ML20094J691

Vendor Equipment Technical Info Program, on Generic Ltr 83-28,Section 2.2.2

ML20094J691
Person / Time
Site:	River Bend
Issue date:	03/31/1984
From:	NUCLEAR UTILITY TASK ACTION COMMITTEE
To:
Shared Package
ML20094J685	List: ML20094J691 RBG-18-521, Forwards Final Response to Sections 1.1,2.1,2.2 & 4.5 of Generic Ltr 83-28, Required Actions Based on Generic Implications of Salem ATWS Events. Response to Sections 3.1 & 3.2 Will Be Provided Prior to Fuel Load
References
GL-83-28, INPO-84-010-01, INPO-84-10-1, NUDOCS 8408140423
Download: ML20094J691 (47)

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4 ABE9WrMTIQM ONLY Nuclear Utility Task Action Committee l

ON GENERIC LETTER 83-28, SECTION 2.2.2 Vendor Equipment Technical Information

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Vendor Equipment Technical Information Program Developed By Nuclear Utility Task Action Committee for Generic Letter 83-28, Section 2.2.2 INPO 84-010 (NUTAC) .

March 1984 Copyright 1984 by institute of Nuclear Power Operations. All rights reserved. Not for sale. Unauthonzed reproduction is a violation of applicable law. Reproduction of not more than ten copies by each recipient for its internal use only is permitted.

n poR INFORMATION W Publications produced by a nuclear utility task action committee (NUTAC) represent a consensus of the utilities participating in the NUTAC. These pub-lications are not intended to be interpreted as industry standards. Instead, the publications are offered as suggested guidance with the understanding that individual utilities are not obligated to use the suggested guidance.

This publication has been produced by the NUTAC on Generic Letter 83-28, Section 2.2.2., with the support of the Institute of Nuclear Power Operations (INPO). The officers of this NUTAC were Chairman Edward P. Griffing and Vice Chairman Walter E. Andrews.

Utilities that participated in this NUTAC include the following:

. Alabama Power Company Nebraska Public Power District American Electric Power Service Corporation New York Power Authority Arizona Public Service Company Niagara Mohawk Power Corporation Arkansas Power & Light Company Northeast Utilities Baltimore Gas and Electric Company Northern States Power Company Boston Edison Company Omaha Public Power District Carolina Power & Light Company Pacific Gas and Electric Company Cincinnati Gas & Electric Company Pennsylvania Power & Light Company The Cleveland Electric Illuminating Company Philadelphia Electric Company Commonwealth Edison Company Portland General Electric Company Consolidated Edison Company of New York, Inc. Public Service Company of Colorado Consumers Power Company Public Service Company of Indiana, Inc.

The Detroit Edison Company Public Service Company of New Hampshire Duke Power Company Public Service Electric and Gas Company Duquesne Light Company Rochester Gas and Electric Corporation Florida Power Corporation Sacramento Municipal Utility District Florida Power & Light Company South Carolina Electric & Gas Company GPU Nuclear Corporation Southern California Edison Company Georgia Power Company Tennessee Valley Authority Gulf States Utilities Company Texas Utilities Generating Company Houston Lighting & Power Company The Toledo Edison Company Illinois Power Company Union Electric Company Iowa Electric Light and Power Company Vermont Yankee Nuclear Power Corporation Kansas Gas and Electric Company Virginia Electric and Power Company Long Island Lighting Company Washington Public Power Supply System Louisiana Power & Light Company Wisconsin Electric Power Company Maine Yankee Atomic Power Company Wisconsin Public Service Corporation Mississippi Power & Light Company Yankee Atomic Electric Company NOTICE: This document was prepared by a nuclear utility task action committee (NUTAC) with the staff support of the Institute of Nuclear Power Operations (INP0).

Neither this NUTAC, INP0, members and participants of INP0, other persons contrib-uting to or assisting in the preparation of the document, nor any person acting on behalf of these parties (a) makes any warranty or representation, expressed or implied, with respect to the accuracy, completeness, or usefulness of the informa-tion contained in this document, or that the use of any information, apparatus, method or process disclosed in this document may not infringe on privately owned rights, or (b) assumes any liabilities with respect to the use of any information, apparatus, methad, or process disclosed in this document.

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EXECUTIVE

SUMMARY

This report was prepared by the Nuclear Utility Task Action Committee (NUTAC) on Generic Letter 83-28 " Required Actions Based on Generic Implications of Salem ATWS Events," Section 2.2.2. It describes the Vendor Equipment Tech-nical Information Program (VETIP) developed by the NUTAC in response to the concerns on vendor information and interface addressed in Section 2.2.2 of the generic letter. VETIP is a program that enhances information exchange and

( evaluation among utilities constructing or operating nuclear power plants and provides for more effective vendor interface.

The NUTAC was comprised of representatives of 56 utilities that are members of the Institute of Nuclear Power Operations (INP0). Staff support for the NUTAC was provided by INP0. This report unanimously presents the final conclusions of the NUTAC and is provided to assist individual utilities in developing specific programs to meet the intent of the generic letter.

Generic Letter 83-28 was developed following investigations by the NRC on the Salem events. As a result of these investigations, the NRC determined that better control and utilization of information regarding safety-related compo-nents might.have helped to prevent these events. The NUTAC identified a program to better ensure that plant personnel have timely access to such information.

The NUTAC efforts were guided by the recognition that individual utilities '

l have the greatest experience with and are most cognizant of the application of safety-related equipment. Vendor involvement with such equipment is generally greatest during construction and initial operation of the plant. Vendors are not familiar with the surveillance or maiatenance histories, nor with the application of the equipment or its environment. This type of information is most readily available at the plant level within individual utilities.

Based on this recognition, the NUTAC investigated the mechanisms currently available to facilitate information exchange among utilities. The NUTAC identified four activities that currently address information about safety-1

• 296L AMF.O?#ATION ONLY related eomponents. These are routine utility / vendor and utility / regulator interchange, and the SEE-IN and NPRDS programs managed by INPO.

It was the assessment of the NUTAC that these existing activities, if properly integrated and implemented, would provide a framework for an overall program to ensure effective communication of safety-related information among all utilities. Accordingly, the program developed to accomplish this goal (VETIP) uses the existing efforts as elements of a more comprehensive program.

The VETIP combines these existing programs, incorporating enhancements, with a coordinated program within each utility. A key element of the VETIP is-the development by each utility of an active internal program to contribute infor-mation to the NPRDS and SEE-IN Programs and to use the results of these programs.

The effectiveness of the VETIP will be determined by the level of utility par-ticipation in these programs. To implement the VETIP, each utility should

. assess the type of information currently being provided to NPRDS and SEE-IN and expand the scope of reporting if appropriate. Additionally, each utility should evaluate current administrative controls for reporting information and for disseminating the results of the NPRDS and SEE-IN Programs to the plant level. These administrative controls may require modification to ensure that effective coordinotion is established. Concurrent with these efforts, enhance-ments will be made to both NPROS and SEE-IN by INP0 within its present insti-tutional objectives.

The VETIP has been developed to ensure that nuclear utilities have prompt access to and effective handling of safety-related equipment technical infor-mation. In addition, VETIP is responsive to the intent of Generic Letter 83-28, Section 2.2.2. Further details are provided in the body of this report.

