17, Response: Covered under peer review. Table A-I AS-R8, SY-Criteria TH-04, TH-06, comments: MAAP4 code was developed and verified by R17, TH-R5, TH-07 qualified trained users. R7, R3 F&O SY-02, SY-08, TH-07, TH-03, l AS-04 Success SC-B2 No TH-04, TH-08 Response: (Use of Expert Judgment) - Not used in STP TH-R5, R2 Criteria PRA F&O TH-02 Success SC-B3 Yes AS-18.TH-04, Response: See Thermohydraulic Analysis Notebook and AS-R8, TH-R5, Criteria TH-05, TH-06, supporting documentation. (Known Issues from Peer R6, R7 F&O TH-07 Review THi-1 to 07) TH-07, TH-03, TH-05, HR-07, SY-08, AS-04 Success SC-B4 Yes AS-18, TH-04, Response: Covered under peer review. Table A-1 AS-R8, TH-R5, Criteria TH-06, TH-07 response: see SC-B13 R7, R3 F&O TH-07, TH-03, AS-04 Success SC-B5 Yes TH-09, TH-07 Response: Known Issues from Peer Review TH-07 TH-R8, R3 Criteria F&O TH-04, TH-03, TH-05, SY-08 Success SC-B6 Yes 0U-27, 0U-28 Response: See Success Criteria, Thermohydraulic QU-R9 F&O Criteria Analysis, and System Notebooks. See also PRA Analysis 0U-03 Assessments for sensitivity studies performed on the PRA model. Also see the IPE, which contains the initial analysis.

(Known Issues from Peer Review THI-1, TH-02, TH-03, TH-05, TH-06, TH-07)

Success SC-Cl Yes ST-13, SY-10, Response: See All PRA Notebooks, specifically Success ST-R1, SY-Criteria SY-17, SY-27, Criteria and Thermohydraulic Notebooks and their RIO, R17, R22, TH-08, TH-09, references. See MAAP analysis Notebooks, Design Basis TH-R2, R8, R9, TH-10, AS-17, Documents and calculations, Table A-1 response: Key AS-R7, R8 AS-18 assumptions as defined In Reg. Guide 1.200T not yet F&O SY-02, documented. SY-08, TH-02, TH-04, TH-05, QU-03 Success SC-C2 No TH-10 Response: (Document Expert Judgment) - N/A, Not used TH-R9 Criteria In STP PRA Success SC-C3 Yes AS-12, AS-13, Response: See IPE Documentation and Thermohydraulic AS-R4,TH-R8, Criteria TH-09, TH-10 analysis Notebook. (Known Issues from Peer Review TH- R9 F&O AS-06, TH-07) 08, TH-04, 05, SY-08 Success SC-C4 Partial AS-24, SY-27, Response: See PRA Notebooks, specifically Level 1 and 2 SY-R27, TH-Criteria TH-09, TH-10, Quantification, Success Criteria, Thermohydraulic Analysis, R8, R9, HR-HR-30 and IPE. (Known Issues from Peer Review TH-04) R17 F&O SY-08, TH-04, TH-05 Table I - STP Self-Assessment Results NOC-AE-04001813 Page 8 of 21

-PRA ASME Included In ; NEI-00-02, STP Evaluation Peer Review Technical SR' NEI002ELEMENTS

Elernent I -f -. '- ' ',

Systems SY-Al Yes SY-04, SY-1 9 Response: See System Notebooks SY-R4, R19 Analysis Systems SY-A2 Yes SY-05, SY-13, Response: See System Notebooks SY-R5, R13, Analysis SY-16, AS-19 Ri , AS-R12 F&O SY-05, SY-04, AS-07 Systems SY-A3 Yes SY-05, SY-06, Response: By procedure OPGP01-ZA-0305, section 4.0, all SY-R5, R6, R8, Analysis SY-08, SY-12, plant information sources used to define and establish the R12, R14 F&O SY-14 PRA must be reviewed during the model update process SY-05, SY-06, and periodically between model updates to Insure that the SY-03, AS-06, PRA represents the 'As Built' plant. See PRA Database of DA-03 Inputs. Items c through h are contained In the systems analysis notebooks. Item a (components and system boundaries needs to be developed as part of submittal.

Item b Is described in the Support System Notebook.

Systems SY-A4 Partial SY-10, DE-11, Response: Plant walk downs and Interviews were SY-R10, DE-Analysis SY Footnote 5 conducted during the initial PRA development, and are R6, RI 1 periodically conducted during the design change process between model updates when a design change Impacts the PRA and periodically during model updates. The GOA working group also reviews the PRA model and assumptions following a model update prior to risk ranking systems and components. This provides additional assurance that the system analysis correctly reflects the as-built, as-operated plant. System high level summaries, which Include components, failure modes, and assumptions, are also reviewed as part of the CRMP program.

Systems SY-A5 Partial SY-08, SY-1 1, Response: Within the STP PRA documentation of systems, SY-R8, Ri 1 Analysis OU-12, OU-13 every system model description includes those conditions QU-R2 F&O that prevent system operation and function Including both SY-09 normal and alternate alignments. See System Notebooks sections 2 and 3 Systems SY-A6 Yes SY-07, SY-08, Response: See System Notebooks SY-R7, R8, Analysis SY-1 2, SY-1 3, R12, R13, R14 SY-14 F&O AS-06, SY-03 Systems SY-A7 Yes SY-06, SY-07, Response: In the STP PRA only the AMSAC system fits SY-R6, R7, R8, Analysis SY-08, SY-09, this description and Is only used in selective sequences, R9, R19 F&O SY-19 See EPONSITE top AM. All other systems are modeled in SY-06, SY-07 detail in fault trees.

Systems SY-AB Partial SY-06, SY-09 Response: Covered under peer review. System notebooks SY-R6, R9 Analysis describe the boundaries of the systems/functions modeled F&O SY-03, In the notebook. SY-06, SY-07,

. DA-03 Systems SY-A9 Yes SY-06, SY-19, Response: See System Notebooks SY-R6, R19, Analysis OU-12, QU-13 OU-R2 F&O

_ I I_SY-06 Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 9 of 21 PRA ASME Included In NEI-00-02 .- STP Evaluation -- Peer Review Technical SR NEI 00-02 -- ELEMENTS Element Systems SY- Partial SY-09 Response: Super components are used in the STP PRA to SY-R9 F&O Analysis A10 simplify system modeling. Whenever a super component Is SY-07 used, measures are taken to ensure that only those components relative to the function being modeled are used. A typical use of super components in the STP PRA would be collecting passive components, such as manual valves, into a single basic event for a train. Most of these can be split Into Individual basic events with the new version of RISKMAN (most of the splitting has been done In Rev 4.1). There is no mixing of systems, and actuation signals are modeled separately. Super components that are made up of multiple components that have different failure probabilities are generally split Collecting component failure data at a higher level, i.e., EDG and associated auxiliaries, does not necessarily result In a super-component. The EDG system model actually splits sequencer, breaker, and engine Into separate basic events.

Super components are heavily scrutinized by the GOQA expert panel during system and component risk ranking following a model update to ensure they are modeled correctly. See System Notebooks, and

. _ GQA risk ranking process.

Systems SY- Yes SY-12, SY-13, Response: See System Notebooks SY-R12, R13, Analysis All SY-17, SY-23, R17, R23 F&O AS-10, AS-13, AS-06, AS-05,

_ AS-16, AS-17 SY-02, SY-08 Systems SY- Partial SY-06, SY-07, Response: Passive critical components whose failure SY-R6, R7, R8, Analysis A12 SY-08, SY-09, affects system operability such as heat exchangers+193 R9, R12, R13, SY-12, SY-13, and tanks are modeled In the STP PRA. Because of STPs R14 F&O SY-SY-14 design, and because piping failure rates are significantly 06, SY-07, SY-lower than other passive components which are modeled, 03 AS-06 piping Is not Included In the STP PRA system models. See System Notebooks for example Safety Injection,

._ _ .Component Cooling Water, or Auxiliary Feedwater.

Systems SY- Yes SY-15, SY-16. Response: See System Notebooks SY-R15 F&O Analysis A13 DA-04 SY-04, DA-02 Systems SY- No SY-08, HR-04, Response: See System Notebooks SY-R8, HR-R3, Analysis A14 HR-05, HR-07 R5 F&O HR-01 Systems SY- Yes SY-08, HR-04, Response: See System Notebooks SY-R8, HR-R3, Analysis A15 HR-05, HR-07 R5 F&O HR-01 Systems SY- Yes SY-08, HR-08, Response: See System Notebooks SY-R8, HR-R6 Analysis A16 HR-09, HR-10 Systems SY- Yes SY-10, SY-11, Response: The STP PRA System Notebooks address for SY-R10, Rlb, Analysis A17 SY-13, AS-13 each system the conditions that cause the system to Isolate R13, AS-R4 or trip. The Support System Model Notebook contains the F&O SY-09 direct system dependency descriptions. Though some dependencies are covered in the system analysis, most direct dependencies are evaluated in the event trees. See Event Tree Notebooks for EPONSITE and MECHSUP.

Systems SY- Yes SY-08, SY-22, Response: See PMET Event Tree for planned SY-R8, R22, Analysis A18 DA-07 unavailability, See system level unplanned unavailability in DA-R3 Individual System Notebooks, System testing frequency and surveillances are located In the individual System Notebooks.

Systems SY- Yes SY-11, SY-13, Response: Under adverse conditions, the STP PRA SY-RI1, R13, Analysis A19 SY-17, AS-18, assumes in most cases the affected systems fail. An R17, AS-R8, DE-10, TH-08 example of an exception to this rule Is EAB HVAC system DE-R9, TH-R2 calculation. This calculation established the mission time for F&O SY-08, loss of EABHVAC for affected system components. See SY-02, DE-06, EAB HVAC System Notebook Sections 2.1.6, 2.1.7, 2.4.4, TH-02 and 3.4. Actual modeling of this dependency Is performed in the event trees.

Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 10 of 21 i PRA , ASME Included In NEI-00-02 ' - - .

S.P Evaluation - - Peer Review Technical- -SR >' NEI 00-02 ELEMENTS .. .-

Element ,,. v . : ; - : .:

Systems SY- Partial SY-05, SY-1 1, Response: The STP PRA systems were developed directly SY-R5, Ri 1, Analysis A20 SY-13, SY-2, from the design basis documents and In most cases no R13, AS-R8, AS-1 9, TH-08 credit is taken beyond the rated or designed capability. For TH-R2 F&O example, the 125V batteries are credited for 4 hours4.62963e-5 days 0.00111 hours 6.613757e-6 weeks 1.522e-6 months (and SY-05, SY-09, are good for more), we credit single train success, we TH-02 exclude ventilation requirements In select areas, etc. For Level 2 analysis, equipment survivability during severe accidents is discussed in section 4.1.4 of the IPE. Also see the Level 2 Analysis notebook where probabilities are used to determine the design limits of SSCs like the containment and the associated justification. Table A-i response: no

__ impact.

Systems SY- Yes SY-18 Response: See System Notebooks SY-R18 Analysis A21 Systems SY- Yes SY-24, DA-15, Response: STP PRA models recovery actions by operators SY-R24, DA-R4 Analysis A22 OU-18 supported by actual plant data and response times. See top

._ _events starting with letter "O_ _

Systems SY-B1 Yes SY-08, DA-08, Response: See System Notebooks SY-R8, DA-R7, Analysis DA-14, DE-08, R12, DE-R7, DE-09 R8 F&O DA-01 Systems SY-B2 No req. for Response: See System Notebooks Analysis Cat II Systems SY-B3 Yes DE-08, DE-09, Response: See System Notebooks DE-R7, R8, DA-Analysis DA-10, DA-12 R9 Systems SY-B4 Yes SY-08, DA-08, Response: See System Notebooks and Data Analysis SY-R8, DA-R7, Analysis DA-1 0, DA-1 1, Notebook R9, RIO, RI1, DA-12, DA-13, R12, DE-R7, DA-14, DE-08, R8, QU-R1 DE-09, CU-09 F&O DA-01 Systems SY-B5 Yes SY-12, DE-04, Response: See Event Tree Notebooks,: System SY-R12, DE-Analysis DE-05, DE-06 Notebooks. For Maintenance dependency see PMET event R2, R3, R4 tree. See Event Sequence Diagrams. F&O AS-06, DE-05, DE-06 Systems SY-B6 Yes SY-12, SY-13 Response: Support system success criteria are established SY-R12, R13 Analysis based upon the variability in the conditions present during F&O AS-06 the postulated accidents for which the system is required to function. In most cases, UFSAR success criteria are used to establish success criteria for support systems, in other cases, plant specific analyses for unique plant conditions establish the success criteria for support systems (e.g.,

room cooling requirements).

