ML20212H223

Summary of 990831-0901 Meeting with STPNOC in Rockville, Maryland Re 990713 Request for 14 risk-informed Exemptions to Exclude safety-related LSSC & non-risk Significant Components from Special Treatment Requirements

ML20212H223
Person / Time
Site:	South Texas
Issue date:	09/24/1999
From:	Alexion T NRC (Affiliation Not Assigned)
To:	NRC (Affiliation Not Assigned)
References
TAC-MA6057, TAC-MA6058, NUDOCS 9909300245
Download: ML20212H223 (61)

Text

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p 4 UNITED STATES g j NUCLEAR REGULATORY COMMISSION WASHINGTON, D.C. 20666 0001 September 24, 1999 LICENSEE: STP NUCLEAR OPERATING COMPANY (STPNOC)

FACILITY: South Texas Project, Units 1 and 2 (STP)

SUMMARY

OF AUGUST 31 - SEPTEMBER 1,1999, MEETING ON STPNOC'S PROPOSED EXEMPTIONS TO EXCLUDE CERTAIN COMPONENTS FROM THE SCOPE OF SPECIAL TREATMENT REQUIREMENTS REQUIRED BY THE REGULATIONS (TAC NOS. MA6057 AND MA6058)

On August 31 and September 1,1999, the NRC staff met with STPNOC to discuss STPNOC's July 13,1999, request for 14 risk informed exemptions to exclude safety related low safety / risk significant (LSS) and safety-related non risk significant (NRS) components from the scope of special treatment requirements required by the regulations. These exemption requests utilize risk-informed insights from STPNOC's probabilistic risk assessment (PRA) and are an extension of STPNOC's risk-informed Graded Quality Assurance (GOA) program that was approved by the NRC on November 6,1997. Meeting attendees are listed in Attachment 1. A handout provided by STPNOC is in Attachment 2. A preliminary list of discussion topics that was generated by the staff is in Attachment 3. .

The licensee met with the staff in two separate forums. One forum was a working-level meeting that took place all day on August 31,1999, and resumed during the moming of September 1, 1999. The other forum was a management-level meeting where the licensee met primarily with NRC's Risk-Informed Licensing Panel (RILP) during the morning of September 1,1999. Both meetings are summarized below.

Workina-Level Meetina 1

Topics that were discussed included STPNOC's GOA background, approach and t implementation; the scope and basis of the exemption requests; and NRC staff questions and I comments. STPNOC provided background information on the evaluation of its risk-informed l processes and how the results from its GOA program will be used to support the exemptions from special treatment provisions. STPNOC provided a vision for utilizing risk insights to make [ l adjustments in the conduct of station operations so that resources are appropriately allocated i based on the relative safety significance of plant equipment / activities. STPNOC provided a -

perspective on how engineering staff would carry out their function in the risk-informed environment. STPNOC's overall perspective is that it has a robust process to determine equipment safety significance and that safety related LSS/NRS items should be thought of as h[

j being equivalent to non-safety related and, therefore, exemptions are warranted.

There was a substantial amount of questions, comments, and discussion from the NRC staff, both during STPNOC's initial presentation and also during the interactive discussions on the set of preliminary review questions and comments (Attachment 3). Some of the major points from NRC staff's perspective are as follows:

1. There is a need to sharpen the focus of the submittal with respect to exactly what items are in and cut of scope for proposed regulatory relief. For example, the staff did not _

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9909300245 990924

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2 understand that relief is sought only for " tagged" equipment, which excludes items such as electrical cable and piping.

2. The submittal requests complete relief from the cited regulatory requirements, when in actuality STPNOC appears to only need relief from portions of the requirements, and in some cases the regulations provide for graded applications already. STPNOC was asked to identify the "real" exemptions it needs (i.e., General Design Criterion 1 and Appendix B allow controls to be applied based on safety significance already, so why is an exemption needed?). STPNOC expressed the point that it wants to have licensing approval to address potential inspection issues.