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' 6fA RMWhATICH (9M FOREWORD On February 22 and 25,1983, during start-ups of the Salem Unit 1 plant, both reactor trip breakers (Westinghouse model DB-50) failed to open on an auto-matic trip signal. As a consequence, the Nuclear Regulatory Commission (NRC) formed an investigating task force to determine the factual information perti-nent to the management and administrative controls that should have ensured proper operation of the trip breakers. The findings and conclusions of the task force are documented in NUREG-0977, "NRC Fact Finding Task Force Report on the ATWS Events at the Salem Nuclear Generating Station, Unit 1, on February 22 and 25,1983." A second task force determined the extent to which these investigative findings were generic in nature. The NRC subsequently issued NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant" and Generic letter 83-28, " Required Actions Based on Generic Implications of Salem ATWS Events."

On September 1,1983, a group of utility representatives met at the offices of the Institute of Nuclear Power Operations (INP0) to discuss the establishment of an ad hoc utility group to address issues relative to the NRC Generic Letter 83-28, Section 2.2.2. The representatives decided that such a group could provide direction that would be of generic benefit to the utilities and consequently formed the Nuclear Utility Task Action Committee (NUTAC) on Generic Letter 62-28, Section 2.2.2. The specific charter for the NUTAC (Appendix A) was adopted, and the target date for completion of activities was established as February 1984.

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FOR INFORMATION ONLY TABLE OF CONTENTS Section Page

1. INTRODUCTION ........................................................ 1

2. ACPONYMS-AND DEFINITIONS.............................................. 3 2.1 Acronyms ........................................................ 3 2.2 Definitions...................................................... 4

3. VENDOR EQUIPMENT TECHNICAL INFORMATION PROGRAM (VETIP) DESCRIPTION.... 7 3.1 Existing Programs................................................ 8 3.1.1 Nuclear Plant Reliability Data System (NPRDS). . .... .. . .. 9 3.1.2 Significant Event Evaluation and Information Network (SEE-IN)............................................... 11 3.1.3 Interaction with Vendors............................... 14 3.1.4 Regul atory Reporti ng Requi rements . . . . . . . . . . . . . . . . . . . . . . 16 3.2 Recommended Enhancements to Existing Programs................... 17

3. 2 .1. Enhancements to NPRDS.................................. 17 3.2.2 Enhancements to SEE-IN................................. 19 3.3 S u mm a ry E x a m p l e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19

4. IMPLEMENTATION OF VETIP.............................................. 21 4.1 Responsibilities for Implementation............................. 21 4.1.1 Utility Implementation Responsibilities................ 21 4.1.2 INP0 Implementation Responsibilities................... 24 4.2 Schedule for Implementation..................................... 25 4.2.1 Existing Programs...................................... 25 4.2.2 Enhancements to Existing Programs...................... 25 l

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Figure 1 - VETIP Block Diagram....................................... 26 Figure 2 - Operating Experience Review Process and Related .

A ct i v i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 7 APPENDIX A: SPECIFIC CHARTER FOR NUCLEAR UTILITY TASK ACTION COMMITTEE ON GENERIC LETTER 83-28, SECTION 2.2.2 APPENDIX B: LIST OF REFERENCES APPENDIX C: SEE-IN FUNCTIONS APPENDIX 0: GENERIC LETTER 83-28, SECTION 2.2.2 y

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i. INTRODUCTION The objective of Generic Letter 83-28, Section 2.2.2 (Appendix D), is to improve the safety and reliability of nuclear power generating stations by ensuring that the utilities are provided with significant and timely tech-nical information concerning reliability of safety-related components. In a typical nuclear station, hundreds of vendors supply the thousands of components that perform safety-related functions. The variations in vintage and design of plants ensure that although common applications of specific components may exist, there are an equal or greater number of unique applications. To attain the objective in a cost-effective and efficient manner, this NUTAC has developed the program outlined in this document. This positive program has been found to be the most realistic approach to attain the objective.

The Vendor Equipment Technical Information Program (VETIP) described in this document establishes a more formal interaction among the major organ-izations involved with commercial nuclear power generation. The goal of the interaction is to improve the quality and availability of equipment technical information for use by the utilities. The major components of the' VETIP are an information transfer system and a centralized evaluation of industry experiences.

This document provides the unanimous NUTAC position on the guidelines for an effective technical information program. The determination of each individual utility to support and utilize these guidelines is the key to .

the effectiveness of this program for the industry as a whole. . The pro-gram does not require the use of nor prescribe standard administrative ,

procedures, but it allows the use of plant-specific procedures compatible with the utility's internal organization and needs. However, the recom-mendations in this document provide the basis for a uniform industry response to NRC questions and requirements relative to a technical infor-mation program. This program will be beneficial to the utilities and, at the same time, it will be responsive to Section 2.2.2 of the NRC Generic Letter 83-28.

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F FOR INFORMATI0 lim 2.* ACRONYMS AND DEFINITIONS 2.1 Acronyms A/E Architect-Engineer AE00 Office of the Analysis and Evaluation of

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Operational Data ATWS Anticipated Transient Without Scram CFR Code of Federal Regulations EPRI Electric Power Research Institute ETI Equipment Technical Information IEB, IEN Inspection and Enforcement Bulletins and Notices, issued by the NRC IEEE Institute of Electrical and Electronics Engineering INPO Institute of Nuclear Power Operations LER Licensee Event Report, issued by a utility MOR Monthly Operating Report NPRDS Nuclear Plant Reliability Data System NRC Nuclear Regulatory Commission NSAC Nuclear Safety Analysis Center NSSS Nuclear Steam Supply System NUTAC Nuclear Utility Task Action Committee-0&MR Operations and' Maintenance Reminder PRA Probabilistic Risk Assessment QA Quality Assurance SEE-IN

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Significant Event Evaluation and Information .

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SER Significant Event Report -

SOER Significant Operating Experience Report VETIP Vendor Equipment Technical Information Program l

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i-o (OR INFCKS%TIOR QMZ 2.2 Definitions Component - A component is a mechanical or electrical assembly (including instruments) of interconnected parts that constitutes an identifiable device or piece of equipment. Examples of electrical components include a drawout circuit breaker, a circuit card, instru-ments, or other subassemblies of a larger device that meet this definition. Examples of mechanical components include valves,

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piping, pumps and pressure vessels, and associated prime movers

and/or operators.

Equipment Technical Information (ETI) - For the purposes of this

. report, this term includes, as a minimum, the following documenta-i tion:

o vendor-supplied engineering and technical information (drawings, manuals, etc.) and changes thereto 9

o equipment qualification data (provided by the equipment vendor or qualification lab) o i'ndustry-developed information, including utility and NRC-originated information (NPRDS, SER, IEB, IEN, etc.)

NUCLEAR NETWORK"* - An information service provided through INP0.

(NUCLEAR NETWORK replaced NUCLEAR NOTEPAD.)

NUREG - These are guidance documents that are issued by the NRC.

Safety-Related - Safety-related structures, systems, and components are those relied upon to remain functional during and following design basis events to ensure (1) the integrity of the reactor cool-ant pressure boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, and (3) the capability

• Trademark application by INP0 for NUCLEAR NETWORK is pending.

gagamTK210FE to prevent or mitigate the consequences of accidents that could result in potential off-site exposures comparable ~to the guidelines of 10 CFR Part 100.