Systems SY-B7 Yes SY-13, SY-17, Response: See System Notebooks SY-R13, R17, Analysis AS-18, TH-07, AS-R8, TH-R3, TH-08 R2 F&O SY-02, SY-08, TH-03, TH-02 Systems SY-B8 Yes SY-10, DE-11 Response: See Spatial Interactions Database, Event Tree SY-R10, DE-Analysis Notebooks, System Notebooks, External Events, Intemal R6, DE-RI 1 Fires and Floods Initiating Events. Event Sequence Diagrams.

Systems SY-B9 Yes SY-10, AS-20, Response: See Level 2 Analysis and Accident Progression SY-10, AS-R9, Analysis 12-08, 12-09, Notebooks, Containment Event Tree Notebook. Event L2-R5 F&O L2-L2-11, L2-13 Sequence Diagrams. 01, 1.202, L2-05, 12-03 Systems SY- Yes SY-12, SY-13 Response: See System Notebooks, System Description. SY-R12, R13 Analysis 810 F&O AS-06 Systems SY- Yes SY-08, SY-12, Response: Systems that are required for initiation or SY-R8, R12, Analysis Bi ISY-13 actuation of systems are specifically modeled In the STP R13 F&OAS-PRA, See ODPS, SSPS, and Reactor Trip System 06 Notebooks. These notebooks describe the conditions needed for automatic actuation along with permissives and lockouts. Event Trees EPONSITE and MECHSUP present the dependencies other systems have on the actuation systems. Event tree macros are also used to define

.... Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 11 of 21 PRA ASME Included In - NEI-O0 @ STPEvaluation Peer Review:

Technical -SR - NEI 00-02 . ELEMENTS , . . - - --

,-Element' . -.

boundary conditions for systems/trains Systems SY- Yes SY-13 Response: The STP PRA models Inventory of tanks, SY-R13 Analysis B12 battery capacity, air, power, and cooling systems. See the associated system notebooks for load and mission time capabilities.

Systems SY- No Response: Proceduralized recovery actions are modeled in Analysis B13 the PRA. Proceduralized recovery actions not eliminate a support system from the model. See applicable System Notebooks.

Systems SY- Partial DE-06, AS-06 Response: Not directly applicable at STP due to system DE-R4, F&O Analysis B14 design and system boundary definitions. Exception DE-06, AS-03, examples, CCW to RHR heat exchanger in RHR top OC, AS-09 and also in Hx top RX. LHSI pumps In Injection and recirculation - event tree rules. Support system dependency Is treated In the Event Trees PMET, EPONSITE, and MECHSUP. Within the System Notebooks, descriptions of basic event components like a common suction valve that can disable multiple trains of that system

. are discussed. See Auxiliary Feed Water or Safety Injection System Notebooks for examples.

Systems SY- Yes SY-1 1 Response: In general, no SSC is credited for operating SY-R11 F&O Analysis B15 beyond its design in the PRA without a calculation to SY-09 support the assumption (Example see SI room cooling calculation for exception). See applicable System Notebooks, Event Tree Notebooks and PRA Analysis/Assessments for operation In adverse conditions.

Systems SY- Yes SY-08 Response: Covered under peer review SY-R8 Analysis B16 Systems SY-Cl Partial SY-23, SY-25, Response: See System Notebooks SY-R23, R25, Analysis SY-26, SY-27 R26, R27 Systems SY-C2 Yes SY-05, SY-06, Response: See System Notebooks and the PRA Model SY-R5, R6, R9, Analysis SY-09, SY-27 R27 F&O SY-05, SY-06, SY-

__ 07 Systems SY-C3 Yes SY-18, SY-27 Response: See System Notebooks SY-R18, R27 Analysis Human HR-Al Yes HR-04, HR-05 Response: A pre-initiator human action analysis has been HR-R3 F&O Reliability performed and Incorporated Into the system analysis. HR-01 Analysis However, this particular analysis has not been updated since the IPE. A specific review of test and maintenance procedures was performed for the STP_1996 and STP_1997 models (all systems). A continuing review of test and maintenance procedures is a standard part of a PRA system analysis update and is performed by all analysts for their respective systems. An initiative Is being considered to review and screen testing and maintenance practices to determine If additional pre-initiator HEPs should be Incorporated in the systems analysis. [CR 04-13754 _ 11 Human HR-A2 Yes HR-04, HR-05 Response: Covered under peer review [See SSPS System HR-R3 F&O Reliability Notebook] HR-01 Analysis Human HR-A3 Yes HR-05, DE-07 Response: Covered under peer review HR-R3, DE-RS Reliability F&O HR-01 Analysis Human HR-B1 Yes HR-05, HR-06 Response: The systems analysis procedure reviews HR-R3, R4 Reliability contain the applicable screening. F&O HR-01, Table 1- STP Self-Assessment Results NOC-AE-04001813 Page 12 of 21 PRA ASME Included In NEI-00-02 .- STP Evaluation. :2: - Peer Review Technical .,SR NEI 00-02 ELEMENTS '< , .;4.. -

'Element ;-. .wN; Analysis HR-04 Human HR-B2 Partial HR-05, HR-06, Response: Covered under peer review HR-R3, R4, R5, Reliability HR-07, HR-26, R16 F&OHR-Analysis DA-05, DA-06 01, HR-04 Human HR-C1 Yes HR-27, SY-08, Response: Covered under peer review HR-R16, SY-Reliability SY-09 R8, R9 F&O Analysis HR-06, SY-07 Human HR-C2 Yes HR-07, HR-27, Response: Unscreened activity unavailability Is Included In HR-R3, R5, Reliability SY-08, SY-09 the system maintenance alignments. Example is top R16, SY-R8, R9 Analysis AFWS. See also HR-B1 F&O HR-06, SY-07 Human HR-C3 Yes HR-05, HR-27, Response: N/A HR-R3, R16, Reliability SY-08, SY-09 SY-R8, R9 Analysis F&O HR-01, HR-06, SY-07 Human HR-D1 Yes HR-06 Response: Covered under peer review. Related F&O HR- HR-R4 F&O Reliability 02. HR-04 Analysis I Human HR-D2 Yes HR-06 Response: Covered under peer review HR-R4 F&O Reliability HR-04 Analvsis Human HR-D3 No Response: The STP dynamic and recovery HEP Reliability development includes performance shaping factors (PSFs)

Analysis that meet this requirement. Due to lack of documentation, it can not be determined If these PSFs were evaluated for pre-initiator HEPs. [CR 04-13754-2-21 Human HR-D4 No Response: From the available STP pre-initiator HEP Reliability documentation (IPE section 3.3.4.3), It does not appear that Analysis STP credited recovery of pre-Initiator errors during development of a particular pre-Initiator HEP, as allowed for In THERP. The available documentation Is lacking In the details of the pre-initiator HEP development. [CR 04-13754-2-21 Human HR-D5 Yes HR-26, HR-27, Response: Covered under peer review HR-R1i6, DE-R5 Reliability DE-07 F&O HR-06 Analysis Human HR-D6 No Response: Developed HEPs are typically a log normal Reliability distribution with associated range factor. The mean values Analysis _ are used in the PRA Level 1 and 2 quantifications. '

Human HR-D7 No Response: When using HEPs In the PRA, analysts judge Reliability the reasonableness of the values prior to use In the models.

Analysis This reasonability check is Inherent In the process, but not well documented. [CR 04-13754-2-3]

Human HR-El Yes HR-09, HR-10, Response: Covered under peer review HR-R6, RIO.

Reliability HR-16, AS-19, AS-R12, SY-R5 Analysis SY-05 F&O HR-04,

__ AS-07, SY-05 Human HR-E2 Yes HR-08, HR-09, Response: Covered under peer review HR-R6, R14, Reliability HR-10, HR-21, R15 F&O HR-Analysis HR-22, HR-23, 04 HR-25 Human HR-E3 Partial HR-10, HR-14, Response: This supporting requirement Is met during the HR-R6, R9, Reliability HR-20 operator Interview process. R13 F&O HR-Analysis 07 Human HR-E4 Partial HR-14, HR-16 Response: This supporting requirement Is met during the HR-R9, R10 Reliability operator interview process. F&O HR-07, Analysis HR-04 Human HR-F1 Yes HR-16, AS-19, Response: Covered under peer review HR-RIO, AS-Reliability SY-05 R12, SY-05 Analysis F&O HR-04, AS-07, SY-05 Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 13 of 21 PRA ASME Included In : NEI-00 -STP Evaluation - Peer Review

-Technical SR 'NEI 00-02 ELEMENTS :

Element .--

Human HR-F2 Partial HR-.1, HR-16, Response: The HEPs developed for dynamic human HR-R7, HR-Reliability HR-17, HR-19, actions Include scenario sheets that define the HFE. The R10, HR-R.1, Analysis HR-20, AS-19, Items Included are 1) scenario description, 2) high level HR-R12, HR-SY-05 specific tasks, and 3) time window for successful R13, AS-R12, completion. Lacking are the specific timing of cues, listing of SY-R5 F&O the specific procedure guidance, and listing of the available HR-02, TH-05, cues/indications. However, the availability of HR-04, HR-05, cues/indications and procedure guidance is specifically AS-07, SY-05 evaluated by the PSFs. Related F&O is HR-02. STP's plan Is to migrate the HEPs to the EPRI HRA Calculator which will result In listing specific cues and procedure guidance.

[CR 04-13754-2-41 Human HR-G1 Yes HR-15, HR-17, Response: Covered under peer review HR-R11 F&O Reliability HR-18 HR-05, HR-07 AnalysisTH-5 Human HR-G2 Yes HR-02, HR-11 Response: At STP, the FLIM method has been used to HR-R2, HR-R7 Reliability determine HEPs. This method accounts for cognition and F&O HR-02, Analysis execution errors via the Performance Shaping Factors. An HR-03 TH-05 example of the cognition-related PSFs is titled 'Plant Man-Machine Interface and Indications". STP plans to migrate the HEPs to the EPRI HRA calculator - this tool provides explicit treatment of Pcog and Pexe via the CBDTM/THERP methods.

Human HR-G3 Partial HR-17, HR-18 Response: The FLIM PSFs evaluate the Impact of (a) HR-R11 F&O Reliability through (h). Items (i) and (j) are not explicitly evaluated in HR-05, HR-07, Analysis the FLIM PSF worksheets. The EPRI HRA Calculator TH-05

__ evaluates all of these supporting requirement elements.

Human HR-G4 Partial HR-18, HR-19, Response: STP time windows generally meet this category HR-RI2, HR-Reliability HR-20, AS-13 11requirement (as clarified In HR-G4 AppA). Time windows R13, AS-R4 Analysis are documented In PLG-0675 (original STP PSA) Volume F&O HR-07, 4, Appendix B, and the TH notebook (MAAP calculations for TH-05 selected HEPs); The point In time for relevant Indications are NOT provided. Also, the recent HRA update assessments do not provide a reference for the time windows. See related F&O HR-07. [CR 04-13754-2-51 Human HR-G5 Partial HR-16, HR-18, Response: The requirement Is met during the operator HR-R10, HR-Reliability HR-20 Interviews which Include a talk-through of the HEP scenario R13 F&O HR-Analysis sheet and applicable procedures. Concurrence of the 04, HR-07, TH-reasonableness of the listed time window is also requested 05 during this process.

Human HR-G6 Yes HR-12 Response: This supporting requirement Is met by the HR-R8 Reliability Inherent review and approval process for developing HEPs.

Analysis The peformance of this consistency check Is not specifically documented. (CR 04-13754-2-61 Human HR-G7 Partial HR-26, DE-07 Response: This systematic dependency analysis has not HR-R16, DE-R5 Reliability been performed - refer to F&O HR-06. [04-13754-2-7] F&O HR-06 Analysis Human HR-G8 No HR-27 Response: This supporting requirement has not been met, HR-R16 F&O Reliability and Is dependent on completing HR-G7. Minimum value of HR-06 Analysis 1E-04 for a single HEP has been described In the latest HRA update, PRA-99-010. [CR 04-13754-2-81 Human HR-G9 No Response: The HEPs are developed In RISKMAN as Reliability lognormal distributions, and thus have an associated error Analysis factor. The mean values are used In the PRA

_ __ _ ___ _ _ _ _ _ quantifications. _ _ _ _ _ _ _

Human HR-H1 Yes HR-21, HR-22, Response: Human recovery actions are included as HR-R14, HR-Reliability HR-23 appropriate in the STP PRA to reduce unnecessary R6 F&O HR-04 Analysis conservatism.

Human HR-H2 Yes HR-22, HR-23 Response: STP use of recovery actions meet these HR-R6 F&O Reliability supporting requirements. In general, recovery actions are HR-04 Analysis only credited If approved procedures support the actions.

Operators are trained on approved casualty mitigation procedures (EOPs, Off-normals, Annunciator Response).