3. The submittal did not sufficiently describe how the program will treat non-safety-related high safety significant equipment for assessing the need for enhanced controls. In order to arrive at findings of risk neutral or risk benefits, the staff will need to understand these controls.

4. There is uncertainty about what commercial-grade controls will be invoked to assure that safety-related LSS/NRS equipment is reasonably assured of being able to perform its design / safety functions.

5. The staff indicated that safety-related LSS/NRS components are still considered safety-related and serve some safety function. Therefore, relaxations of controls or alternative controls may be more appropriate than exemptions to the regulations.

6. The staff indicated that a substantial amount of additionalinformation needs to be submitted on the docket in order for the staff to be able to complete its review of the application and craft a safety evaluation.

The staff found the licensee's presentation to be very informative and thanked the licensee for its presentation and candid participation during the question and comment portion of the meeting. STPNOC staff also indicated that they appreciated the staff comments and indicated that they plan to supplement the application to address the NRC's comments. In addition, STPNOC invited the staff to visit the site for the staff to perform an in-depth review cf the systems that have had their components risk ranked in the Risk Significance Basis Documents.

RlLP Meetino STPNOC provided an overview of its exemption request and its relationship to rulemaking initiatives underway by the NRC to risk inform 10 CFR Part 50. STPNOC indicated that it wanted the staff to approve its process for determining the safety significance of plant equipment so that special treatment controls could be applied commensurate with the safety significance of the equipment. STPNOC proposed to provide the staff with feedback at a 6-month interval following approval of the exemptions, which would include implementation details of the changes made under its proposed program.

The NRC staff pointed out that it will be necessary for STPNOC to provide an explicit description of the alternative design codes and standards that will be relied upon to ensure that safety related LSS/NRS equipment can reasonably be assumed to be capable of performing its intended functions. The staff also stated the need for an enhanced description (i.e., in revised

\- . . . . -

1 3

updated final safety analysis report content) of the process that would be implemented by STPNOC, along with a proposed regulatory change control for the process description (i.e.,

50.59 or 50.54(a)-type controls).

The NRC staff expressed their appreciation for STPNOC support of the meeting and also indicated that future interactions with the RILP were likely as the RlLP will be providing management level integration of the exemption review along with other staff initiatives in the risk-informed arena.

ORIGINAL SIGNED BY Thomas W. Alexion, Project Manager, Section 1 Project Directorate IV & Decommissioning Division of Licensing Project Management Office of Nuclear Reactor Regulation Docket Nos. 50-498 and 50-499 Attachments: As stated (3) j cc w/atts: See next page DISTRIBUTION Hard Conv E-MAIL Docket File J.Zwolinski/S. Black S. Turk PUBLIC S.Richards D.Lange (DJL)

PD IV-1 Reading R.Gramm  ;

T. Alexion C. Jamerson OGC- NRC Meeting Participants  !

ACRS *See previous concurrence To receive a copy of this document, indicate "c" in the box OFFICE PDIV-1/PM C PDIV-D/LA C PDIV-1/SC (., *PDIV/D NAME- TAlexion:db CJamerson SDembeck for f (

RGra}rpi q(/) SRichards O/

DATE 799 9/d/99 h 99 09/22/99 Ee a copyWoc in e ox -

OFFICE *DD/ DRIP *DD/DIPM *D/DE, *D/DSSA NAME SNewberry FGillespio JStrosnider GHolahan DATE 09/21/99 09/23/99 09/21/99 09/22/99 10 receive a copy of Inis document, indicate "C"in the box OFFICE *DD/DLPM NAME SBlack DATE 09/21/99 DOCUMENT NAME: G:\PDIV-1\SouthTexas\MTSA6057.wpd OFFICIAL RECORD COPY.