. Vendor - For the purposes of this report, this term is used to iden-tify the manufacturer of the component concerned and/or those who provide the related equipment technical information.

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3. VENDOR EQUIPMENT TECHNICAL INFORMATION PROGRAM (VETIP) DESCRIPTION The VETIP includes interactions among the major organizations involved with commercial nuclear power generation. As illustrated in Figure 1, a utility' exchanges safety-related equipment information with vendors, NRC,

. -INP0, and other utilities via reports, bulletins, notices, newsletters, and meetings. The purpose of these information exchanges is to share

- equipment technical information to improve the safety and reliability of nuclear power generating stations. The NUTAC concluded that the lack of information is not a problem, but that the various information systems available are not integrated properly. The purpose of VETIP is to ensure that current information and data will be available to those personnel responsible for developing and maintaining plant instructions and pro-cedures. These information systems and programs currently exist and are capable of identifying to the industry precursors that could lead to a Salem-type event. VETIP is an industry-controlled and mainly hardware-oriented program that does not rely on vendor action, other than the NSSS supplier, to provide information to utilities. Instead, VETIP provides information developed by industry experience through SERs and SOERs to the vendor for comment before it is circulated to the utilities concerned.

The majority of information provided by vendors is commercial in nature.

This usually is provided voluntarily by the vendor, but does little to improve the safety or reliability of existing equipment.

A vendor-oriented program to provide information that would improve the safety and reliability of_ existing equipment relies on the vendor having an internal program to develop the information. Such programs typically are not in existence. Following design and qualification testing, vendors normally do not continue extensive testing or engineering programs in anticipation of equipment problems. Subsequent failures discovered during

-operations require several steps to complete the information feedback l oop_. For example,'when a problem occurs and a' local vendor represen-tative provides a solution, he would have to provide that information to the vendor headquarters. Then, the headquarters would need a tracking program to idc:tify ? trend and subsequently a program to provide the

. Information to the industry. In addition, the vendor often is not in the O _ . _ . . _ _ - __ __ _ _ _-

{~ FOR INFO %%TiiOM M best position to analyze the failure. The vendor is not always aware of the component's application and environment nor its maintenance and sur-veillance history.

The VETIP recognizes that the utility user is in a unique position. The

, utilityLuser alone has immediate access to the maintenance and surveil-lance history of the equipment. The utility, not the manufacturer, knows the component's actual application and environment. The utility is the primary source of information on the failure, and the utility has the greatest need for the solution. As such, the utility is the central organizer in any approach to the solution, whether or not the manufacturer gets involved. The utility is in the position to know of the failure analysis and its solution at the earliest possible time. The utility can then disseminate the information to other utilities, with an indication of its significance and urgency.

By sharing the operating history, problems, and solutions within the nuclear industry, independent of any normal vendor contacts, the other users will be informed in a much more timely and uniform way. In this way, the distribution of information is controlled entirely by the nuclear utility industry. The programs that comprise the VETIP currently are in existence. The recommended enhancements contained within this report are suggested ways to improve the current use and application of these exist-ing programs.

3.1 Existing Programs The existing systems and programs included in the VETIP are the Nuclear Plant Reliability Data System (NPRDS) and the Significant Event Evaluation and Information Network (SEE-IN), both managed by INP0. Also, the VETIP includes existing programs that the utilities now conduct with vendors and other sources of ETI, particularly the NSSS vendor interaction programs and the NRC reporting programs that disseminate significant failure information. Utility-vendor inter-action is further enhanced by the INP0 supplier participant practices.

Through participation in this program, NSSS vendors and A/E firms are working toward greater participation in the NPRDS and SEE-IN Programs.

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f MINKPWATEOMCdM3, 3.1.1 Nuclear Plant Reliability Data System (NPRDS)

NPRDS js an industrywide system managed by INP0 for monitoring the performance of selected estems and components at nuclear power plants. INP0 member utilities have agreed to partici-

pate in the program. United States plants in commercial ,,

operation (except for six atypical, early vintage units) supply basic engineering information and subsequent failure data on the selected systems and components (typically six to seven thousand components from some 30 systems per unit). The value of NPRDS lies in the ready availability of this data ,

base to operation and engineering groups for a broad range of applications. The criteria used to determine the scope of NPRDS reports are as follows:

o systems and components that provide functions necessary for accident mitigation o systems and components for which loss of function can initiate a significant plant transient Uniform scoping and reporting criteria are set forth in the Nuclear Plant Reliability Data System (NPRDS) Reportable System and Component Scope Manual (INP0 83-020) and in the Reporting Procedures Manual for the Nuclear Plant Reliability Data System (INP0 84-011).

To support the benefits that can be obtained from NPRDS usage, i

utilities submit three kinds of information to the NPRDS data l base: engineering / test information, failure reports, and operating history. The engineering / test record on a component contains information necessary to identify the component and its application, such as manufacturer, model number, operating environment, size, horsepower, and test frequencies. The information is submitted when the component is placed in

, service and is stored in the data base. If that component fails to perform as intended, a report is submitted containing a description of the failure mode and cause, the failure's-f effect on plant operations, corrective actions taken, and other

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.w information necessary to assess the failure. On a quarterly basis, utilities submit information on the number of hours the plant'is in different modes of operation. This information is used in conjunction with the engineering and failure reports-to generate failure statistics for systems and components.

The data is retrievable from a computer, and the engineering and failure information can be combined in various ways. A search of the failure records can identify problems experi-enced with components in other plants and the corrective

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actions taken. There are several hundred searches of the data base in a typical month. Following are some example uses of the data base:

Utility and Plant Staffs o accessing comprehensive equipment history files to support maintenance planning and repair o avoidance of forced or prolonged outages by identifying other plants with similar or identical equipment that may have spares for a possible loan o determination of spare parts stocking, based on industry mean time between failures o comparison of component failure rates at a given plant with the industry average failure rates Design Groups o identification of common failure modes and causes o selection of vendors based on component application and performance o identification of component wearout and aging patterns o studies of component performance as a function of operating characteristics, such as test frequency and operating g environment o input to plant availability improvement programs l p,

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o identification of significant failure modes affecting safety or availability o trending of component failure rates o development of failure probability estimates for use in fault-tree analyses (reliability or PRA studies)

NPRDS data is available to users through various periodic reports and through on-line access of sthe data from a computer terminal.

3.1.2 Significant Event Evaluation and Information Network (SEE-IN)

Since the early days of nuclear power plant operations, utili-ties and manufacturers have attempted to share what has been learned from plant operating experience. As nuclear tech-nology becomes more complex and more demanding, the need for sharing operating experience continues to grow and becomes more important. The safety benefits of avoiding problems already encountered and resolved more than justifies the costs and extra effort required for utilities to keep each other informed. The Nuclear Safety Analysis Center (NSAC), with the support of its utility advisory group, began developing a program-to share-information learned from analyzing nuclear plant experiences. Shortly after its formation in late 1979, the Institute of Nuclear Power Operations (INP0) joined NSAC in the development and implementation of the program. The .