These procedures typically contain the applicable cues.

Attention is given to the appropriate elements of HR-G3 for Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 14 of 21 PRA ASME Included In NEI-00-02 -. SW Evaluaton Peer Review Technical SR NEI 0-2 -ELEMENTS :; - . .

Element - -- - * - ,,, . . -- - -

PSFs.

Human HR-H3 Yes HR-26 Response: Dependency analysis of multiple recovery HFEs HR-R16 Reliability In a sequence has not been systematically performed -

Analysis refer to F&O HR-06.

Human HR-11 Partial HR-28, HR-30 Response: Documentation of the HRA Is contained in PLG- HR-R17 Reliability 0675 Vol.4 Section 14, the IPEEE, and assessment PRA-Analysis99-010. Enough detail is contained in these documents to understand the STP HRA. Some of the documentation specified In this supporting requirement is not available for certain SRs. Examples include:

1) documentation of pre-initiator screening - see F&O HR-01,

2) dependency analysis - see F&O HR-06,

3) summarized source of timing information, and

4) basis for minimum probability for multiple HEPs occurring In a sequence.

_CR 04-13754-2-101 Data DA-Al Yes DA-04, DA-05, Response: Covered under peer review SY-R8, SY-R14 Analysis DA-1 5, SY-08, F&O DA-02, SY-14 SY-03 Data DA-A2 No Response: Lognormal distributions have predominantly Analysis been used for the STP PRA data analysis.

Data DA-A3 Yes DA-04, DA-05, Response: Plant specific data updates meet this DA-R3, SY-R8 Analysis DA-06, DA-07, requirement. F&O DA-02 SY-08 Data DA-Bi Yes DA-05 Response: Covered under peer review Analysis Data DA-B2 Yes DA-05, DA-06 Response: A review of plant specific data updates indicates Analysis that this requirement has been met.

Data DA-Cl Yes DA-04, DA-07, Response: Covered under peer review DA-R3, DA-R8, Analysis DA-09, DA-1 9, DA-R6 F&O DA-20 DA-02, DA-03, DA-04 Data DA-C2 Yes DA-04, DA-05, Response: Covered under peer review DA-R3, DA-Analysis DA-06, DA-07, R12, DA-R6, DA-14, DA-15, MU-R4 F&O DA-19, DA-20, DA-02, DA-03, MU-05 DA-04 Data DA-C3 Partial DA-04, DA-05, Response: The plant specific Data Update Notebook DA-R3, MU-R4 Analysis DA-06, DA-07, documents this requirement See IE Notebook. F&O DA-02 MU-O5 Data DA-C4 No Response: The plant specific Data Update notebook Analysis documents this requirement. Refer to Table I in the notebook.

Data DA-C5 No Response: Review of the data updates Indicate this Analysis requirement was followed.

Data DA-C6 Yes DA-06, DA-07 Response: The plant specific Data Update notebook DA-R3 Analysis _ documents this requirement.

Data DA-C7 Yes DA-06, DA-07 Response: Covered under peer review DA-R3 Analysis Data DA-C8 No Response: The plant specific data update has not Identified Analysis a need to use operational records to determine the time that components were configured in their standby status.

Reasonable assumptions based on support system operating status satisfies this requirement.

Data DA-C9 Yes DA-04, DA-06, Response: A review of plant specific data updates indicates DA-R3 F&O Analysis DA-07 that this requirement has been met. DA-02 Table I - STP Self-Assessment Results NOC-AE-04001813 Page 15 of 21 PRA ASME Included In - NEI-00 STP Evaluation Peer Review

-Technical'- SR NEI 00-02 ELEMENTS . . -

Element -. *-.-..i_ ; i . al ' ,,.

Data DA- No Response: Surveillance test procedures are reviewed and Analysis C10 credited appropriately for demands. Refer to the Data Update notebook Rev.4.

Data DA- No Response: Maintenance unavailabilities have been updated Analysis C1l based on RAsCal data. RAsCal data meets this requirement Data DA- No Response: Maintenance unavailabilities have been updated Analysis C12 based on RAsCal data. RAsCal data meets this requirement.

Data DA- No Response: Maintenance unavailabilities have been updated Analysis C13 based on RAsCal data. RAsCal data meets this requirement.

Data DA- Yes DA-15, AS-16, Response: Maintenance unavailabilities have been updated AS-R6, SY-R24 Analysis C14 SY-24 based on RAsCal data. RAsCal data meets this requirement.

Data DA- Yes DA-15, IE-13, Response: Repair times for the support system Initiators IE-R1O, IE-R13, Analysis C15 IE-15, IE-16, have been collected from actual unplanned maintenance IE-R7, AS-R6, AS-16, SY-24, events. Recovery times for LOOP events have not been SY-R24 OU-18 collected since STP has not experienced a LOOP Initiator as defined In the model. The LOOP updates Include recoverV times from STP's grid data.

Data DA-DI No Response: The Bayesian update process Is used to Analysis calculate parameter estimates. The update process for component failures Is generally limited to components that have experienced a large number of failures over the update period (e.g., MRPSAF criteria exceeded). For the data not updated with plant specific experience, STP should either update the data with plant specific experience or update with recent industry generic data. [CR 04-13754 1 1

Data DA-D2 No , Response: STPs data variables have been developed Analysis consistent with this requirement. Most data Is based on

_generic estimates or plant-specific updates.

Data DA-D3 Partial OU-30 Response: All STP data parameters Include the mean value QU-R10 Analysis and statistical parameters associated with a lognormal distribution as represented by a DPD.

Data DA-D4 No Response: This was addressed In the WOG peer reviewl Analysis and associated F&Os. A data update guideline would better support this requirement. [CR 04-13754-3-21 Data DA-D5 Partial DA-08, DA-09, Response: STPs CCF parameters are based on the DA-R7. DA-R8, Analysis DA-1 0, DA-1 1, Multiple Greek Letter model, which meets this requirement. DA-R9, DA-DA-12, DA-13, Ri0, DA-R1, DA-14 DA-R12 F&O DA-01 Data DA-D6 Partial DA-08, DA-09, Response: The STP WOG PRA Peer review F&O DA-01 DA-R7, DA-R8, Analysis DA-10, DA-i1, addresses this requirement. DA-R9, DA-DA-12, DA-13, RIO, DA-R 1, DA-14 DA-R12 F&O DA-01 Data DA-D7 No Response: STP's model update process and design Analysis change Impact review ensures that appropriate data Is used to support the system models.

Data DA-El Partial DA-01, DA-19, Response: STPs data update documentation lacks some of DA-R1, DA-R6 Analysis DA-20 the requirements. The following documentation needs to be F&O DA-03, generated to meet the requirement: (a) system and DA-04 component boundaries used to establish component failure probabilities (c) sources for generic parameter estimates (Q) key assumptions made In the Interpretation of data and the reasoning (based on engineering, systems modeling, operations, and statistical knowledge) supporting its use In parameter estimation (i) the rationale for any distributions used as priors for Bayesian updates, where applicable [CR 04-13754-3-31 Intemal IF-Al No Response: Flooding areas are defined b all three Items DE-R9, IF-5 Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 16 of 21

,-PRA ASME Included In -' NEI-WO-02 ,- ;...

- -STPEvaluation - ,, :-- -. - Peer Revlew .

Technical SR NEI!2 l . ELEMEN * >:1.i.-

Flooding identified In the standard. F&O DE-06 Internal IF-A2 No Response: The Spatial Interactions Database covers this DE-R9, IF-6 Flooding Information. All equipment potentially affected by Internal F&O DE-06 floods are Identified. Since most of the Internal flooding scenarios were screened out In the early screening process, spatial locations were not required. Only flood scenario, Z123-FW-01 required further analysis based on spatial Information (Ref. IPE section 3.4.3.3). This scenario also screened out below the significance threshold.

Internal IF-A3 No Response: Spatial Interactions Database contains SSCs DE-R9, IF-6 Flooding within flood areas. F&O DE-06 Internal IF-A4 No Response: A plant walkdown was performed to verify/obtain DE-R9 F&O Flooding spatial Information, SSCs and potential flood sources. DE-06

Reference:

Original PSA and IPE Internal IF-B1 No Response: Flooding Sources are identified In Spatial IF-7 F&O DE-Flooding Interactions Database. Identification is performed by 02 analyzing the type of flooding source (e.g., Fire Hoses, Moderate/High Energy Unes, etc.). Reference Table D-3 In Original PRA.

Internal IF-B2 No Response: WOG PRA Peer Cert, CR 02-6188-7-3 Pipe IF-7 F&O DE-Flooding breaks and tank ruptures appear to be the only cause of 03 flooding considered In the 1988 analysis. Floods caused by human errors during maintenance, water hammer, and failures during off-normal operations were not considered as flooding Initiators. Will be corrected In flooding update.

Internal IF-B3 No Response: WOG PRA Peer Cert. CR 02-6188-7-2 The IF-9 F&O DE-Flooding maximum flow rate of the flood was not considered. The 04 screening analysis appears to be based on the flood water volume caused by the design basis flood. Flow rates, duration of the flow rates and ultimate water volumes

- . produced during the flood were not stated. Reference to the drain size was not mentioned. Will be corrected In flooding update.

Internal IF-B4 No Remarks: Floor drains are credited for limiting the IF-10 F&O DE-Flooding propagation of Internal floods but not for limiting the effect 04 on flooding of the room with which the drains are located.

This Is being reevaluated In the Internal flood hazard update (on-going).

Internal IF-Cl No Response: WOG PRA Peer Cert. CR 02-6188-7-1 IF-1I F&O DE-Flooding Propagation pathways: Flood propagation through drains, 01 stairwells, and cracks under doors were considered. It Is not apparent that pathways such as HVAC ducts, pipe chases and penetrations, pipe tunnels were considered In the same detail. All flood barriers were assumed to be In their functional position. That Is, doors being open, structure failure of doors, dikes being removed for maintenance were not considered. Drains being blocked or drain line check valves being failed open were not considered. All rooms were screened based on room alone. No propagation analysis was done. STPs Initial Response: Spatial Interactions Database contains multiple examples of flood propagation from one zone to another. It Is assumed that propagation of water from one room to another will flood all equipment within the room. Will be corrected in flooding update.

Internal IF-C2 No Remarks: The STP PRA contains no Internal flooding IF-12 Flooding events, therefore, this Is not an issue. Any Justification for screening Internal flooding scenarios is documented In the

_spatial interactions.

Internal IF-C3 No Response: The Spatial Interactions Database IF-13 Flooding conservatively fails equipment. (See IPE Table 3.4.1.7 Equipment Susceptibility)

Table I - STP Self-Assessment Results NOC-AE-04001813 Page 17 of 21

' PRA'-- ASME Included In NEI-00-02 - STP.Evaluation Peer Revlew.

Technical SR - - NEI02 ELEMENTS Element - ' i -

Internal IF-C4 No Response: Propagation pathways were developed from IF-14 Flooding plant walkdowns. No credit was given to operator with the exception of the Control Room (Manned 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months a day).

Internal IF-C5 No Remarks: This Information Is documented In the Spatial IF-14 Flooding Interactions Database.

Internal IF-C6 No Response: With the exception of floods within the Control IF-14 Flooding Room, no human mitigation was credited.

Internal IF-Di No Response: See section 8 in the Original PRA for Flooding documentation of the structured, systematic' process for

.__ _ _developing the spatial Interactions database.

Internal IF-D2 No Response: Flooding scenarios were binned Into different IF-14 Flooding classes (i.e., type of scenarios), Including scenarios that result In initiating events. All internal flooding events were screened out early In the screening process. However, If further evaluation had been required, then systems alignments, Including support systems, would have been performed. See PRA fire analysis for examples.

Internal IF-D3 No Response: No Internal flooding scenarios required grouping Flooding of Initiating events. Therefore, this element Is not a concern I at STP.

Internal IF-D4 No Remarks: STP does not have any shared systems or Flooding structures that would Impact the internal flooding analysis.

Internal IF-D5 No Response: WOG PRA Peer Cert. CR 02-6188-7-5 IF-15 F&O DE-Flooding Flooding frequencies were based on a 1983 paper, which 07 provided an overall frequency for flooding In the Aux, DG, turbine buildings. These frequencies were apportioned to rooms of Interest based on square footage. Continued use of flooding frequencies based on 19-year-old data Is not appropriate. Further, the method of apportioning the data may no longer reflect current Industry experience. STP Response: Disagree with F&O. The flooding frequencies were developed from 2 sources: 1) LER database and then

2) updated/reanalyzed to support shutdown events at Seabrook (Ref PLG-0624). STP is currently In the process of updating the Internal flooding analysis with TR-11880, Piping System Failure Rates. Will be corrected in flooding update.

Internal IF-El No Remarks: The STP PRA contains no Internal flooding Flooding scenarios, therefore, this is not an Issue.