P, i

l South Texas, Units 1 & 2 i

cc:

Mr. Cornelius F. O'Keefe Mr. J. J. Sheppard, Vice President Senior Resident inspector Engineering & Technical Services i U.S. Nuclear Regulatory Commission STP Nuclear Operating Company P. O. Box 910 P. O. Box 289 Bay City, TX 77414 Wadsworth,TX 77483 1

A. Ramirez/C. M. Canady S. M. Head, Supervisor, Licensing Cityof Austin Quality & Licensing Department l Electric Utility Department 'STP Nuclear Operating Company  !

721 Barton Springs Road P. O. Box 289 Austin, TX 78704 l

- Wadsworth, TX 77483 Mr. M. T. Hardt Office of the Governor  !

Mr. W. C. Gunst ATTN: John Howard, Director i City Public Service Board Environmental and Natural P. O. Box 1771 Resources Policy San Antonio, TX 78296 P. O. Box 12428 )

Austin, TX 78711 l Mr. G. E. Vaughn/C. A. Johnson l Central Power and Light Company Jon C. Wood P. O. Box 289 Matthews & Branscomb MailCode: N5012 One Alamo Center l Wadsworth, TX 74483 - 106 S. St. Mary's Street, Suite 700 San Antonio,TX 78205-3692 INPO Records Center A-thur C. Tate, Director 700 Galleria Parkway Division of Compliance & Inspection Atlanta, GA 30339-3064 Bureau of Radiation Control Texas Department of Health Regional Administrator, Region IV 1100 West 49th Street U.S. Nuclear Regulatory Commission Austin, TX 78756 611 Ryan Plaza Drive, Suite 400 e

' Arlington, TX 76011 Jim Calloway Public Utility Commission of Texas D. G. Tees /R. L. Balcom Electric Industry Analysis Houston Lighting & Power Co. P. O. Box 13326 P. O. Box 1700 Austin, TX 78711-3326 Houston, TX 77251 Mr. William T. Cottle Judge, Matagorda County President and Chief Executive Officer Matagorda County Courthouse STP Nuclear Operating Company 1700 Seventh Street South Texas Project Electric Bay City, TX 77414 Generating Station P. O. Box 289 Jack R. Newman, Esq. Wadsworth, TX 77483 Morgan, Lewis & Bockius 1800 M Street, N.W. l Washington, DC 20036-5869 May 1999

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l MEETING WITH STP NUCLEAR OPERATING COMPANY (STPNOC) ON PROPOSED l EXEMPTIONS TO EXCLUDE CERTAIN COMPONENTS FROM THE SCOPE OF SPECIAL TREATMENT REQUIREMENTS REQUIRED BY THE REGULATIONS August 31 - September 1,1999 L

Name Oraanization S. Rosen STPNOC R. Grantom STPNOC  ;

S. Thomas STPNOC W. Harrison STPNOC G. Schinzel STPNOC K. Work STPNOC A. Moldenhauer STPNOC S. Frantz Morgan, Lewis & Bockius R. Huston Licensing Support Services A. Heymer NEl B. Christie - Performance Technology D. Steiifox McGraw-Hill R. Gramm NRC R.Fretz NRC T. Alexion NRC J. Williams NRC P.Y. Chen NHC K. Manoly NRC D. Terao NRC Y L. Li NRC S. Lee NRC G. Parry NRC C. Gratton NRC J. Calvo NRC E. McKenna NRC W. Bateman NRC R. Young NRC T. Scarbrough NRC J. Knox NRC D. Thatcher NRC F.Ashe NRC H. Garg NRC D. Dorman NRC M. Rubin NRC P. Balmain NRC M. Shuaibi NRC R.Lobel -NRC l (continued on next page) ATTACHMENT 1 l

m MEETING WITH STP NUCLEAR OPERATING COMPANY (STPNOC) ON PROPOSED EXEMPTIONS TO EXCLUDE CERTAIN COMPONENTS FROM THE SCOPE OF SPECIAL TREATMENT REQUIREMENTS REQUIRED BY THE REGULATIONS August 31 - September 1,1999 (continued from previous page)