I program has been named "Significant Event Evaluation and l Information Network" (SEE-IN). In 1981, the management of the -

l SEE-IN Program became the sole responsibility of INP0.

l Objective l The objective of SEE-IN is to ensure that the cumulative l learning process from operating and maintenance experience is

! effective and that the lessons learned are reported and cor-rective action taken in a timely manner to improve plant safety, reliability, and availability. .This objective is met L-

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-s by screening available nuclear plant event information system-atically, identifying and evaluating the important or signifi-cant events, and communicating the results to the utilities and appropriate designers and manufacturers.

Scope The functional approach to SEE-IN is an eight-step process outlined in Appendix C. While INP0 has the program management function, no single organization is responsible for performing all of these functions; rather, the responsibility is spread among key participants in the network. The principle organi-zations involved in the initial screening of plant event data are the utilities and INP0. Each nuclear utility has an in-house program to sc,een events that occur in its nuclear plant (s).. INPO has a broader charter to screen all nuclear j; plant events. The sources of input to the screening process include NPRDS, NUCLEAR NETWORK, h;C-mandated reports, IEBs, IENs,.etc. The provision to control the data normally is governed by agreements between INPO'and the supplying organi-zation (e.g., utilities, NRC, NSSS vendors, international partici;, nts, etc.). When a significant event 'or. trend has been identified from the screening process, a Significant

! Event Report (SER) is prepared by INP0 and transmitted to the utilities and other participants on NUCLEAR NETWORK. This event then undergoes an action analysis by INP0. The purpose i of the action analysis is to investigate the event or trend in i more detail and to develop and evaluate practical remedies.

L For events requiring utility action, the results of the action analysis are communicated to the utilities, normally in the L form of a Significant Operating Experience Report (SOER). In these instances, recommendations are made to resolve the under-lying problems. The implementation of applicable recommended i remedial actions is the responsibility of the individual util-ity. Implementation may include changes in plant procedures, equipment design, and/or operator training programs. The two final steps in the SEE-IN process are (1) feedback and INPO i

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assessment during plant evaluation of actions taken by the

. utilities as a result of information provided through SEE-IN and (2) periodic assessment of the process effectiveness by INPO.

For events which, through the screening process, are deter-

, mined not significant but have valuable operations and mainte-nance information, an Operations & Maintenance Reminder (0&MR) is prepared and processed in the same way as SERs.

The SEE-IN Program provides copies of draft SERs, 0&MRs, and SOERs to the affected vendors for review. Vendor comments are considered in preparation of final SEE-IN reports. Once finalized, the reports are sent to the utilities.

The SEE-IN Program includes a cross-reference capability to identify SERs, 0&MRs, S0ERs, LERs, etc., which report compo-nent problems that could cause a significant event. This cross-reference facilitates utility review of the component's prior history before using that component in a safety-related application.

Program Operation Plant operating experience data is reviewed from several perspectives including design, component and system perfor-mance, plant procedures, human factors, personnel training, maintenance and testing practices, and management systems to identify significant events and trends. '

Formal Review Sources A formal review is conducted on NRC information notices, bulletins, AE00 reports, event-related generic letters, etc.

A formal review also is conducted on industry-prepared infor-mation (including those required by NRC) such as LERs, monthly operating reports, NRC event-related reports, NSSS technical bulletins, NPRDS data, NUCLEAR NETWORK operating experience k_

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' tion deficiency reports, safety defect reports, and trends identified as significant in the INP0 NPRDS and LER data bases. The formal review includes a dual, independent screen-ing process. The review status is documented and tracked by computer.

Other sources of operating experience information are used by the SEE-IN Program on an ad hoc basis as reference or supple-mental material but do not receive a formal review. The sources include such items as NRC NUREG documents, EPRI and NSAC reports, and other industry reports or data concerned with plant operating experience. The INP0 process for screen-ing is shown in Figure 2.

Utility Contact (SEE-IN)

In addition to the formal and reference information sources, another vital information source is direct contact with power plant technical personnel on an ad hoc basis. Each utility designates a SEE-IN contact to respond to questions from INP0 on plant events. The majority of such communications was handled over the telephone or via NUCLEAR NETWORK. Files are maintained by INP0 on nuclear utilities and contain names and telephone ntmbers of designated contacts, telecopier numbers, status of nuclear units (i.e., operating, under construction or planned), and NSSS vendor (s).

3.1.3 Interaction With Vendors In the interest of operating the plant safely and efficiently, the utility-vendor contact is essential. To accomplish this goal, utilities already interact with various vendors.

The contractual obligations for furnishing equipment and software (manuals, drawings, etc.) are fulfilled upon accep-tance at the plant site. Interaction between utilities and vendors, due to deficiencies, may be brought about by the

I Foi &NFOih% TION G4LY reporting requirements of 10 CFR 21 and 10 CFR 50.55(e). The continuing contract with vendors for warranty obligations or maintenance work are two examples of active interaction after an initial purchase. In addition, much of the interaction with the vendors during plant life is initiated in response to significant failures, to failure trends experienced at the

, plant, to spare parts procurement, or to subsequent purchase orders of new equipment.

The interaction with the NSSS vendor, who typically supplies a large portion of the safety-related plant equipment, generally is more active than with the other vendors. There are exist-ing channels through which the NSSS suppliers disseminate information of interest to their client utilities. These include the following:

o In regular meetings, NSSS representatives outline recent developments and maintenance / design recommendations. Any special concerns of the utility can be addressed in follow-up correspondence with the NSSS supplier's service depart-ment.

o Bulletins or advisories from the NSSS supplier's service department alert client utilities to special problems experienced by similar plants. Typically included in this correspondence are a description of the problem and the corrective actions taken to resolve it. Recommendations for preventive actions or for particular cautions to be considered by the utility usually are included.

o Owners groups provide an additional forum for the exchange of information that may be of generic interest to member utilities. For example, problems in the design or oper-ation of a system or component may be shared with the group and potential resolutions identified. The owners groups' efforts often are directed at seeking improvements or anti-cipating problems rather than being only reactive in nature.

% E Improvements in availability or testing and maintenance procedures are examples of positive'results that have.come about through owners groups activities. The NSSS supplier makes his broadly-based knowledge available to the group for.the specialized evaluations that may be required.

3.1.4 Regulatory Reporting Requirements Other existing sources of information are the documents that result from the NRC's reporting requirements. These documents include 10 CFR 21 reports,10 CFR 50.55(e) reports, Licensee Event Reports, and NRC Inspection & Enforcement (IE) Bulletins and Information Notices. 10 CFR 21 specifies reporting-require-ments relating to component or system deficiencies that may create a substantial safety hazard. This reporting provides the nuclear utility industry notification of significant noncompliances and defects identified by other utilities, architect-engineers, constructors, vendors, and manufacturers associated with nuclear facilities.

10 CFR 50.55(e) requires that the holder of a construction permit notify the NRC of each deficiency found in design and construction, which, if uncorrected, could affect the safe operation of the nuclear power plant adversely.

10 CFR 50.73 requires the holder of an operating license for a nuclear power plant to submit a Licensee Event Report (LER)

! for events described in 50.73(a)(2). These LERs are incor-porated into the INPO LER data base, which provides informa-l tion to identify and isolate precursor events and identify emerging trends or patterns of potential safety significance.

l The NRC Office of Inspection and Enforcement (IE) issues various documents, including bulletins and information notices, to inform licensees and construction permit holders lz of significant concerns that may result from the NRC evalua-tion of reports, as required by 10 CFR 21.21, 50.55(e), and '

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50.73. These documents provide the nuclear utilities with information on events and concerns that are considered sig-nificant by the NRC.

i 3.2 Recommended Enhancements to Existing Programs The following are recommended enhancements to the existing programs.