Internal IF-E2 No Remarks: The STP PRA contains no Internal flooding F&O DE-04 Flooding scenarios, therefore, this Is not an Issue.

Internal IF-E3 No Remarks: The STP PRA contains no Internal flooding Flooding scenarios, therefore, this Is not an Issue.

Internal IF-E4 No Remarks: The STP PRA contains no Internal flooding Flooding scenarios, therefore, this is not an Issue.

Internal IF-E5 No Remarks: The STP PRA contains no Internal flooding IF-12 Flooding scenarios, therefore, this Is not an Issue.

Internal IF-E6 No Remarks: The STP PRA contains no Internal flooding IF-12 Flooding scenarios, therefore, this Is not an Issue.

Internal IF-E7 No Remarks: The STP PRA contains no Internal flooding F&O DE-09 Flooding scenarios, therefore, this Is not an Issue.

Internal IF-Fl No Remarks: The Spatial Interactions Database is well Flooding documented. The WOG PRA Peer resulted In a Level of Significance of "S' with the following documentation: In all aspect of spatial dependencies, the STPEGS PRA (in 1988) performed a rigorous hazard analysis which considered jet water, spray water, explosive canisters, equipment drops, high temperatures and missiles. The work was largely completed In an extensive walk down. All rooms were walked down and documented.

Table I - STP Self-Assessment Results NOC-AE-04001813 Page 18 of 21 i-PRA ASME Included In E NEI-OO-02 Evaluation Peer Review

STP Technical- SR: NEI 0002 ELEMENTS -j

.Eleent , -. -

Internal IF-F2 No Remarks: This Information is documented in the IPE and Flooding Original PSA Including Table D-6.

Quantificati QU-A1 Yes AS-04, AS-05, Response: STP PRA performs this by default In RISKMAN AS-Ri, AS-R2, on Analysis AS-06, AS-07, software AS-R3, AS-R12 AS-08, AS-09, F&O AS-O1, AS-10, AS-19 AS-03, AS-09, AS-1 0, SY-06, AS-06, AS-04, TH-04, AS-05,AS-07 Quantificati QU-A2 Yes QU-08 Response: Conditional split fractions used In the event tree QU-R13 F&O on Analysis quantification process incorporate the effects of 'The State AS-10 of Knowledge" dependence In component failure data.

Quantificati QU-A3 Yes QU-04, OU- Response: Default capability of RISKMAN software QU-R13, QU-on Analysis 08, QU-09, Rl, OU-R2 FO QU-10, QU- AS-10, HR-O6, 11, QU-12, HR-07, QU-05 QU-13 Quantificati QU-A4 Yes aU-18, QU-19 Response: Recovery credited in STP PRA see System OU-R4 on Analysis Notebooks and Event Tree Notebooks. Most operator recovery top events start with letter"O."

Quantificati QU-B1 Yes QU-04, QU- Response: RISKMAN software used and sensitivity cases F&O QU-01 on Analysis 05, QU-06 after quantification using different methodologies are performed to Insure appropriate solutions. User group tracks the limitations of the code and known problems and resolutions Quantificati QU-B2 Yes QU-21, QU- Response: Sensitivity studies are performed at various QU-R5, OU-R6, on Analysis 22, QU-23, cutoff frequencies to Insure stable results for final solution. QU-R8

__- QU-24 See Level 1 and Level 2 Quantification Notebooks.

Quantificati QU-B3 Partial QU-19, QU- Response: Sensitivity studies are performed at various QU-R4, QU-R5, on Analysis 22, QU-24 cutoff frequencies to Insure stable results for final solution. QU-R6 See Level 1 and Level 2 Quantification Notebooks Quantificati QU-B4 Yes QU-04 Response: RISKMAN software uses both mean and rare on Analysis event approximation solutions and Is now capable of producing exact solutions utilizing binary decision diagrams (this capability Is not used In the current model). Sensitivity studies are performed to Insure reasonable results. See System Notebooks for System Level Quantification.

Additional software passes SQA requirements to Insure it produces reasonable results.

Quantificati QU-B5 Yes QU-14 Response: Covered under peer review QU-R3 on Analysis Quantificati QU-B6 Yes QU-04, QU- Response: Inherent property of event tree quantification OU-R15, AS-R3 on Analysis 20, QU-25, F&O AS-06, AS-08, AS-09 AS-04, TH-04 Quantificati QU-B7 Yes QU-26 Response: This function performed In model update QU-R7 on Analysis verification.

Quantificati QU-B8 No Response: Not directly applicable to RISKMAN models.

on Analysis 'Logic flags' in event trees are typically macros like those found in PMET event tree are either set to failure by the associated logic statements or are by definition 'Not Failed". In system fault trees, 'Logic Flags' are typically House events whose status Is explicitly controlled by split fraction definition equations. See the various system notebooks or the event tree macros and split fraction rules.

Quantificati QU-B9 Partial SY-09 Response: See System Notebooks, and Event Tree SY-R9 F&O on Analysis Notebooks. Risk significance of components is SY-07 accomplished using the RISKMAN software and appropriate mapping of conditional split fraction groups in the model. See also response to SY-AlO.

Table 1 - STP Self-Assessment Results NOC-AE-04001813 Page 19 of 21 PRA. ASME Included In i -NEI-00-02 STP Evaluation Peer Review Technical - SR :NEI00-02 ELEMENTS: , .,-

, Element .

Quantificati 0U-Cl Yes 0U-10,0U- Response: Sensitivity studies are performed following HR-R16 F&O on Analysis 17, HR-26 model quantification. However see HR Section comments HR-06, HR-07 specifically HR-G6, 7, 8 comments and F&O HR-06.

Quantificati 0U-C2 Partial QU-10, 0U-17 Response: See HR Section comments and F&O HR-06. F&O HR-06, on Analysis HR-07 Quantificati QU-C3 Yes 0U-20 Response: Validated during model update, see Event Tree QU-R15 on Analysis Notebooks and model using RISKMAN software.

Quantificati QU-D1 Yes 0U-08, 0U- Response: See F&O QU-02, 04, 05, for related Issues. A QU-R13, 0U-on Analysis 09, QU-10, review Is conducted during the model update process and R1, QU-R2, QU-11, QU- prior to application updates such as GQA risk ranking. OU-R3, QU-12, QU-13, R14 F&OAS-0U-14, 0U- 10, HR-06, HR-15, 0U-16, 07, 0U-02, 0U-OU-17 04, 0U-05 Quantificati OU-D2 Partial 0U-27, 0U- Response: A review Is conducted during the model update QU-R9, SY-R22 on Analysis 28, SY-22 process. Periodically Interviews and simulator experiments F&O 0U-03, are performed to validate operator actions, F&O HR-04, 06. HR-04, HR-06

._ _This issue is an open item.

Quantificati 0U-D3 Yes 0U-08, 0U- Response: See F&O QU-05. CR 02-618-9-5, open item QU-R13 F&O on Analysis 11, QU-31 AS-10, QU-05, OU-02, OU-04 CR 02-618-9-5 Quantificati 0U-D4 Yes 0U-15 Response: Performed during Model update OU-R14 on Analysis Quantificati QU-D5 Yes QU-08, OU-31 Response: See Level 1 Quantification Notebook for overall QU-R13 F&O on Analysis Importance and system level importance. See component AS-10, 0U-02, Importance In PRA 03-013R1 analysis assessment for risk OU-04 ranking of PRA modeled components used in GOQA

. . exemption from special treatment application. -

Quantificati QU-El Yes QU-30 Response: See F&O 0U-03, Uncertainty Is evaluated 0U-R10 F&O on Analysis during model update process In accordance with procedure 0U-03

. OPGP03-ZA-0305 Quantificati QU-E2 Yes OU-27, QU-28 Response: See F&O U4-03, Uncertainty Is evaluated QU-R9 F&O on Analysis during model update process OPGP03-ZA-0305 OU-03 Quantificati QU-E3 Partial QU-30 Response: See F&O U4-03, Uncertainty Is evaluated 0U-R10 F&O on Analysis during model update process OPGP03-ZA-0305 OU-03 Quantificati QU-E4 Partial OU-28, 0U- Response: See F&O OU-03, Uncertainty Is evaluated 0U-R10 F&O on Analysis 29, 0U-30 during model update process ZA-305 See Level 1 QU-03, U4-02 Quantification Notebook. Table A-1 response: Key assumptions / uncertainty as defined In Reg. Guide 1.200 not yet documented.

Quantificati U1-F1 Partial 0U-31, QU- Response: See Level 1 and Level 2 Quantification QU-R11 F&O on Analysis 32, 0U-34 Notebooks for overall results, System Notebooks for system U4-02, 0U-04 level results. Table A-1 response: part G significant basic events causing accident sequences to be non-significant Is

__ not documented.

Quantificati QU-F2 Yes QU-31 Response: See Level 1 and Level 2 Quantification on Analysis Notebooks for overall results, System Notebooks for system level results. Significant sequences reviewed In both system level quantification and event tree quantification during model updates.

Quantificati QU-F3 Yes 0U-27, 0U- Response: See Level 1 and Level 2 Quantification QU-R9, 0U-on Analysis 28, 0U-32 Notebooks for overall results, System Notebooks for system R1 1 F&O 0U-level results. Key sources of uncertainty and key 03 assumptions are not yet Identified or analyzed In the PRA

_ _ model.

Quantificati QU-F4 Yes QU-12, 0U-13 Response: See Level 1 and Level 2 Quantification QU-R2 on Analysis Notebooks for overall results, System Notebooks for system I _level results.

Table I - STP Self-Assessment Results NOC-AE-04001813 Page 20 of 21

, PRA ASME Included In NEI-00-02 STP Evaluation Peer Review

-Technical SR -NEI 0-2 ELEMENTS; - - - -

Element -

Quantificati QU-F5 Yes QU-04, MU-07 Response: RISKMAN program has been verified by the MU-R5 F&8 on Analysis vendor SQA program and by site SQA evaluation of the MU-04 model. Vender retains all documentation 'proof' for code capability and yielding correct results. When a revision to the software takes place STP verifies same results with

_ ___ _ __ _ _ __ _ _ __._ _ single m odel. _ _ _ _ _ _

Quantificati QU-F6 No Response: See application analysis assessments, and on Analysis PRA assumptions.

LERF LE-Al Yes 12-07, 12-08, Response: See Draft Level 2 Report. L2-R4, AS-R5, Analysis 12-22, AS-1 4, AS-R9, AS-AS-20, AS-21, Ri0, AS-R11 AS-22, AS-23 F&O 12-01, L2-02, L2-04 TH-01 LERF LE-A2 Yes 12-07, 12-08, Response: See Draft Level 2 Report. L2-R4, AS-R9 Analysis AS-21 F&O 12-01, L2-02 LERF LE-A3 Yes 12-07, 12-08, Response: See Draft Level 2 Report. L2-R4 F&O L2-Analysis 12-21 01, 12-02, L2-

._ 06 LERF LE-A4 Yes 12-07, 12-08, Response: See Draft Level 2 Report. L2-R4, AS-R9 Analysis 12-21, AS-20, F&O L2-01, L2-AS-21 02, 12-06 LERF LE-A5 Yes 12-08, 12-21, Response: See Draft Level 2 Report. AS-R9 F&O Analysis AS-20 12-01, 12-02, 12-06 LERF LE-B1 Yes 1l2-08, 12-10, Response: See Draft Level 2 Report. L2-R6, L2-R7, Analysis 12-15, 12-16, L2-R3 F&O L2-12-17, 12-19 01, 12-02, L2-03 LERF LE-B2 Yes 12-13, L2-14 Response: See Draft Level 2 Report. L2-R7 F&O L2-Analysis 03 LERF LE-B3 Yes 12-14, 12-15, Response: See Draft Level 2 Report. L2-R7, ST-R2 Analysis ST-04 F&O TH-03 LERF LE-C1 Yes L2-24 Response: See Draft Level 2 Report. L2-R8 Analysis LERF LE-C2 Yes 12-09, 12-12, Response: See Draft Level 2 Report. 12-R5 F&O L2-Analysis L2-25 05 LERF LE-C3 Yes 12-08, 12-24, Response: See Draft Level 2 Report. L2-R8 F&O L2-Analysis L2-25 01, 12-02, L2-

__ 05 LERF LE-C4 Yes 12-04, 12-05, Response: See Draft Level 2 Report. L2-R3, L2-R3 Analysis L2-06 F&O 12-01, L2-02, L2-04 LERF LE-C5 Yes 12-07, L2-11, Response: See Draft Level 2 Report. L2-R4, AS-R9 Analysis 12-25, AS-20, F&O 12-05

_ _ AS-21 LERF LE-C6 Yes 12-12, 12-24, Response: See Draft Level 2 Report. L2-R8 F&O L2-Analysis 12-25 05 LERF LE-C7 Yes 12-07, L2-11, Response: See Draft Level 2 Report. L2-R4, L2-R8 Analysis 12-12, 12-24 F&O L2-05 LERF LE-C8 Yes 12-11, 12-12 Response: See Draft Level 2 Report. F&O 12-05 Analysis _ _ _ _ _ _

LERF LE-C9 Yes 12-11, 12-12, Response: See Draft Level 2 Report. L2-R3, L2-R8, Analysis 12-16, 12-24, AS-R9 F&O 12-25. AS-20 L2-05 LERF LE- No Response: See Draft Level 2 Report.