Name Oraanization K. Heck NRC S. Richards NRC W. Jones NRC E. Sullivan NRC G. Georgiev NRC S. Ali NRC M. Snodderly NRC T. Reed NRC S. Black NRC J. Strosnider NRC F. Gillespie NRC T. King NRC R. Barrett NRC C. A. Carpenter NRC P. Kadambi NRC G. Holahan NRC J. Moore NRC S. Newberry NRC o

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l POTENTIAL DISCUSSION TOPICS FOR 8-31-99 STP RXEMPTION REQUEST MEETING l 1

1 Comment #1:

The licensee's submittal of July 13,1999 does not list RPS, ESFAS and other instrumentation system for downgrading the requirements. Please confirm that no component from these i systems are downgraded. Section 7.1 of the submittalidentifies other systems for which i downgrading review has been completed. In order for staff to understand the process used for downgrading, we would like licensee to walk us thru the following systems using P& ids and logic diagrams to justify the categorization used to downgrade the requirements:  ;

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1. Auxiliary Feedwater System

2. Emergency Diesel Generator

3. Safety injection Also provide us the list of components with the categorization for these systems.

Comment #2:

Potential To, es for STP Exemotion Meetina

1. Efft on design basis accidents

a. Are any LSS/NRS components relied upon in FSAR accident analyses (full .

scope, from AOO and DBA) I

b. What is the effect of assuming that LSS/NRS components with reduced special treatment requirements can not be credited in FSAR analyses?

c. How are functional requirements for LSS/NRS components controlled under 10 CFR 50.59?

d. Staff should be aware that there is probably already some set of components where items (a) and (b) are relevant under the existing GOA program. That is, there are probably some cornponents which do not receive full QA, but are nonetheless relied upon to demonstrate acceptable accident analysis performance.

2. How will special treatment requirements for NSR items added to scope be identified and controlled?

3. How does the risk ranking process address spatial relationships such as seismic interactions or fire?

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Comment #3:

Reolacement of ASME Components with Non-ASME Comoonents

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Section 4.2.1 of STPNOC's July 13,1999, exemption request states in part that: )

.The requested exemption from Section 50.59 applies not only to UFSAR special treatment provisions required by the regulations, but also to other UFSAR special treatment provisions that are not directly required by the regulations. For example, under 10 CFR 50.55a, the installation requirements in Section lli of the

- ASME Code are only applicable technically to holders of construction permits and not to holders of operating licenses. However, in general, the UFSAR commitments to Section ill of the ASME Code are not limited to construction, but also apply to replacement components installed during operation. For safety-related LSS and NRS components, the requested exemption would enable STP to replace an ASME component with a non ASME component without the need to perform a detailed 50.59 evaluation or seek prior NRC approval. Similarly, STP would not need to perform a detailed 50.59 evaluation or seek prior NRC approval for UFSAR changes involving other inspection, maintenance, testing qualification, and QA provisions applicable to safety-related LSS and NRS components that are not specifically covered by the regulations '

(for example, such changes could involve the testing provisions for motor-

operated valves or air operated valves, and inspection of snubbers).

10 CFR 50.55a requires that systems and components of boiling and pressurized water-cooled nuclear power reactors must meet the requirements of the ASME Boiler and Pressure Vessel Code,Section XI. The ASME Code Section XI rules become effective when the requirements of the ASME Construction Code (Section lli rules) have been satisfied. Division 1 of Section XI provides rules for inservice inspection of light water cooled nuclear power plants. Subsections IWB, IWC, IWD, and IWE specify requirements for Code Class 1,2,3, and MC components, respectively. Subsection IWA provides general requirements that apply to the other

- subsections. Within each suosection Article 4000 addresses repair / replacement activities. (In

. earlier versions of the Code, Article 7000 addressed replacement requirements separately from repair requirements.) In general, these articles require that the materials for repair and replacement comply with the requirements to which the original compon6nt or part was ,

constructed (or to a later Edition of Section ill). Therefore, components which are part of the reactor coolant pressure boundary must meet the requirements for Class 1 components in Section lli of the ASME Boiler and Pressure Vessel Code. Components classified Quality l