INP0 and the NPRDS User's Group should investigate the feasibility of these recommendations. If found feasible, an implementation program should be developed.

3.2.1 Enhancements to NPRDS o The present definition of component in NPRDS (extracted from IEEE 603-1980) is more applicable to electrical components.

The definition should be improved to describe mechanical components better.

o The present failure reporting guidance needs improvement in the following areas:

-- Guidance is needed to provide better information for analyzing the role of piece parts as a factor in causing component failures.

-- The guidance should be revised to indicate that utilities should supply information when inadequate vendor informa-tion is identified as a causal or contributing factor in a failure. The guidance should provide users of the data base the ability to retrieve readily those failures involving inadequate vendor information (example, key

• word sorting, coding).

-- Present failure reports are often sketchy in providing details of the failure analysis conducted by utilities.

The guidance should emphasize the importance of providing more complete results of failure analysis when one is conducted. Although detailed failure analyses are not j

h

MNM always conducted for every fa' lure, when they are con-ducted they should be provided in NPRDS failure reports.

In this way, the SEE-IN Program and other utilities can derive more benefit from the work of each utility.

o Utilities should develop internal methods to ensure that their NPRDS reports are clear and complete and that the program guidance is followed appropriately.

o For some failures it may not be possible for utilities to provide a complete failure description within the time frames for reporting to NPRDS. Utilities should still submit preliminary failure reports within the established time frame. Utilities should revise these reports when the necessary information is available. However, the present system does not provide methods for utilities to indicate that reports will be revised later. NPRDS should be modi-fied to permit each utility to readily identify which of their reports still requires follow-up information. Utili-ties should report a failure event promptly and include an initial analysis. Detailed and complete information should be provided in a timely manner once final analysis has been completed.

o The present scope of NPRDS reporting may not meet all the needs of individual utilities for monitoring the relia-bility of their own safety-related components. Each utility that decides that additional systems and components should be added to their basic scope of NPRDS systems and components should request that INP0 accept these systems. INP0 will consider these requests, identify the additional resource requirements needed to handle these requests, and notify utilities when it is able to accept additional information.

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ea m m m ass 3.2.2 Enhancements to SEE-IN o Reports should be generated for potential failures caused by faulty or missing vendor-supp led information or other ETI. The VETIP recognizes that the utility will uncover

. errors in ETI (e.g., during review of the information,

writing of instructions, testing, etc.) before anyone

( else. It is recommended that ETI faults be reported over l'

NUCLEAR NETWORK for review by INP0 under the SEE-IN Program.

o The SEE-IN Program should be broadened by INPO to improve i the ability to trend NPRDS data. Present methods of trend-

. ing are largely qualitative and subjective in nature. .They l' depend largely on the ability of analysts to recognize the

!- need to look for degrading or unacceptable system and com-ponent reliability. INPO should develop methods to use NPRDS in a' more quantitative fashion to detect trend prob-j lems. This enhancement is presently under development by

t. INPO.

1 3.3 Summary Examole, One problem that led to the Salem event was that the information con-1 tained in the NSSS vendor technical bulletin (issued in 1974) was not processed appropriately and therefore not incorporated into plant procedures. If the systems that comprise the.VETIP were' functional in the early.1970s, this oversight probably would not have occurred

+

or would have been rectified. Westinghouse had prepared the techni-cal bulletin tased on a precursor event that occurred at another nuclear unit. This. type of precursor event would have required that -

an LER be written and submitted to the NRC.. INP0 also would have p reviewed the Westinghouse technical bulletin and the LER. The cur-

• - rent criteria for significance screening used by INPO personnel identify this type event as a significant single failure. It is highly likely that'an SER would have been generated by INP0 and disseminated to utilities via NUCLEAR NETWORK. Utilities would have reviewed the SER through 'their operating experience report review programs.

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PM SUTR.WE"E M In addition, utilities would have had an ongoing program with their NSSS vendors to obtain ETI. Utilities would have had systems in place to track and process this information. Therefore, there are two pathways that would have ensured this type of information was received and evaluated by the utility:

o NPRDS/SEE-IN (SERs, SOERs) 3 NSSS vendor technical bulletins The utility's VETIP procedures would have assessed this information and effected positive action to correct the failed component.

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4. IMPLEMENTATION OF VETIP 4.1 Responsibilities for Implementation 4.1.1 Utility Implementation Responsibilities 4.1.1.1 Existing Programs NSSS Vendor Contact Each utility should have a program in place with its NSSS supplier to obtain technical information. This program consists of a technical bulletin system and necessary direct contact with the NSSS supplier.

NPRDS/SEE-IN

, Each utility should indicate or reaffirm its active participation in the NPRDS and SEE-IN Programs. The utility should supply the nectssary basic information and should report failures and problems on a timely basis. Adequate internal controls should be in place to ensure that this activity is timely, consistent, and controlled and should include incorporation of future revisions to these programs.

Other Vendors Each utility should continue to seek assistance and ETI from other safety-related equipment vendors when the utility's evaluation of an equipment or ETI problem concludes that such direct interaction is .

necessary or would be beneficial. These problems and those of lesser significance will continue to be reported by means of the NPRDS and/or the SEE-IN Programs.

Internal Handling of Equipment Technical Information The utility should process incoming ETI so the objec-tives noted below are achieved.

F FOR INFORMATION0NW t

o Administrative procedures should provide control of incoming ETI whether it arrives directly from the vendor or from other industry or regulatory sources (i.e., NUCLEAR NETWORK, NPRDS, SEE-IN, NRC bulletins, etc.), so it receives the appropriate engineering / technical review, evaluation, and distribution for the following:

prompt warnings to key personnel

-- timely incorporation into maintenance or operating procedures, equipment data /

purchasing records, and training programs future procedure review and revision cycles notification on NUCLEAR NETWORK of significant ETI The incorporation of such safety-related infor-mation (or changes) remains within the scope of the utility's review and ' approval requirements.

o The administrative program should require that maintenance or operating procedures cite appro-priate ETI in the reference section of the proce-dure.

o Within the performance section of the procedure, appropriate ETI should be incorporated and approved in the engineering, technical, and quality review of the safety-related procedure.

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FOR INF0iiMAT.10EM o Internal Handling of Vendor Services The vendor, contractor, or technical represen-tative who will perform safety-related services should be a QA approved / qualified supplier of such nuclear safety-related services. Furthermore, the services should be specified in the procurement .

documentation so that a combination of procedural and QA/QC controls are established.

A vendor service may be performed using utility procedures. If so, the procurement documentation should specify that the service is performed using utility procedures that have been approved after a technical and quality review cycle typical for other utility service, maintenance, repair, or operating procedures. As an alternative, the service may be performed using vendor or contractor procedures. In this case, the documentation should specify that the service is performed using vendor, contractor, or technical representative procedures that have been reviewed and approved in accordance with the utility procurement program, QA program, and administrative review program. This is to ensure that their documents are processed and approved in a manner equivalent to the utility procedures concerning similar activities.