Analysis C10 LERF LE-DI Yes 12-14, 12-15, Response: See Draft Level 2 Report. L2-R7, L2-R3, Analysis 12-16, 12-17, L2-R8 F&O L2-12-18, 12-19, 01, 12-02, L2-12-20, ST-05, 03 Table I - STP Self-Assessment Results NOC-AE-04001813 Page 21 of 21

. - PRA -ASME In E;N Included E102 -STP Evaluatlon -. Peer Review-Technical 'SR - NEI 00-02 'ELEMENTS -. ,

Element :

ST-06 LERF LE-D2 No Response: See Draft Level 2 Report.

Analvsis _______________________________________

LERF LE-D3 Yes IE-14, ST-09 Response: See Draft Level 2 Report. IE-R6, ST-R3 Analysis F&O IE-02, IE-03, ST-01 LERF LE-D4 No Response: See Draft Level 2 Report.

Analysis LERF LE-D5 No Response: See Draft Level 2 Report.

Analysis LERF LE-D6 Yes 12-16, 12-18, Response: See Draft Level 2 Report. L2-R3, L2-R7, Analysis 12-19, 12-24, L2-R8 F&O L2-12-25 03, 12-05 LERF LE-El No 12-05, 12-11, Response: See Draft Level 2 Report. L2-R3 F&O L2-Analysis L2-12 01, 12-02, L2-04, 12-05 LERF LE-E2 Yes 12-12, 12-13, Response: See Draft Level 2 Report. L2-R7, L2-R8 Analysis 12-17, 12-18, F&O 12-01, L2-DA-04, HR-15, 02, 12-03, DA-12-19, L2-20 02 LERF LE-E3 Yes QU sub- Response: See Draft Level 2 Report.

Analysis elements applicable to LERF LERF LE-FI Yes QU-08, QU- Response: See Draft Level 2 Report. QU-R13 F&O Analysis 09, au-10, AS-10, HR-06, OU-11, QU-31 HR-07, QU-02,

_._ QU-05 LERF LE-F2 No QU-27 Response: See Draft Level 2 Report. OU-R9 F&O Analysis . QU-03 LERF LE-GI Partial L2-26, 12-27, Response: See Draft Level 2 Report. 12-R9, L2-Rio Analysis L2-28 LERF LE-G2 Partial L2-26, 12-27, Response: See Draft Level 2 Report. L2-R9, L2-R1O Analysis __ L2-28 .

LERF LE-G3 Partial 12-26, 12-27, Response: See Draft Level 2 Report. L2-R9, L2-RiO Analysis 12-28 LERF LE-G4 Partial 12-26, 12-27, Response: See Draft Level 2 Report. L2-R9, L2-Rio Analysis L2-28 _

LERF LE-G5 Partial 12-26, L2-27, Response: See Draft Level 2 Report. L2-R9, 12-R10 Analysis L2-28 LERF LE-G6 Partial L2-26, L2-27, Response: See Draft Level 2 Report. L2-R9, L2-R1O Analysis L2-28 LERF LE-G7 Partial 12-26, 12-27, Response: See Draft Level 2 Report. L2-R9, L2-RiO Analvsis 12-28 LERF LE-G8 Partial 12-26, 12-27, Response: See Draft Level 2 Report. L2-R9, L2-Rio Analysis _ 12-281 NOC-AE-04001813 Attachment 4 Key Assumptions and Approximations NOC-AE-04001813 Page I of I Identification of Key Assumptions and Approximations The following are key sources of uncertainty in the STP PRA. Sensitivity analyses for the key sources of uncertainty are being performed.

1. Reactor Coolant Pump Seal LOCA Modeling - The STP PRA uses a Rhodes model for seal LOCA behavior, which is incorporated into loss of offsite power (LOOP) and station blackout (SBO) recovery to determine key timing issues. Seal LOCA as a result of loss of cooling (i.e.,

component cooling water) is not a significant contributor at STP due to unique plant design features, including improved reactor coolant pump seal O-rings, a diesel-backed positive displacement charging pump that is independent from other cooling requirements, and high pressure injection pumps that are independent of component cooling water for injection and independent of external room cooling (one train only).

2. BRA Modeling - HRA modeling is described in Attachment 2.

3. SBO Recovery - LOOP and SBO are leading contributors to CDF and LERF at STP. Offsite power recovery was updated for the last model, STP_REV4. Sensitivity studies using different estimates of offsite power recovery under uncertain grid conditions (from deregulation) have not been performed. These sensitivity studies will be completed in support of the next model revision.

4. Balance of Plant (BOP) Modeling for Post-Trip Response - The STP PRA does not include BOP systems in the current PRA, which was a peer review finding. Sensitivity studies under different assumptions for operation of BOP systems after a plant trip are being performed. BOP system models will be incorporated in the next PRA update.

NOC-AE-04001813 Attachment 5 Resolution of Peer Review Comments NOC-AE-04001813 Page I of 19 Resolution of Peer Review Comments In April 2002, the STP PRA underwent an industry peer review performed in accordance with NEI-00-02, "Industry PRA Peer Review Process." All technical elements within the scope of the peer review were graded as sufficient to support applications requiring the capabilities of a Grade 2 (risk-ranking applications). Several other technical elements were further graded as sufficient to support applications requiring the capabilities defined for Grade 3 (risk-informed applications supported by deterministic insights). The overall assessment of the peer reviewer was that the STP PRA could effectively be used to support applications involving risk-significance evaluations supported by deterministic input once the items noted in the technical element summaries and in the Fact & Observations (F&O) sheets were addressed appropriately for specific applications.

STPNOC is using its Corrective Action Program as a tracking mechanism for resolving the items identified by the peer review team. Most F&O items identified by the peer team have been completed and incorporated into the latest revision of the STP PRA (Revision 4). Other F&O items are currently being addressed and will be completed in 2005 prior to the implementation phase of RITS 4B. The STP PRA Revision 4 model is the basis for this application of Risk-Informed Technical Specifications. The full report of the peer review is available in the archival information.

Table 2 provides the results of the Peer Review and identifies F&O items that were generated.

F&O items that are highlighted in Table 2 are Significance Level A or Significance Level B.

Table 2 - Peer Review Results NOC-AE-04001813 Page 2 of 19

-CRITERIA-- -NEI :ASME, NEI CRITERIA

= PRA Contingent Related Supporting CATEGORY: 02 SR -Grade PRA Grade Facts & Criteria and DESIG. DESG- Obs. Notes GUIDANCE IE-01 Describes the process used 3 RI IE-02 Consistent with industry practices 3 R2 IE-03 Sufficient detail provided for 3 Ri1 I__ - reproducing the evaluation IDENTIFICATION IE-04 Grouped Initiators by plant response 3 R3 AND GROUPING consistent with event tree structure

.___ and success criteria. .

IE-05 IE-A9 The class of initiating events that Is 3 IE-01 R5 caused by failure of part or all of a system that supports the front-line safety function are addressed:

- Cooling water systems (e.g., service water, component cooling water, etc.)

- AC Power

- DC Power

- HV_

IE-06 For multi-unit sites with shared NA, R12 systems, the Impact of Initiators requiring simultaneous response (e.g.,

LOOP, loss of cooling source due to Ice, loss of an AC or DC bus, etc.) are Included IE-07 Initiators considered cover the 3 R2, R3 spectrum of Intemal event challenges IE-08 All experienced initiators are 3 R8 accounted for In the model IE-09 It typical Initiators cited In NUREG- 3 R9 1150 or industry PSAs have been

. _ excluded, the basis is documented IE-10 A structured approach for plant 3 support systems is performed to determine If a loss of support system Initiator presents a unique challenge to the plant SUBSUMED IE-11(3) Treatment of subsumed Initiating 3R4 INITIATING events Istraceable EVENTS IE-12 Subsumed Initiating events are 3 R4 included, in non-risk significant sequences or non-risk significant

_ initiators DATA IE-13 Initiating event frequencies and 3 Rio recovery are consistent with industry experience or analysis IE-14 The features that lead to the frequency 3 R6 of interfacing system LOCA (e.g.,

surveillance test practices, start up procedures, etc.) are modeled explicitly or Identified in the PSA

_documentation.

IE-15(3) IE-Cl 1 Plant specific features are reflected in 3 R13 the Initiating event frequency and recoverv inputs where appropriate I I __

IE-1 6(3) IE-C2 Plant specific experience is reflected in t

3 R7 the Initiating event definitions and frequency plus recovery Inputs where appropriate IE-17 A systematic process Is used to 3 R14 Identify the need for and application of techniques such as plant specific models or FMEAs, to quantify Initiating event frequencies and recovery. (See also SY-21)

Table 2 - Peer Review Results NOC-AE-04001813 Page 3 of 19 CRITERIA NEI ASME NEICRITERIA; PRA: Contingent Related Supporting

'.-CATEGORY ' .02- 7 ',.SR- ; .. *-. ;Grade PRAGrade Facts & Criterla and.

DESIG .DESIG ' ' '.';: - -- ;' : ' Obs. ' Notes-'

DOCUMENTATION IE-18 Documentation provides the basis of 3 FiRl I the quantified values and Istraceable IE-19 Documentation reflects the process 3 used IE-20 Documentation provides the basis for 3 the Initiating event frequency groupings_

IE-21 Independent review provided for the 3 documented results GUIDANCE AS-01 Describes the process used 3 R13 AS-02 Consistent with industry practices 3 R13 AS-03 Sufficient detail provided for R13 reproducing the evaluation ACCIDENT AS-04 The event trees reflect the Initiating 3 AS-01 R1 SCENARIO event groupings EVALUATION AS-05 The models and analysis are 3 R2 consistent with the as-built plant (as could be confirmed during the Peer Review process)(6)

AS-06 The necessary critical safety functions 3 AS-03, are modeled In each sequence AS-09 AS-07 All relevant systems are credited for 3 each functions AS-08 The branching structure and transfers 3 AS-06 R3 among event trees maintain and resolve the failure paths AS-09 Success paths are defined correctly 3 ,

AS-1 0 Dependencies among top events are 4AS-05, Identified and addressed AS-06 AS-11 The method of treating dependencies 4 AS02, Is documented and consistently AS-06 applied to capture the dependencies among top events.

AS-12 PWRs: An appropriate model for the reactor coolant pump seal LOCA, which may result from a loss of seal cooling due to various causes, Is used and documented. Appropriate seal cooling dependencies are considered.

AS-13 Time phased evaluation Isincluded for 3R4 sequences with significant time dependent failure modes (e.g.,

batteries for SBO. PWR RCP seal LOCA) and significant recoveries (e.g.,

AC recovery for SBO)

AS-1 4 Functions and structure are adequate 3 R5 to discriminate among plant conditions necessary for Level 2 analysis AS-15 Transfers among event trees are 3 R3 performed correctly to avoid loss of

.__ __ information in the transfer AS-16 System/component repair and 3 R6 recovery, if Included In the accident sequences. are correctly modeled SUCCESSI AS-171 Functional success criteria are 3 R7 CRITERIA IIdentified I I Table 2 - Peer Review Results NOC-AE.04001813 Page 4 of 19

-, CRITERIA

NEI -t ASME; NEI CRITERIA

.. PRA. Contingent Related Supporting.

CATEGORY -'.02 -S; Grade PRA Grade Facts & Criteria and

- DESIG, D ESIGY* :U - Obs. - Notes

. -~- £- ,- -. ., ..4. C -. ...,,-:--:.

SUCCESS AS- Success criteria are based on a 3 R8 CRITERIA BASES 18(7) combination of generic realistic and plant-specific realistic thermal hydraulic analyses INTERFACE WITH AS-19 Reflects the EOPs and AOPs. (The 3 AS-07 R12 EOPs/AOPs functions and structure of the event trees are consistent with the EOPs and abnormal procedures). (See also SY-5)

ACCIDENT AS-20 The development of plant damage - _ NA, R9 SEQUENCE END- states, their relationship to functional STATES (PLANT failures, and their relationship to Level DAMAGE 1 event tree end states or linked fault STATES)(5) tree cut sets is documented.