Group B and C must meet the requirements for Class 2 and 3 components in Section Ill, j respectively. Guidance for quality group classifications of components which are to be included 4 l

in the safety analysis reports pursuant to 6 50.34(a) and 9 50.34(b) may be found in Regulatory Guide 1.26 and in Section 3.2.2 of NUREG-800," Standard Review Plan for Review of Safety Analysis Reports for Nuclear Power Plants." In categorizing components, Regulatory Guide 1.26 and NUREG-800 use the terms safety related and important to safety. Therefore,

, re-defining the terms " safety-related" and "important to safety" to make them more L risk-informed in the regulation would not, in itself, allow the licensee to repair or replace low or l non-risk significant components with non-Code components. Changes would need to be made to Regulatory Guide 1.26, other regulatory guidance documents, and to the licensee's USAR i

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and procedures. In order to allow licensees to repair or replace ASME Code components with non-Code components, the NRC would have to grant licensees an exemption from the 10 CFR 50.55a requirements to meet the ASME Boiler and Pressure Vessel Code or the NRC would have to specifically authorize alternative repair / replacement activities (i.e., pursuant to 10 CFR 50.55a(a)(3)(i)).: Then, and to the extent 10 CFR 50.59 is revised to make it risk-informed, licensees would be able to change their FSARs to allow the replacement of safety-related but non-risk-significant components with non-Code class components without prior NRC approval.

l The staff concludes that STPNOC needs to request an exemption from 10 CFR 50.55a or provide an alternative pursuant to 10 CFR 50.55c(a)(3) in order to replace Code class 2 and 3 components that are determined to be "non-essential" or "non-Risk-informed Safety Class" with non-Code components. Because ASME Code Class 1 components are part of the reactor coolant pressure boundary, they represent one of the principal fission-product barriers.

Consistent with the defense-in-depth philosophy, tne staff does not recommend risk-informed changes to the repair / replacement activities for reactor coolant pressure boundary components.

Similarly, the staff does not recommend risk-informed changes to the repair / replacement activities for containment components classified as MC or CC.

I Comment #4:

1. The description of the scope of the exemption to Appendix J is not match the i justification for the exemption.- The justification is more detailed / specific. What is the licensee's intent: to consider all isolation valve configurations or only the closed system in the justification?

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2. Is " closed system" used in the GDC 57 sense as defined in SRP 6.2.4, i.e., missile i protected, seismic, etc.

3. Appendix J is written in terms of cumulative leakage rates. If only some systems are I

excluded, how is a total or cumulative leakage rate determined?

Comment #5: *

1. Provide examples of how typical motor control center, battery, MOV, RPS logic cards would be processed thru the safety significance determination process. Provide specific discussion about the classification of the AFW pump /turoine classification.

2. Discuss conceptually how equipment associated with the control room ventilation system would be processed through the safety significance determination process.

3. STP should re-consider the justification of the usefulness of normal plant operational feedback mechanisms to demonstrate the functionality of LSS/NRS equipment with respect to special treatment provisions. Absent the occurrence of a situation that '

challenges the equipment qualification, no information would be gained about whether the equipment remains qualified (need a seismic event).

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4. Discuss examples of how commercial practices would be implemented when procuring a re,:'acement LSS/NRS motor driven pump for design / procurement / vendor fabrication / receipt / installation, post mod verifications, and maintenance.

5. Discuss any potential issues with respect to eventual plant license renewalif exemptions were to be granted. Normally renewal credits many of the programs where exemptions are being requested.

6. Why is full exemption sought for Appendix B criterion IV which requires measures to assure that applicable regulatory, design bases, and other requirements are satisfied in documents for procurement of material? STP stillis counting on purchase requirements to ensure availability of replacement components to function under design conditions (Attachment 1, page 7, section 4.1.2). This seems inconsistent.