In addition to specifying the procedures that will be used, the QA/QC program to be used should also be specified. The utility QA/QC program may be used. In this case, the procurement documentation should specify that the activity will be performed under the cognizance of the utility QA/QC program. Alternately, the vendor or contractor QA/QC program may be used. In this case, the f.

FOR INNM documentation should specify that the activity will be performed under the cognizance of the vendor, contractor, or technical representative QA/QC program that has been reviewed separately and approved in accordance with the utility QA program. In addition, during the performance of the service, the utility QA program will monitor the effectiveness of their performance and compliance with its approved program by suitable surveillance, inspection, and audit.

4.1.1.2 Enhanced Programs o NPRDS Each utility should incorporate the enhancements to the NPRDS recommended in Section 3.2. This could involve revisions to existing administrative programs or procedures. It also could require revised training or other actions needed to ensure a meaningful and effective implementation of the NPRDS program enhancements.

o SEE-IN Each utility should incorporate the enhancements to the SEE-IN program recommended in Section 3.2. As in the NPRDS program, this could involve revisions to existing administrative programs or procedures or to training or other activities so the data reported to the SEE-IN Program is com-plete and detailed enough to support the system enhancements being undertaken by INP0.

4.1.2 INP0 Implementation Responsibilities o Existing Programs The NUTAC determined that present NPRDS/SEE-IN Programs, properly used, currently provide an adequate framework for the effective exchange of information.

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. EOR 18EomATMMM o Enhanced Programs INPO should implement the enhancements of the NPRDS and SEE-IN Programs (noted in Section 3.2) to augment this VETIP.

4.2 Schedule for Implementation 4.2.1 Existing Programs Utilities that find that their existing internal program and procedures do not support those outlined in Sections 3.1 and 4.1.1.A above should make the necessary timely revisions as part of the established review and updating cycle for such documentation. A specific schedule should be established by the individual utility with a target date for full implementa-tion by January 1, 1985.

4.2.2 Enhancements to Existing Programs 4.2.2.1 INP0 should work with the NPRDS user's group with the goal of establishing schedules by July 1, 1984, for implementation of the enhancements of the NPRDS program.

4.2.2.2 Utilities should incorporate the enhancements to the NPRDS and SEE-IN programs, recommended in Section 3.2 and 4.1.1.B above into their internal program and procedures on a timely basis.

4.2.2.3 Schedules should be established that are consistent *-

with an overall goal to implement the recommended enhancements to both programs by January 1,1986.

. gQR N ORMA M N NSSS Vendor Spare Parts O Problems $

/ Tech. Reps.

Direct Contact Tech. Bulletins Owners Groups IEN/IEB/EER

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NPRDS Generic Letter Other Utilities Spare Paris Problems t

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I lEB/IEN/LER 10CFR 21 10 CFR 50.550 Generic Letters Board Notification 4

VETIP Block Diagram Figure 1

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e Misc. Event Info From Other Sources, e.g. NPRDS v

INPO Receives LERs, MORs, Reviewer Evaluates Event For And Other Significance EventInfo if involved.

Vendor Contacted u For Comment If Appropriate, Reviewer Prepares And issues An Event O&MR "

• ls Not -

Reviewer SER is n .

Significant Drafts SER

Revised As i

Necessary U

Eventis And Sent

, Disctmed At ,_

To Member Penodic Meeting Utilities Via The Event

'" NETWORK May Be -

-g Evt nt is Significant Sign icant -.---* No (Sign cant by

-* Further Action; ers)

Document Reasons Event is Studied Event Second Further By INPO,

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Review. OR ReviewerAssigns A Significance - -* Affected Utility. ->

f On -* Develop An --

Significant Updates Further SOER Of "NS- And Others As Action Necessary Operating Experience Review Process . ["ps^"

And Related Activities Evaluation Figure 2 k

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folt UWOINATION ONLY,

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APPENDIX A SPECIFIC CHARTER FOR NUCLEAR UTILITY TASK ACTION COMMITTEE ON GENERIC LETTER 83-28, SECT 10'N 2.2.2 This Nuclear Utility Task Action Committee (NUTAC) has been established by a group of utility representatives who have recognized a need for nuclear indus-try guidance on Generic Letter 83-28, Section 2.2.2. The estab1'shment of this NUTAC has been in accordance with the general charter governing the organization and operation of a NUTAC, as approved by the Institute of Nuclear Power Operations (INPO) Board of Directors. This NUTAC is committed to com-pliance with this specific charter, its bylaws, and the general charter. This charter has been reviewed and approved by the chairman of the Analysis and Engineering Division Industry Review Group and the president of INPO, and the president of INP0 authorizes staff support for this NUTAC.

This committee has adopted the following objective to ensure fulfillment of the goal of achieving industry consensus and guidance on Generic Letter 83-28, Section 2.2.2:

o development of guidance for use by utilities in response to Generic Letter 83-28, Section 2.2.2 To ensure that this objective results in products that are of generic benefit to the utilities, voting membership on this committee is limited to permanent employees of U.S. nuclear utilities. The chairman and vice chairman of this .

committee will be elected by the NUTAC from a list of candidates approved by the chairman of the sponsoring IRG. To further ensure that this NUTAC pro-vides products that are of generic benefit to utilities, the NUTAC chairman will maintain close liaison with the sponso' ring INP0 Industry Review Group.

A-1 V _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _

(

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) APPENDIX A SPECIFIC CHARTER FOR NUCLEAR UTILITY TASK ACTION COMMITTEE ON GENERIC LETTER 33-28, SECTION 2.2.2 a

)

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FOR DEORMAILOM CtGZ Additionally, this NUTAC should establish liaison with other recognized indus-try groups, such as AIF, ANS, EEI, EPRI, and NSSS owners groups and will maintain communication on this industry initiative with the NRC, as appropri-ate.

l Approved 7 $ h I 3 Chairman, NUTAC hDate Chairman, IRG Date 4 Af0L Vice Chairman, NUTAC PNrftG & m Date President, INP0 1/z'/s a Date l

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APPENDIX B l 1

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LIST OF REFERENCES l

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QW List of References

1. " Required Actions Based on Generic Implications of Salem ATWS Events" (Generic Letter 83-2P). Washington, D.C.: U.S. Nuclear Regulatory Commission, July 8, 1983.

2. NRC Fact-finding Task Force Report on the ATWS Events at Salem Nuclear Generating Station Unit 1 on February 22 and 25,1983 (NUREG-0977).

Washington, D.C.: U.S. Nuclear Regulatory Commission, March 1983.

3. Generic linplication of ATWS Events at the Salem Nuclear Power Plant (NUREG-1000). Washington, D.C.: U.S. Nuclear Regulatory Commission, April 1983.

4. Sionificant Event Evaluation and Information Network (SEE-IN)

(INPO 83-001). Atlanta, Ga.: Institute of Nuclear Power Operations, February 1983.

5. Nuclear Plant Reliability Data System (NPRDS) Reportable System and Compo-nent Scope Manual (INP0 83-020). Atlanta, Ga.: Institute of Nuclear Power Operations, 1983.

6. Reporting Procedures Manual for Nuclear Plant Reliability Data Systems (INP0 84-011). Atlanta, Ga.: Institute of Nuclear Power Operations, 1984.

7. 10 CFR 21, Code of Federal Regulations: Title 10 Part 21, " Reporting of Defects and Noncompliance." Washington, D.C.: U.S. Government Printing Office.