AS-21 Plant damage states are sufficient to . NA, R9 support the transfer of Information to Level 2 _

AS-22 Plant damage states are based on a 3 TH-01 RIO clear, consistent definition of CDF that Isconsistent with industry usage AS-23 Plant damage states are based on 3 R11 mission time of 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months or separately justified DOCUMENTATION AS-24 Documentation provides the basis of 3 event tree structure and Is traceable to plant specific or generic analysis AS-25 Documentation reflects the process 3 used AS-26 Documentation Includes an 3 independent review for the documented results . .

GUIDANCE TH-01 Describes the process used 3 TH-06 TH-02Consistent with industry practices 3 TH-01 Ri TH-03 Sufficient detail provided for 3 TH-06 R4 reproducing the evaluation T&H ANALYSES TH- Combination of Generic realistic and 3 R5 04(1) Plant-specific realistic thermal hydraulic analysis are used MULTIPLE T&H TH-05 A combination of plant specific, 2 R6 INPUTS generic and FSAR calculations are used to support success criteria and I HRA timing. i GENERIC TH-06 Application of the generic 3 R7 ASSESSMENTS assessments account for limitations of the generic analysis when applied to the specific plant BEST ESTIMATE TH-07 Application of the T & H codes account 3 R3 CALCULATIONS for the limitations of each of the codes (e.g., MAAP, RETRAN, SAFER-GESTER)_

ROOM HEATUP TH-08 Documented evaluation available to 3 R2 CALCULATIONS support the modeling decisions, _

DOCUMENTATION TH-09 Documentation provides the basis of 3S RB the Thermal Hydraulic Analysis, Is traceable to plant specific or generic :

analysis, and demonstrates the reasonableness of the success criteria. i _

TH-10 Documentation reflects the process 3 R9 used TH-1 1 Documentation Includes an 3 R10 Independent review for the documented results Table 2 - Peer Review Results NOC-AE-04001813 Page 5 of 19

CRITERIA. NEI-00-- ASME: a -NEI CRITERIA . PRA* Contingent Related Supporting.

CATEGORY 02 SR . Grade PRA Grade' Facts & Criteria and DESI DESIG Obs. Notes GUIDANCE SY-01 Describes the process used 3 SY-01 R1 SY-02Consistent with Industry practices 3 R2 SY-03 Sufficient detail provided for 3 R3 I reproducing the evaluation SYSTEM MODELS SY-04 The system models are available for 3 R4 (e.g., Fault Trees) review SY-05 The models and analyses are 3 SY-05 R5 consistent with the as-built, as-operated plant Including EOPs and AOPs (See also AS-1 9)

SY-06 The structure of the system model 3 R6 provides detail down to at least the major active component level (e.g.,

pumps and valves)

SY-07 The level of detail of the system 3R7 models reflects certain passive components that may impact CDF.(6)

SY-08 The system models contain at a 3 R8 minimum the following (if applicable):

- Common cause failure contributors

- Test and maintenance unavailabilities

- Operator errors that can Influence system operability (where appropriate)

- False Instrument signals that can cause failures of the system

- Operator interface dependencies

_across systems or trains SY-09 Modules used In the system models 3 SY-07 R9 are well correlated to their constituent components and capable of providing Importance and parametric effects on a component level.

SY-10 Spatial or environmental dependencies 3 R10 (e.g., internal floods, room cooling, etc.) are addressed for each system within the system model or in the accident sequence evaluation.(5)

SY-1 I In some accident sequences, systems 3 SY-09 Ri1l are expected to perform in degraded environments (e.g., Inside containment after a LOCA). While equipment Is generally qualified for such an environment, there should be some evidence that a search has been made for equipment that Is not so qualified (e.g., statements that necessary equipment Is qualified.) Other examples of degraded environments Include:

- Steamline breaks outside containment

- Debris that could plug screens/filters (both internal and external to the plant), and

- heating of the water supply (e.g.,

PWR containment sump) that could affect pump operability SY-12 Support system requirements are 4 AS-06 R12 accounted for SY-13 The Inventories of air, power, and 3 R13 cooling sufficient to support the mission time (or potential deficiencies)

Table 2 - Peer Review Results NOC-AE-64001813 Page 6 of 19

. CRITERIA'.,- NEI-M ASME i; NEI CRITERIA -PRA Contingent Related Supporting

-CATEGORY - 02

-- SR;. -. .y :- .: Grade PRA Grade Facts & Criteria and DESIG DESIG >:"... . <. y-- :-;: - -Obs. , Notes are Identified and Included in the model as appropriate. (Also refer to Elements TH and DE regarding definition of success criteria) . -

SY-14 The system boundary Included In the 2 SY-03 R14 system model Is clearly discerned from a simplified schematic of system SY-15 The system model analysis considered 3 R15 generic system failure modes observed in industry(9)

SY-1 6 The system model analysis Included 3 SY-04 R16 plant specific failure modes(7), (9)

SYSTEM MODELS SY- Combination of Generic realistic and 3 SY-02, R17 (e.g., Fault Trees) 17(11) Plant-specific realistic thermal SY-08 continued hydraulic analysis SY-1 8 The system model nomenclature Is 3 R18 developed in a consistent manner to allow model manipulation and to represent the same designator when a component failure mode is used in multiple systems or trains.

SY-19 The systems used in the event trees 3 R19 have detailed system model development to support them unless they are generally treated with point estimate values, e.g.,:

- RPS

- Diesel Generators

- Switchyards The following Impact on !grades is suggested for the above sample Items (4):

- Conditional Probabilities (Split Fractions)

SY-20 The system models(4) are used to 3 R20 quantify the accident sequences by:

- Conditional Probabilities (Split Fractions)

SY-21 The impact of the system model on 3 R21 Initiating events has been examined (see also IE-10, IE-17)

SY-22 The assumptions for the system logic 3 R22 SY' model are Identified (12)

SY-23 The system operation under accident 3 R23 conditions Is Identified In the system notebook SY-24 System/component repair and 3 R24 recovery actions and modeling, if used, are Identified and documented (see also QU-1 8)

DOCUMENTATION SY-25 Reflects the process used 3 R25 SY-26 Includes an independent review for the 3 R26 documented results SY-27 Provides the basis of the system 3 R27 model and Is traceable to plant specific I or generic analysis GUIDANCE DA-01

I*

Describes the process used 4 3 4 4

RI DA-02

--. 4 . . __

Consistent with industry

-- I.

practices 3 R2 DA-03 Sufficient detail provided for 3 4- 4 renroducino the evaluation 4 4 FAILURE DA-04 The random Independent component 3 PROBABILITIES failure probability data used In the evaluation and where it can be justified Table 2 - Peer Review Results NOC-AE-04001813 Page 7 of 19 x CRITERIA . NEI ASME NEI CRITERIA ,PRA 4: Contingent Related Supporting

-'CATEGORY-: . 02I ,;,,SR' Grade- PRA Grade Facts & Criteria and

'DESld DESIG, Obs. . Notes' --

- ..-' .; .j 1 .- I ,,- i :

, .I

.- i- . ..

I. t is based on accumulated plant specific experience; otherwise, realistic generic data is used.

+ 4- I I DA-05 For plant specific data development, 3 similar components- have been grouped together in a reasonable manner and the grouping is supported by the documentation.

DA-06 For basic events derived using NA standby failure rate data, the plant specific surveillance test Intervals have been identified and used In the analysis.

I no The system/train maintenance I 3 -

R3 SYSTEM/TRAIN I -

e no u/i-U' MAINTENANCE unavailabilities are derived based on UNAVAILABILITIES plant specific data.

(1)

COMMON CAUSE DA-08 The common cause failure 2 R7 FAILURE probabilities are referenced to PROBABILITIES acceptable data sources.(2) I I 4 -

DA-09 The common cause failure 3 R8 probabilities are realistic based on generic data source comparisons.

DA-10 Common cause groups to which the 3 R9 common cause failure probability applies have been derived based on sound judgment and are documented.

DA-1 1 Justification Is provided for treatment 3 R10 of common cause failure of on-site AC sources. Treatment Includes consideration of: (4)

- Design diversity

- Common maintenance crews

- Common I&C technicians

- Similarity of procedures

- Common fuel oil

- Common Heating/Cooling Designs DA-12 NUREG/CR-4780 (EPRI NP-5613 or 3 R9 equivalent) systematic approach used to provide plant specific grouping of similar system components for CCF treatment DA-13 Dominant contributors for sequences 4 R11 include MGL for more than 2 redundant trains (5)

UA-1'4 ruii intent or NURtICR¶-47io tI~rr 4 -ilz NP-5613 or equivalent) Included: -

Plant specific screening of common cause data.

Table 2 - Peer Review Results NOC-AE-04001813 Page 8 of 19

CRITERIA NEI-OD- ASME -NEI CRITERIA -'. PRA Contingent Related -Supporting

,-'CATEGORY 02 SR,..: 7, ' Grade PRA Grade Facts & Criteria and DESIG. IDESIG 'KY. a ,.. -Notes Obs.

UNIQUE DA-15 Documentation and bases are 3 R4 UNAVAILABILITIES provided for the failure probabilities OR MODELING from plant specific or generic sources ITEMS that do not fit into the basic event database, e.g.:(6)

- AC Power Recovery

- EDG Mission Time

- Repair and Recovery Model

- LOOP Given Transient

- BOP Unavailability

PipelTank Rupture Failure Probability

- ATWS Related RPS Failures

- RCP Seal Failure (for PWRs)

- % of time Pressurizer PORVs Blocked during operation (PWRs)

- PORV demand probability given an initiating event

- % of time SG PORVs or atmospheric dump valves blocked during operation DA-16 The unique unavailabilities are based 3 R4 on:(7)

- These failure probabilities are justified to the current state of the

. _ technology DOCUMENTATION DA-17 Reflects the process used 3 DA-1 8 Includes an independent review for the 3 R5 documented results DA-19 Provides the basis of the data a treatment and Is traceable to plant specific or generic analysis.

DA-20 The generic and plant specific 3 R6 databases are available for Inspection and use.

GUIDANCE HR-01 Describes the process used 3 R1 HR-02 Consistent with industry practices 2 - R2 HR-03 Sufficient detail provided for 3 Ri reproducing the evaluation PRE-INITIATOR HR-04 Pre-initiator Human Interactions (HIs) 3 R3 HUMAN ACTIONS were considered In the PRA HR-05 A systematic process Is used to 3 HR-01 R3 Identify the Pre-initiator Human Errors to be Included In the PRA (e.g.,

miscalibration of Instruments)

HR-06 Best estimate HEPs are used In the 2 R4 quantification of pre-Initiator HEPs for dominant contributorso HR-07 Those pre-initiator actions with the 3 R3, R5 possibility of adversely Impacting baseline CDF or LERF are Included In

__ _______the

____ quantification.

POST-INITIATOR HR-08 Post-initiator HIs were considered In 3 R6 HUMAN ACTIONS the PRA HR-09 A systematic process Is used to 3 R6 Identify the Post-Initiator Human Errors to be Included In the PRA.

HR- Assessment of plant procedures and 3 R6 10(3) plant specific operating experience are explicitly Included In the Identification I___ and quantification process for the HIs.

I____ I I I Table 2 - Peer Review Results NOC-AE-04001813 Page 9 of 19 CRITERIA :- NEI *ASME, NEI CRITERIA,.,-,. -PRA~; Contingent Related *Supporting

^.CATEGORY- --.02-. ' SR :'. ., . .Gdrade: PRA Grade Facts & Criteria and

- . .. DESIG :DESIG .. Obs. ':- Notes

i. -i F:

I

~,

,- I-,

i r ;- -~

HR-11 The symptoms available during the 3 R7 postulated accident sequence are evaluated and input into the HRA 4 4 Drocess. 4 4 HR-12 HEP values are Internally consistent 3 R8 4

within the PRA. 4

4 I 4

HR- Screening HEPs are internally NA 13(1) consistent within the PRA HR-14 Operator actions have been reviewed 3 R9 by the operating staff and their impact is Included in the HRA evaluation HR- Best estimate HEPs are used in the 3 15(1) quantification of dominant contributors. I I HR- Emphasis of the Human Reliability 2 R10 16(2) Analysis Is to Identify that the Hi Is folded correctly Into the model and that the HI:

- Reflects the procedures (EOPs &

AOPs)

HR-17 The performance shaping factors such 4 R11 as time available, time to perform, stress, complexity, etc. are Included in the quantification.

HR-18 The performance shaping factor for 3 time available for an action and the time required to take an action are developed on a plant specific basis.

HR-19 Time available for action is based on: 3 R12

- plant-specific T &H analysis HR-20 The time required to complete the 3 R13 actions is based on observation or operations staff input.

HR-21 The recovery actions are Included 3R14 systematically in the model HR-22 The models and analysis are 1 consistent with the operating procedures and training. l _

HR-23 Operator actions Including recovery 3 R6 are not credited unless a procedure Is available or operator training has Included the action as part of crew's training.