7. Appendix B controls at level of 18 criterion form an integrated management control process that provides assurance that equipment will perform satisfactorily in service.

The high level program description in Appendix B is very similar to those types of '

controls imposed commercially (ISO 9001 program). It is not clear how a truncation of 15 Appendix B criterion will yield confidence that equipment will function as intended.

Why wasn't a grading of application controls for those 15 criterion proposed rather than full exemption?

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Comment #6:

1. Q50.12(a)(2)(ii) states: " Application of the regulation in the particular circumstances would not serve the underlying purpose of the rule or is not necessary to achieve the underlying purpose of the rule" The underlying purpose of the EO rule (10CFR50.49) is to ensure the protection, independence, and reliability of safety related (or risk significant HSS and MSS) structures, systems, and components. At STP, the application of the EQ rule to safety related (and also to non-safety related but important to safety) non-risk significant components (i.e., LSS or NRS components) serves this purpose and is necessary to meet independence and single failure requirements of Criterion 17 of 10CFR Part 50, Appendix A, for safety related (or risk significant HSS or MSS) structures, systems, and  ;

components. Explain why the application of the EO rule to these non risk significant j components does not serve this underlying purpose or is not necessary to ensure the l protection, independence, and reliability of safety related (or risk significant HSS and MSS) structures, systems, and components. I

2. G50.12(a)(2)(iv) states: "The exemption would result in benefit to the public health and i safety that compensates for any decrease in safety that may result from the grant of the  !

exemption".

l STP's exempwn request indicated or implied the following:

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Components classified Non-safety related and judged to be risk significant (i.e.,

l HSS or IUSS) will receive augmented controls; Components classified safety related and judged not to be risk significant (LSS or NRS) will be maintained sufficient to meet their associated system performance levels; and i

Components classified non-safety related and considered important to safety )

because their failure may affect structures, systems, or components classified j safety related and judged risk significant (i.e., HSS or MSS) may (or may not) be i maintained sufficient to meet their as'sociated system performance levels. I

• ' In STP's estimation based on its experience with the risk-informed categorization ;

process, do the increased controls for MSS and HSS non safety-related components continue to result in a net decrease in risk when the increased risk from non-safety related but important to safety components are considered in addition to the increased risk from safety related but not risk significant 1 components.

Explain how decreased risk (or maintaining risk level) is related to (and 1 compensates for) the decrease in safety that may result when the protection, independence, and reliability of safety related (or risk significant HSS and MSS) structures, systems, and components is either eliminated or reduced by some j level given that safety related (and also non-safety related but important to safety) non-risk significant structures, system, and components no longer meet special treatment requirements for qualification, quality assurance controls, maintenance requirements, and inspection and testing.

3. $50.12(a)(2)(vi) states: "There is present any other material circumstance not i

considered when the regulation was adopted for which it would be in the public interest i to grant an exemption."

Section 5.2.4.1 of the exemption request states:"...the changes in the special treatment i requirements do not lend themselves to a quantitative risk assessment, because the l relationship between the special treatment requirements and equipment performance has not been firmly established." Explain how risk can be considered a new material circumstance when changes in the special treatment requirements (e.g., exemption from 50.49 requirements) do not lend themselves to a quantitative risk assesement.

4. Section 5.1 of the exemption request states that "The exemption would apply only to safety-related LSS and NRS components, but would include non-safety related LSS and NRS components as it applies to the Maintenance Rule." Does this statement mean that exemption to 50.49(b)(2) and 50.49(b)(3) is not being requested? It appears that structures, systems, and components that are classified non safety related but important to safety are not included as part of the exemption. Provide clarification.

5. Section 5.2.2 of the exemption request states that " Design provisions for redundancy, I L diversity, and independence will be maintained. Given that (1) the underlying purpose of l- the EQ rule (10CFR50.49) is to ensure the protection, independence, and reliability of E

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safety related (or risk significant HSS and MSS) structures, systems, and components and (2) exemption is being requested from this underlying purpose, clarify how design provisions for independence are being maintained when exemption from these design provisions is being requested.