8. 10 CFR 50, Code of Federal Regulations: Title 10 Part 50, " Domestic Licensing of Production and Utilization Facilities." Washington, D.C.:

U.S. Government Printing Office.

9. Standard Criteria for Safety Systems for Nuclear Generating Stations (IEEE 603-80). New York, N.Y.: Institute of Electrical and Electronics Engineers, 1980.

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3-FORINFQWWQl1Sl$4 APPENDIX C SEE-IN FUNCTIONS e

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% M NOIf.M SEE-IN Functions

1. Provide basic report of plant event (utilities).

2. Screen events for significance and transmit Significant Event Reports (SERs) via NUCLEAR NETWORK (utilities and INP0 with vendor input solicited whcn specific product is identified).

3. Provide backup data on contributing factors and probable causes and con-sequences (utilities and vendors).

4. Perform action analysis on significant events to evaluate possible options for short-term remedies and feasible long-term solutions that might be implemented (utilities, INP0, and vendors).

5. Disseminate information, along with an alert of potential implication, to the utilities (INPO).

6. Evaluate the information and implement remedies as appropriate (utili-ties).

7. Provide feedback on implementation actions (utilities and INPO).

8. Evaluate periodically the effectiveness of the process, including steps 1-7 above (INP0).

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E6M APPENDIX D GENERIC LETTER 83-28 SECTION 2.2.2 (Generic Letter 83-28, Section 2.2.2 is enclosed verbatim) t e

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r U* UNITED STATES

• /[. 4, NUCLEAR REGULATORY COMMISSION

y. g WASHINGTON, D. C. 20555 k July 8, 1983

% $j TO ALL LICENSEES OF OPERATING REACTORS, APPLICANTS FOR OPERATING LICENSE, AND HOLDERS OF CONSTRUCTION PERMITS Gentlemen:

SUBJECT:

REQUIRED ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EVENTS (Generic Letter 83-28)

The Commission has recently reviewed intermediate-tenn actions to be taken by licensees and applicants as a result of the Salem anticipated transient without scram (ATWS) evehts. These actions have been developed by the staff based on information contained in NUREG-1000, " Generic Implications of A1WS Events at the Salem Nuclear Power Plant." These actions address issues related to reactor

trip system reliability and general management capability.

l The actions covered by this letter fall into the following four areas:

1. . Post-Trip Review - This action addresses the program, procedures and data collection capability to assure that the causes for unscheduled reactor shutdowns, as well as the response of safety-related equipment, are fully understood prior to plant restart.

2. Equipment Classification and Vendor Interface - This action addresses the programs for assuring that all components necessary for accomplishing required safety-related functions are properly identified in doctanents, procedures, and infonnation handling systems that are used to control

safety-related plant activities. In addition, this action addresses the l establishment and maintenance of a program to ensure that vendor information i for safety-related components is complete.

-3. Post-Maintenance Testing - This action addresses post-maintenance operability I testing of safety-related components.

l l 4. . Reactor Trip System Reliablity Improvements - This action is aimed at -

l assuring that vendor-recommended reactor trip breaker modifications and l associated reactor protection system changes are completed in PWRs, that

a' comprehensive program of preventive maintenance and surveillance testing ,

is implemented for the reactor trip breakers in PWRs, that the shunt trip attachment activates automatically in all PWRs that use circuit breakers in their reactor trip system, and to ensure that on-line functional testing i of the reactor trip system is performed on all LWRr..

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The enclosure to this letter breaks down these actions into several components.

You will . find that all actions, except four (Action 1.2, 4.1, 4.3, and 4.5) ,

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require sof tware (procedures, training, etc.) changes and/or modifications ,

and do not affect equipment changes or require reactor shutdown to complete.

Action 1.2 may result in some changes to the sequence of events recorder or existing plant computers, but will not result.in a plant shutdown to implement' .

Actions 4.1,'4.3 and 4.5.2, if applicable, would require the plant to be shutdown in order to implement.

The reactor trip system is fundamental to reactor safety for all nuclear power plant designs. All transient and accident analyses are predicated on its successful operation to assure acceptable consequences. Therefore, the actions listed below, which relate directly to the reactor trip system, are of the

- highest priority and should be integrated into existing plant schedules first.

1.1 Post-Trip Review (Program Description and Procedure) 2.1 Equipment Classification and Vendor Interface (Reactor Trip System Components) 3.1 Post-Maintenance Testing (Reactor Trip System Components) 4.1 Reactor Trip System Reliability (Vendor-Related Modifications) 4.2.1 and 4.2.2 Reactor Trip System Reliability (Preventive Maintenance and Surveillance Program for Reactor Trip Breakers) 4.3 Reactor Trip System Reliability (Automatic Actuation of Shunt-trip Attachment for Westinghouse and B&W plants)

Most of the remaining intermediate-term actions concern all other safety-related systems. These systems, while not sharing the same relative importance

to safety as the reactor trip system, are essential in mitigating the conse-quences of transients and accidents. Therefore, these actions should be integrated into existing plant schedules over the longer-tenn on a medium priority basis. Some of the actions discussed in the enclosure will best be i served by Owners' Group participation, and this is encouraged to the extent practical.

~

. Accordingly, pursuant to 1,0 CFR 50.54(f), operating reactor licensees and i

applicants for an operating license (this letter is for information only  ;

for those utilities that have not applied for an operating license) are requested to furnish, under oath and affinnation, no later than 120 days from the date of this letter, the status of current conformance with the positions contained herein, and plans and schedules for any needed improvements for

, conformance with the positions. The schedule for the implementation of these i improvements is to be negotiated with the Project Manager.

9

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r F.O B t% EC B M A M G M Licensees and applicants may request an extension of time for submittals of the required infomation. Such a request must set forth a proposed schedule and justification for the delay. Such a request shall be directed to the Director, Division of Licensing, NRR. Any such request must be submitted no later than 60 days from the date of this letter. 1.' : licensee or applicant does not intend to implement any of the enclosed items, the' response should so indicate and a safety basis should be provided for each item not intended to be implemented. Value-impact analys.is can be used to support such responses or to argue in favor of alternative positions that licensees might propose.

For Operating Reactors, the schedules for implementation of these actions shall I be developed consistent with the staff's goal of integrating new requirements, considering the unique status of each plant and the relative safety importance of the improvements, combined with all other existing plant programs. Therefore, schedules for implementation of these actions will be negotiated between the NRC Project Manager and licensees.

For plants undergoing. operating license review at this time, plant-specific schedules for the implementation of these requirements shall be developed 4 in a manner similar to that being used for operating reactors, taking into consideration the degree of completion of the power plant. For construction pemit holders not under OL review and for construction pemit applicants, the requirements of this letter shall be implemented prior to the issuance of an operating license.

This request for infomation was approved by the Office of Management and Budget under clearance number 3150-0011 which expires April 30, 1985.

Comments on burden and duplication may be directed to the Office of Management and Budget, Reports Management Room 3208, New Executive Office Building, Washington, D. C. 20503.

Sincerely,

$$ LC Darrell G. 'Eisenhut Director Division of Licensing .