HR-24 Inter-unit cross-ties are only credited if _ NA, R15

_____ procedures and training are available. . .___

HR-25 Inter-unit cross-ties are accurately NA, R15 accounted for under conditions of outage for the other unit and special Initiating events.

DEPENDENCE HR-26 The dependence among human 3 R16 AMONG ACTIONS actions Is evaluated In the PSA process.

HR-27 Identification of sequences that, but for I R16 low human error rates In recovery actions, would have been dominant contributors to core damage frequency Is included as a test of modeling adequacy. Equivalent techniques may also be used.

DOCUMENTATION HR-28 Reflects the process used 3 I I R17 HR-29 Includes an Independent review for the R17 Idocumented results j __ I Table 2 - Peer Review Results NOC-AE-04001813 Page 10 of 19 CRITERIA . NEI-O-. ASME - .-NEI CRITERIA -- - PRA y Contingent Related Supporting CATEGORY 02 SR.

S --. ;,: - "Grade PRA Grade' Facts & Criteria and

DESIG DESIG ,-;,K,,;-* >L.' - -Obs. -:Notes.!

HR-30 Provides the basis of the HRA and is - NA, R17 traceable to plant specific or generic analysis.

GUIDANCE DE-01 Describes the process used 3 Ri DE-02 Consistent with Industry practices 3 Ri DE-03 Sufficient detail provided for 3 Ri reproducing the evaluation l INTER SYSTEM DE-04 The dependencies of the front-line 4 DE-05 R2 DEPENDENCIES system to support systems and support systems to support systems are identified.(i)

SYSTEM / DE-05 The dependencies of the support 4 DE-05 R3 INITIATOR systems and front-line systems to the DEPENDENCIES Initiating events are Identified METHODOLOGY DE-06 Support system and system to system 4 DE-06 R4 Interactions are treated In the event trees or linked fault trees. (See Element AS-6)

HUMAN DE-07 The human Interactions that can cut .. NA, R5 INTERACTIONS across system trains and can cause failure of multiple trains due to pre-Initiator and post Initiator human interactions (His) are Identified and documented. (See Element HR-26).

Examples include:

- Common cause miscalibration of similar sensors

- Operator procedure-based actions to terminate injection COMMON CAUSE DE-08 Similar components within a system NA, R7 are Included In a common cause group. (See Element DA-10)

DE-09 NUREG/CR-4780 methodology or - _ NA, R8 equivalent Is used to develop the component groups, OR NUREG/CR-4780 methodology or equivalent supported by plant specific operating experience is used to ensure grouping is adequate OR Full NUREG/CR-4780 Application or Its equivalent (See Elements DA-12 and DA-14)

SPATIAL DE-10 Spatial challenges that can result In 4 DE-06 R9 DEPENDENCIES dependencies among components are included In the model for:

- Flooding

- High temperature

- Inadvertent sprinkler operation

- Missiles (HPCVRCIC turbines for BWRs, turbine-driven EFW/AFW pumps for PWRs)

- Intake anomalies (e.g., Ice frazil, bio-fouling)

WALKDOWN DE-1 1 Specifically examines the spatial 2 R6, R1I dependencies that could affect the system or Intersystem reliabilities or

___ ._._._._ ___ Initiating events. . __ _

DOCUMENTATION DE-12 Reflects the process used; 2 DE-08 For Intemal Flooding, documentation reflects the process used to Identify flood sources, flood pathways, flood scenarios, and their screening and Table 2 - Peer Review Results NOC-AE-04001813 Page 11 of 19 CRITERIA- NEI0- ASME : NEICRiTERIA PR Contingent Related Supporting CATEGORY '.- 02 ' SR Grade'% PRA Grade- Facts & Criteria and

- DESIG DESIG ! '. . - Obs. - Notes Internal flood modeling DE-13 Includes an Independent review for the 3 Edocumented results I I DE-1 4 Provides the basis of the dependency 3 DE-08 treatment and Is traceable to plant specific or generic analysis.

AREAS AND SSC'S IF-04 IF-Al Plant Is divided Into physically 3 INVOLVED separate or combined flood areas generally on the same elevation.

- Presence of physical barriers Identified (e.g., walls, floors, dikes, watertight doors)

- Mitigation features (e.g., sumps and drains) identified Propagation pathways are Identified (e.g., open hatches, doors)

IF-05 IF-A2 SSC's located In each flood area are 4 IF-5 Identified, Including mitigating features (e.g., shielding) for SSC's which can challenge normal plant -operations.

- - .t t I IF-06 IF-A3, Plant walkdown Is performed to verify 3 IF-6 IF-A4 Information obtained from plant sources (e.g., drawings, operator interviews) spatial information, SSCs located In flood areas, and potential

+

flood sources In the areas. I 4 4

FLOODING IF-07 IF-B1 Potential sources for flooding water 3 .F-7 I SOURCES AND are Identified, Including:

MECHANISMS - Equipment In the area (e.g., pipes, valves, pumps) connected to fluid systems (e.g., circulating water systems, service water, feed water, reactor cooling water)

- Plant Internal sources (e.g., tanks or pools), and

- External sources (rivers or reservoirs) connected to the area throunh some svstam or structire.

IF-08 IF-B2 Flooding mechanisms from each 2

0 NIF-7 potential flooding water source are Identified (e.g., pipe break, human error In overfill of tanks, Inadvertent actuation of fire suppression systems, . IF-7 maintenance errors, any other events I_

__ which could release water in the area.) I I___ _I IF-09 IF-13 For each flooding mechanism the type 2 of water release and capacity are Identified:

- Breach (e.g., leak, rupture, spray)

- Flow rate of water, Capacity (e.g.,

gallons of water source), and

- Characterization of flow (e.g., spray, let. potential for pipe whip)

FLOODING I IF-10 I IF-1341 In each flood area the capacity of . NA, IF-10 SOURCESAND drains and sumps Is Identified. o MECHANISMS I Table 2 - Peer Review Results NOC-AE-04001813 Page 12 of 19

I. CRUTERIA - NEt-00-I ASME~ o NEI CRITERIA

PRA- Contingent Related Supporting

CATEGORY . . SR"' .Grade PRA Grade Facts & Criteria and

,; : ~~ I DESIG; DESIG . -,;.... . - Obs.- Notes

, i..-.

4- I 4 4-POTENTIAL IF-11 IF-Cl Propagation path for each flood source 2 IIF-11 FLOODING Is Identified including:

SCENARIOS - Normal flow path via drain lines

- Back flow through drain lines with failed check valves

- Pipe and cable penetrations

- Doors

- Stairwells

Hatchways

- Structural failure of doors or walls, and

- HVAC ducts.

IF-12 IF-C2 Plant design features or operator NA, IF-12 actions that can mitigate or terminate flood propagation are Identified and justified. Included availability of flood alarms, dikes, curbs, drains, sumps, shields, water-tight doors, and operator actions.

IF-13 IF-C3 Susceptibility of each SSC In the flood 3 IF-13 area to flood Induced failure mechanisms Is Identified and any exclusions are justified. Included are submergence, impingement, spray, pipe whip, resulting area dampness, etc.

POTENTIAlL IF-14 IF-C4 Flood scenarios are developed by NA, IF-14 FLOODIN( examining propagation paths.

SCENARIO' )1 Scenarios verified by walkdowns.

- (continued IF-15 IF-C5 Flooding scenarios are screened In 3 - IF-14 areas where:

- There Is no mitigating equipment modeled In the PRA or equipment failure does not cause an Initiating event

- There is no significant flooding potential, and

- There are adequate mitigating

.____ _._____ systems _

IF-16 IF-C6 Flooding scenarios are screened NA, IF-14 where the time to damage of safe shutdown equipment Is greater than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , if there is flood Indication In the control room and flood sources can be isolated, OR Mitigating action can be performed with high reliability for the worst flooding Initiator, OR No screening Is performed for scenarios that rely on operator action for mitigation

- + + + --

FLOODING IF-171 IF-D1 Flood initiators that challenge normal NA INDUCED plant operation have been Identified, INITIATING Including the potential for flooding EVENTS AND induced transient or LOCA.

THEIR FREQUENCIES Table 2 - Peer Review Results NOC-AE-04001813 Page 13 of 19

-. CRITERIA ' - NEI ASME i '- NEI CRITERIA . PRAf Contingent Related Supporting

.,-CATEGORY - 02 SR-.' f~> ; --- Grade PRA Grade Facts & Criteria'and

.DESIG' DESIG -. ,. No".

IF-18 IF-D2 Impact of plant specific Initiating event _ _ NA, IF-1 4 precursors and system alignments is not evaluated.

OR For Included flood-induced initiating events, the Impact of plant-specific initiating event precursors and system alignments, and alignments of supporting systems are addressed.

IF-19 IF-D3 Initiators are grouped only when: NA Events can be considered similar In terms of plant response, success criteria, and timing; OR Events can be subsumed Into a group and bounded by the worst case Impacts within the 'new' group.

Induced initiating events grouped only when:

Events can be considered similar In terms of plant response, success criteria, and timing, and the effect on the operability and performance of operators and relevant mitigating systems; or Events can be subsumed into a group and bounded by the worst case Impacts within the 'new' group.

continued IF-19 X X FLOODING IF-20 IF-D4 For multi-unit sites with shared NA INDUCED systems, a qualitative evaluation has INITIATING been performed to ensure that relative EVENTS AND risk significance of modeled SSC's is THEIR not distorted If multi-unit flood initiators FREQUENCIES are excluded (continued) OR For multi-unit sites with shared systems dual unit flood Initiators are treated and quantified explicitly.

IF-21 IF-D5 Flooding initiating event frequencies 2 IF-15 are determined by crieria specified in high level requirement IE-C of the ASME PRA Standard, generic data sources, and plant specific sources.

OR Flooding Initiating event frequencies are determined by criteria specified In supporting requirement IE-Cl-5 of the ASME PRA Standard and generic data enhanced by plant specific operating experience, or a combination of one of

___ __ __ the above with expert ludgement.

IF-22 IF-El The accident sequence results NA developed In AS are reviewed and modified as necessary to account for any flood-induced phenomena. I I QUANTIFICATION IF-23 IF-E2 Engineering calculations for flood rate, NA OF FLOODING time to reach vulnerable equipment, INDUCED and structural capacity of SSCs per ACCIDENT Success Criteria have been SEQUENCES performed.

IF-24 IF-E3 The systems analysis results obtained NA by following the applicable have been

- I I Table 2 - Peer Review Results NOC-AE-04001813 Page 14 of 19

- NEI-O-ASME jg; NE E ., ---PRCRITERIA Contingent Related Supporting CAEOY 0 S > ~ ~ s.. Grd RAGrade Facts & Criteria and DEDESIDEG - - 7, 'Obs., .Notes ,.

- .:M. .; - - ..- -

reviewed to include flood-induced failures.

IF-25 IF-E4 Additional data analysis performed to NA, IF-12 the applicable requirements for Human Reliability.

IF-26 IF-E5 Human reliability analyses were . .. NA, IF-12 performed Including PSFs for (a) Additional workload, (b) Uncertainties for event progression, and (c) Effect of flooding on mitigation, required response, and flooding-specific lob aids and training.

IF-27 IF-E6 Flood sequence quantification has 2 been performed, Including quantitative screening and both direct and indirect failures caused by flooding.

IF-28 IF-E7 Level 2 and LERF analyses were - NA reviewed to account for flood induced phenomena.

QUANTIFICATION IF-30 IF-F2 The following Intemal Flooding results - NA OF FLOODING are documented:

INDUCED - flood sources identified in the ACCIDENT analysis, any rules used to screen out SEQUENCES these sources, and the resulting list of (continued) sources to be further examined;

- flood areas used in the analysis and the reason for eliminating any of these areas from further analysis;

- propagation pathways between flood areas and any assumptions, calculations, or other bases for eliminating or justifying any of these propagation pathways;

- accident mitigating features and barriers credited in the analysis, the extent to which they were credited, and associated Justification;

- component fragilities and any associated assumptions or calculations used In the determination of the impacts of submergence, spray, temperature, or other flood-induced effects on equipment operability;

- screening criteria used in the analysis:

IF-30 - flooding scenarios considered, (continu screened, and the remaining ed) scenarios, as well as how the internal event analysis models were modified to model these remaining scenarios for

._ _the Internal flooding analysis .