6. Define the UFSAR requirements for physical and electrical protection of safety related (or risk significant HSS and MSS) structures, systems, and components from electrical and/or mechanical failure of safety related (and also non-safety related but important to safety) non-risk significant structures, system, and components.

i 7. In section 5.2.3 of the exemption request, it is indicated that STP does not expect the need to change any of the safety analyses in the USFAR. The USFAR safety analysis j assumes protection will be provided safety related (or risk significant HSS and MSS) ,

structures, systems, and components. As part of the proposed exemption, this i l

protection provided by qualification, quality assurance controls, maintenance requirements, and inspection and testing (assumed by the USFAR safety analysis) is i being eliminated or reduced. Explain or clarify why STP does not expect the need to I change any of the safety analyses in the USFAR.

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8. In section 5.2.3 of the exemption request, STI' indicates that as a result of their process for categorization of a component, failure of safety relates (and also non-safety related but important to safety) non-risk significant structures, system, and components would not cause the failure of a significant safety function.

i Provide the methodology which assures that failure of one (or more) non-risk significant components will not cause failure (i.e., loss of function) of safety related (or risk significant HSS and MSS) structures, systems, and components when needed.

Define the criteria which will be applied as part of the categorization process which willlimit the probability for failure of safety related (or risk significant HSS and MSS) structures, systems, and components when needed due to failure of one (or more) non-risk significant components.

Provide results of analysis that demonstrates the probability for failure (or loss of function) of safety related (or risk significant HSS and MSS) structures, systems, l

and components will not increase given that safety related (and also non safety related but important to safety) non-risk significant structures, system, and i

components no longer meet special treatment requirements for qualification, quality assurance controls, maintenance requirements, and inspection and l testing.

Comment #7:

EElB analysis of the STP document that addresses risk informing 10CFR50.49 (EQ Rule) has raised the following fundamental issue: *Is the current EO rule amenable to being risk-informed l consistent with current regulatory policy?"

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l The EO task action plan (TAP) was established to evaluate and resolve EQ concerns. A risk I assessment was one task noted in the EO-TAP which along with the other tasks the staff i

periodically provided status reports to the Commission. In a status report dated November 15, '

1996, the staff reported to the Commission the results of a preliminary risk assessment scoping study. These staff reported results indicated EO equipment failures in a harsh environment could significantly effect core damage frequency (CDF). The results from this risk assessment scoping study indicated that if the reliability of EO components was reduced by the presence of harsh environments, the probability of CDF estimates could increase significantly. The staff further reported to the Commission in the November 1996 status report that a follow-up study performed by the Argonne National Laboratory concluded that available information and reliability data for components in harsh environments cannot adequately support a probabilistic risk assessment (PRA) for EO. Based on the findings from the initial scoping and the Argonne National Laboratory studies, the staff recommended that no further work be performed to assess the impact on CDF due to EO equipment.

In addition, although not required by Generic Letter (GL) 88 20, all utilities chose to perform a PRA in response to this GL. These PRAs were not performed to spc',,ric standards and no requirements were specified for the data or models. The level of detail contained these PRAs varies widely across the operating plant units. In general, the existing PRAs model system level responses versus component level responses and were not performed to support risk-informed performance based regulations. The EO rule clearly focuses on component level responses and provides defense in-depth to address common cause equipment failures since EO has clear common cause implications. To upgrade existing PRA models to incorporate component level responses will require significant amounts of licensee resources. Thus, within the current framework, it is unclear how risk informed means/ techniques will truly establish those components with little or no safety implications regarding EO.

The above not withstanding, we need to clearly establish a response to the above fundamental issue prior to entertaining the STP proposal on the equipment governed by the EO rule.

Comment #8:

EMEB Issues: '

1. MOV/AOV and Snubber Testina Proaram Chances: The South Texas licensee states that an exemption is not required to remove low-risk safety-related MOVs and AOVs from plant-specific programs. With respect to MOVs, the NRC issued Generic Letters 89-10 and 96-05 under 10 CFR 50.109 as a requirement for compliance with the NRC regulations including 10 CFR 50, Appendix A, Criteria 1 and 4. If the licensee removes safety-related MOVs from the scope of those programs, the licensee must demonstrate that the MOVs continue to be capable of performing their safety related functions. With respect to AOVs, the licensee will continue to be required to satisfy the NRC regulations in ensuring adequate confidence in the capability of safety-related AOVs in performing their safety functions. The licensee may establish less rigorous performance criteria for its low-risk safety-related MOVs and AOVs provided adequate confidence is maintained in their capability.

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L in addition, STP states that its snubber testing program will be modified to remove ,

, safety-related LSS and NRS snubbers from the scope of the program. The licensee has  !

not explained how safety-related snubbers can be categorized as LSS or NRS since I snubbers protect the safety function of a system and not necessarily the function of a specific component. Therefore, it is unclear whether these snubbers can be removed

! from tha scope of the snubber testing program.

2. Exemotion from GDC 2. GDC 4. and Part 100. Aooendix A.VI: The STP licensee is requesting an exemption to exclude safety-related LSS and NRS components from the scope of the regulations covering seismic and dynamic qualification testing l requirements. The licensee has not addressed the impact of common-cause failures in

, multiple LSS and NRS components resulting from earthquake loadings. In addition, it is l

unclear how an LSS or NRS component can be expected to perform its safety-related function under seismic conditions without a qualification test or analysis.

In addition, with regards to dynamic effects resulting from pipe breaks, the licensee has ,

not discussed how spatial effects associated with failures of LSS and NRS components j l

as well as cascading effects resulting from failures of high-energy LSS and NRS l components are treated in evaluating the risk consequences.

3. 10 CFR Part 50. Aooendix J Exemotion: The STP licensee is requesting an exemption to exclude safety-related LSS and NRS components from the scope of Appendix J leak testing requirements for the primary reactor containment. Because the reactor containment provides the final fission-product barrier for defense-in-depth, the licensee's request does not satisfy one of the five principles of risk informed itegrated j l decisionmaking as described in RG 1.174.' Furthermore, the inservice inspection '

methods for ASME Code Class MC (metallic containment) and CC (concrete containment) components as required by 10 CFR 50.55a relies heavily on the Appendix {

J leak test to assure the pressure integrity of the containment. It is unclear how the '

leak-tightness of the containment can be assured without such a periodic test.

Comment #9:

t EMCB issues l

1. If materials are no longer specified to Section Ill, Class 1,2, or 3 requirements, how will ;

these components meet their design, material, testing and ISI requirements? What l tests will be performed to demonstrate that materials will satisiy their design, materials, j and ISI requirements?.

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2. How will PRA be performed for components that are replaced due to leaks, cracks, etc during plant operation?

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3. There are other risk-informed initiatives being pursued by the industry and the NRC i staff, such as risk-informed IS1, IST, Graded QA, and Tech Spec. How does the l

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9 request for multiple exemptions to change the classification of components relate to l these initiatives?

4. Are the criteria utilized to re-classify components consistent with the guidelines of RG 1.174 and the associated RGs for risk-informed ISI, IST, Graded QA, and Tech Spec.

5. Which systems or portions of systems are being considered for exemption? What would  ;

be the systems classification using the guidance provided in RG 1.26 for those systems l or portions of systems? l I

Comment #10:

1. Explain the difference between GOA performance monitoring and maintenance rule monitoring?

2. Would a revision to RG 1.160 regarding component level monitoring for the maintenance rule provide more appropriate clarification that the maintenance rule does not require component level monitoring?

3. How does the exemption effect the corrective action requirements of the maintenance rule (i.e. if a train or system or plant levelloss of function results from failure of an LSS or NRS component)?

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