GUIDANCE ST-01 Describes the process used 3 RI ST-02t _ Consistent with Industry practices _ _3 . _ _ . R_

ST-03 Sufficient detail provided for 3 RI renroducina the evaluation RPV CAPABILITY ST-04 Best estimate failure condition 3 R2 (ATWS) considered (ASME Service Level C Iused)

CONTAINMENT ST-05 Conservative estimate of failure NA. See L2 probability Is used OR Realistic estimate of failure probability Table 2 - Peer Review Results NOC-AE-04001813 Page 15 of 19

-* CRITERIA .- :-EILASME - NEI CRITERIA *- PRA r Contingent Related :Supporting CATEGORY . ..02 - -SR - Grade PRA Grade Facts & Criteria and DESIG DESIG ! .- . Obs.

,:jX Notes Is used based on detailed plant specific structural examination ST-06 Level 2 analysis considers multiple . NA, See L2 pathways from the containment REACTOR ST-07 Blowout panels considered _ NA BUILDING (for BWRs)

ST-08 Level 2 analysis considers multiple _ NA pathways from the reactor building PIPE ST-09 Conservative estimate is used NA, R3 OVERPRESSURE OR (ISLOCA) Generic realistic estimate Is used OR Plant specific realistic estimate is used FLOOD BARRIER ST-10 Internal flooding analysis considers -BNAR4 INTEGRITY flood barrier (e.g., doors) structural capability and features when these barriers are credited for limiting flood propagation DOCUMENTATION ST-1 1 Reflects the process used 3 RI ST-12 Includes an Independent review for the 3 R1 documented results ST-13 Provides the basis of the treatment 3 R1 and is traceable to plant specific or generic analysis.

GUIDANCE OU-01 Describes the process used 3 R11 OU-02 Consistent with Industry practices 3 OU-03 Sufficient detail provided for 3 reproducing the evaluation CODE OU-04 The base computer code and Its inputs 3 have been tested and demonstrated to Produce reasonable answers.(3), (4)

OU-05 The simplified model (cutset model) Is - NA demonstrated to produce reasonable results for typical applications. (2)

QU-06 Applications are not limited by the 3U-O1

_. _._____ _ _.. _capabilities

___.__ of the computer code. . . ..

SIMPLIFIED MODEL OU-07 The simplified model (e.g., solved NA

_+ . . . .. ....

cutset) limitations are _-

cleariv identified. . . - ,..

DOMINANT QU-08 The dominant cut sets or 3 R13 SEQUENCES/CUTS sequences(1) - Make physical sense ETS OU-09 Include common cause potential 3 Ri where appropriate OU-10 Include dependency among human 3 actions when multiple HEPs are In the E same cut set or sequence QU-1 1 Are not missing potentially dominant 3 QU-05 cut sets or sequences for similar plants. Possible reasons for differences Include:

(a) physical plant or procedural differences among plants; (b) documented assumptions; (c) detailed modeling or data to supplant assumptions

.... Table 2 - Peer Review Results NOC-AE-04001813 Page 16 of 19 CRITERIA NEI-OD- ASME NEI CRITERIA

- PRA Contingent Related Supporting

- CATEGORY a: ' 02 ;"--?SR - Grade PRA Grade Facts' & Criteria and DESIG DESIG . *-*tes b QU-12 Asymmetry: The model asymmetry Is 3 R2 well described In terms of:

- modeling

- plant support systems

- normally running equipment

- cross-ties to an adjacent unit.

QU-13 Asymmetry: Any modeling quantitative 3 R2 asymmetry (e.g., one train of dual-train system modeled as In-service, other in standby) Is documented and Is well understood so that applications affected by asymmetry can be determined.

QU-14 Circular logic can sometimes occur 3 R3 when using linked fault trees. The PSA process appropriately accounts for support system dependencies In a consistent fashion that avoids so-called circular logic. (5)

NON-DOMINANT OU-15 The non-dominant cut sets or 3 R14 SEQUENCES/CUTS sequences:

ETS(1) - Make physical sense OU-16 - Include common cause potential or 3R1 there are equivalent cut sets that do Include the common cause potential QU-17 - Include dependency among human 3 actions when multiple HEPs are In the same cut set or sequence l _

RECOVERY QU-18 Recovery actions credited In the 3 ANALYSIS evaluation are either proceduralized or have reasonable likelihood of success when the Technical support Center I Emergency Operations Facility are manned.

OU-19 Recovery actions that are Included In 3 R4 the quantification process are Included In all applicable sequences and cut sets QU-20 Transfers of sequences among event 3 R15 trees are treated explicitly.

TRUNCATION OU-21 There Is evidence of consideration of 3 R5 the effects of quantification truncation values on the results. (6)

QU-22 Example truncation values used In a 4 R6 base PSA are given. The screening truncation of events or failure modes retained In the model are as follows for screened out events:

< 0.00001 ^ CDF Base < 0.00001

__ LERF Base QU-23 The truncation values used in the 3 R8 system fault trees and accident sequences are sufficiently low to support their use In representative applications.

QU-24 There Is evidence of convergence 3 R5 towards a stable result CUTSET QU-25 If the fault tree linking approach Is NA COMPLEMENTS & used, 'delete' terms (cutset MUTUALLY complements) are used to account for EXCLUSIVE the successes In event sequences as EVENTS appropriate to assure that the correct cut sets are generated.

IQU-26 The quantification process Identifies 3 R7 Table 2 - Peer Review Results NOC-AE-04001813 Page 17 of 19 CRITERIA NEI-G ;.ASME. - .'-:..NEI

%f CRITERIA--, : -, PRA - Contingent Related Supporting

CATEGORY - -A02* SR ;- .-

Grade PRA Grade Facts & Criteria and DESIG DESIG , lObi.. Notes.

and deletes mutually exclusive cut sets.

UNCERTAINTY QU-27 A search Is performed for unique or 2 R9 unusual sources of uncertainty not present In the typical or generic plant analysis.

OU-28 If there are unusual sources of 2 R9 uncertainty, special sensitivity evaluations or quantitative uncertainty assessments are performed to support the base conclusion and future applications. _

OU-29 The capability to perform focused 3

s R12 sensitivities to support the PSA applications Is available. _ _

OU-30 A quantification of selected 3R10 uncertainties Is performed, or the Impact of the selected uncertainties on the final risk measures Is estimated.

RESULTS OU-31 The PSA results summary Identifies 3

SUMMARY

the dominant contributors. (7)

OU-32 Reflects the process used. 3 R11 QU-33 Includes an Independent review for the 3 documented results.

QU-34 Provides the basis and Is traceable to 3 plant specific or generic analysis.

GUIDANCE L2-01 Describes the process used 3 _ R1 L2-02 Consistent with Industry practices 3 R1 L2-03 Sufficient detail provided for 3 R1 reproducing the evaluation .

SUCCESS L2-04 The success criteria are Identified 3 R2 CRITERIA L2-05 The success criteria are supported by 3 R3 thermal hydraulic analysis, system capability evaluations, or Industry studies __

L2-06 The success criteria are judged 3 realistic L1IL2 INTERFACE L2-07 The link between the Level 1 and 3 R4 Level 2 Is sufficient and adequately documented to provide the transfer of Information from the Level 1 analysis to the Level 2 containment evaluation.

PHELNOMENA Lz-08s Tne phenomena tnat may control the Z Er!A CONSIDERED LERF radionuclide release (1),(3) - A characterization are Included. 4

L2-09(4) (PWRs): If plant specific features are 3 R5 not consistent with those assumed In Owners Group SAMG analyses, the L2 model addresses any plant-specific phenomena that may affect accident 4 ______

management actions and plannina. 4 4 L2-10 The phenomena that may Influence 2 R6 applications are Included.

__ _ _ __ .. 4 HEPs AND SYSTEM 12-11 System performance has been 3 PERFORMANCE evaluated to account for the adverse conditions that may be present during the

-' -- core melt

- -' - -' progression

'--- _ response.

' - D'- - - - -- -- - - --- - -

L2-12 Success of human actions has been 3 evaluated to account for the adverse conditions that may be present during the core melt progression response.

Table 2 - Peer Review Results NOC-AE-04001813 Page 18 of 19

CRITERIA . NEI -ASME NtCIEA. - PA Coinent Related Supporting CATEGORY.' 02 ' SR.. 'Grade PRA Grade Facts & Criterla'and

.DESIG DESIG O s.;. Notes; L2-13 Containment and system functional 3 12-03 failures are treated realistically for dominant contributors CONTAINMENT 12-14 Containment capability Is analyzed 3 R7 CAPABILITY under severe accident conditions for ASSESSMENT Its survivability 12-15 Both static and dynamic effects are 3 R7 Included (2), (3) 12-16 All postulated failure modes Identified 3 R3 by IDCOR or NRC Staff In NUREG-1150 are considered (2), (3) 12-17 For Ice Condenser and BWR Mark III -. NA containments only 12-18 Both leakage and large failures are 3 R8 Included In the analysis 1-2-19 Containment failure modes are 3 12-03 R7 treated realistically In the analysis 12-20 The containment analysis Is: 2

- Conservative ENDSTATE 12-21 The Level 2 end states support the 3 DEFINITION applications currently envisioned. _

LERF DEFINITION 12-22 The LERF definition Is consistent with 3 the following guidance, and is documented:

- PSA Applications Guide or other Owners Group-specific definitions (5) 12-23 - The LERF definitions use Emergency 3 Action Levels (EAL) bases If required; and the EAL bases are documented. ___

CONTAINMENT 12-24 The CETs: 3R8 EVENT TREES - Include all the functional events (CETs) required to meet a safe stable condition

- Include the phenomena cited under phenomena.

12-25 The CETs: 3

- Include the systems and HEPs necessary

- Are consistent with the EOPs

- Include reasonable recovery actions.c DOCUMENTATION 12-26 Documentation reflects the process 3 R9 used 12-27 Includes an independent review for the 3R10 documented results 12-28 Provides the basis of the containment 3R9 performance analysis and the analysis Is traceable to plant specific or generic analysis.

GUIDANCE MU-01 Describes the process used 3 R1 MU-02 X Consistent with Industry practices _______ R2 MU-03 Sufficient detail provided to update the 3 R3 evaluation _

Table 2 - Peer Review Results NOC-AE-04001813 Page 19 of 19

-CRITERIA .- .NEI ~NE100-- -ASME NEI CRITERIA - PRA.- Contingent Related Supporting CATEGORY ; 02 .- SR -Grade. PRA Grade Facts & Criteria and DESIG DESIG 11' l L' I

--Obs. .- Notes

, : . _' l..

4 -- t 4- - I I INPUT- MU-04 Each of the following Information 3 R4 MONITORING AND sources Is part of the PSA update COLLECTING NEW process for monitoring new Information INFORMATION (2) associated with the following:

- Operational Experience

- Plant Design

- New Maintenance Policies

- Operator Training Program

- Technical Specifications

- Revised Engineering Calculations

- Emergency and Abnormal Procedures

- Operating Procedures

- Emergency Plan

- Accident Management Programs

- Industry Studies ..~4 . 4 MU-05 Plant specific data is included for 3 R4 quantitative reevaluation.

MODEL CONTROL MU-06 The computer models of the PRA are 4 MU-04 R5 stored In a controlled manner. This also applies to sensitivity cases that may be performed to support a specific application.

COMPUTER CODE MU-07 Computer code controls are formalized 4 MU-04 R5 CONTROL to ensure that the effect on the PRA of changes to these codes are understood and addressed if appropriate PRA UPDATE MU-08 A process Is in place to maintain the 3 R1, R2, R3, PRA. The PRA update model process R4, R7 consists of the elements Identified and the steps In the process. The model update process consists of the following:

- Identification of Affected Model Elements

- Modification of PRA Models

- Requantification of PRA Models

- Evaluation of Results

- Re-Evaluation of Past PRA Applications e !

MU-09 The plant has defined a fixed update 3 R6 schedule or a reasonable criteria upon which to base the need for an update.

EVALUATION OF MU-10 The PRA results are evaluated by 3 R2 RESULTS knowledgeable personnel before the

.___ ._ _ results are used. . _ .

RE-EVALUATION MU-11 Past PRA Applications are evaluated - _ NAR7 OF PAST PRA qualitatively to assure that the APPLICATIONS (3) conclusions remain valid. l _

MU-12 Past PRA Applications that may be 3R7 affected by the latest information and update are re-performed.

DOCUMENTATION MU-13 Documentation reflects the process 3 R8 used MU-14 Includes an independent review for the 3R8 documented results MU-15 Provides the basis of the update 3 R8 process and the results are traceable to specific changes In design, procedures, training, or operating experience.

v t e Letter Sequence Other
TAC:MC3923, (Withdrawn, Closed) TAC:MC3924, (Withdrawn, Closed)
Initiation Acceptance... Administration Questions Answers Results Withdrawal Other: ML043070448, ML061640018, NOC-AE-04001813, Technical Adequacy of the South Texas Project Probabilistic Risk Assessment
MONTHYEAR2006-06-13 [Table View]

NOC-AE-04001813, Technical Adequacy of the South Texas Project Probabilistic Risk Assessment

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