Enclosure:

Required Actions Based on Generic Implications of Salem AlWS Events l

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O ENCLOSURE REQUIRED -ACTIONS BASED ON GENERIC IMPLICATIONS OF SALEM ATWS EVENTS 1.1 POST-TRIP REVIEW (PROGRAM DESCRIPTION AND PROCEDURE)

Position

Licensees and applicants shall describe their program for ensuring i

that unscheduled reactor shutdowns are analyzed and that a detemination is made that the plant can be restarted safely. A report describing the program for review and analysis of such unscheduled reactor shutdowns should include, as a minimum:

1. The criteria for determining the acceptability of restart.

1 2. The responsibilities and authorities of personnel who will perform the review and analysis of these events.

3. The necessary qualificat' ions and training for the responsible personnel.

4. The sources of plant information necessary to conduct the review and analysis. The sources of information should include the measures' and equipment that provide the necessary detail and

type of information to reconstruct the event accurately and in sufficient detail for proper understanding. (See Action 1.2)

5. The methods and criteria for comparing the event information with known or expected plant behavior (e.g., that safety-related equip-ment operates as required by the Technical Specifications or other performance specifications related to the safety function).

6. The criteria for determining the need for independent assessment of an event (e.g., a case in which the cause of the event

! cannot be positively identified, a competent group such as the

. Plant Operations Review Committee, will be consulted prior to authorizing restart) and guidelines on the preservation of physical -

( evidence (both har,dware and software) to support independent L analysis of the event.

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7. Items 1 through 6 above are considered to be the basis for the establishment of a systematic method to assess unscheduled reactor shutdowns. The systematic safety assessment procedures compiled from the above items, which are to be used in conducting the evaluation, should be in the report.

Applicability This position applies to all licensees and OL applicants.

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, O Type of Review For licensees, a post-implementation review of the program and procedures will be conducted or the staff will perform a pre-implementation review if desired by the licensee. NRR will perform the review and issue Safety Evaluations.

For OL applicants, the NRR review will be performed consistent with the licensing schedule.

Documentation Required Licensees and applicants shall submit a report describing their program addressing all the items in the position.

Techncial Specification Changes Required

- No-changes to Technical Specifications are required.

References Section 2.2 of NUREG-1000

-Regulatory Guide 1.33 ANSI N18.7-1976/ANS-3.2 Item I.C.5 of NUREG-0660 10 CFR 50 - 50.72 5

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1.2 POST-TRIP REVIEW - DATA AND INFORMATION CAPABILITY Position Licensees and applicants shall have or have planned a capability to record, recall and display data and information to permit diagnosing the causes of unscheduled reactor shutdowns prior to restart and for ascertaining the proper functioning of safety-related equipment.

Adequate data and infonnation shall be provided to correctly diagnose the cause of unscheduled reactor shutdowns and the proper functioning of safety-related equipment during these events using systematic safety assessment procedures (Action 1.1). The data and information shall be displayed in a form that permits ease of assimilation and analysis by persons trained in the use of systematic safety assessment procedures.

A report shall be prepared which describes and justifies the adequacy of equipment for diagnosing an unscheduled reactor shutdown. The report shall describe as a minimum:

1. Capability for assessing sequence of events (on-off indications)

1. Brief description of equipment (e.g., plant computer, dedicated computer, strip chart) .

2. Parameters monitored

3. Time discrimination between events

4. Format for displaying data and information

5. Capability for retention of data and information

6. Power source ( ) (e.g., Class IE, non-Class IE, non-interruptable

2. Capability for assessing the time history of analog variables needed to determine the cause of unscheduled reactcr shutdowns, and the ,

functioning of safety-related equipment.

1. Brief description of equipment (e.g., plant computer, dedicated computer, strip charts)

2. Parameters monitored, sampling rate, and basis for selecting parameters and sampling rate

3. Duration of time history (minutes before trip and minutes after trip)

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4. Format for displaying data including scale (readability) of time histories

5. Capability for retention of data, information, and physical evidence (both hardware and software)

6. Power source (s) (e.g., Class IE, non-Class IE, non-1 interruptable)

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3. Other' data and information provided to assess the cause of unscheduled reactor shutdowns.

4. Schedule for any planned changes to existing data and information capability.

Applicability This position applies to all licensees and OL applicants.

Type of Review Data and information capability will be reviewed by NRR to determine whether adequate data and information will be available to support the systematic safety assessment of unscheduled re. actor shutdowns. NRR will perform the reviews and issue a Safety Evaluation.

For licensees, a post-implementation review of the program and procedures will be conducted by NRR or the. staff will perform a pre-implementation review if desired by the licensee.

For OL applicants, the NRR review will be performed consistent with the licensing schedule.

Documentation Required i Licensees and applicants shall submit a report describing their data and information capability for unscheduled reactor shutdowns. .

Technical Specification Changes Required To be determined based on evaluation of required documentation.

References Section 2.2 of NUREG-1000.

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EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (REACTOR TRIP SYSTEM

, COMP 0NENTS)

Position Licensees and. applicants shall confirm that all components whose function-ing is . required .to trip the . reactor are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities, including maintenance, work ~ orders, and parts replacement. In addition, for these components, licensees and applicants shall establish, implement and maintain a continuing program to ensu're that vendor information is complete, current and controlled throughout the life of the plant,'and appropriately referenced or incorporated in plant instructions

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and procedures. Vendors of these components should be contacted and an inter-face established. Where vendors can not be identified, have gone out of business, or will not supply the information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reactor trip system reliability. The vendor . interface program shall include periodic communication with vendors to assure that all applicable information has been received. The program should use a system of positive feedback with vendors for mailings containing technical information. This '

could be accomplished by licensee acknowledgement for receipt of technical mailings. The program shall also define the interface and division of responsibilities among the licensees and the nuclear and nonnuclear divisions of their vendors that provide service on reactor trip system components to assure that requisite control of and applicable instructions for maintenance

work are provided.

Applicability This action applies to all licensees and OL applicants.

Type of Review For licensees, a post-implementation review will be conducted. NRR will perform these licensing reviews and issue a Safety Evaluation. ,

For OL applicants, the NRR review will be performed consistent with the licensing schedule.

Documentation Required Licensees and applicants should submit a statement confirming that they have reviewed the Reactor Trip System components and conform to the position regarding equipment classification. In addition, e summary report describing the vendor interface program shall be submitted for staff review. Vendor lists of technical information, and the techncial information itself, shcIl be available for inspection at each reactor site.

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UMM Technical Specification Changes Required No changes to Technical Specifications are required.

Reference Section 2.3.1 of NUREG-1000.

Section 2.3.2 of NUREG-1000.

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E AE%MEWiQ&E 2.2 EQUIPMENT CLASSIFICATION AND VENDOR INTERFACE (PROGRAMS FOR ALL SAFETY-RELATED COMPONENTS)

Position Licensees and applicants shall submit, for staff review,.a description of their programs for safety-related* equipment classification and vendor interface as- described below:

1. For equipment classification, licensees and applicants shall describe their program for ensuring that all components of safety-related systems necessary for accomplishing required safety functions are identified as safety-related on documents, procedures, and information handling systems used in the plant to control safety-related activities,

, including maintenance, work' orders and replacement parts. This

description shall include

1. The criteria for identifying components as safety-related within systems currently classified as safety-related.

This shall not be interpreted to require changes in safety classification at the systems level.

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2. A description of the information handling system used to identify safety-related components (e.g., computerized equipment list) and the methods used for its development l,

and validation.

3. A description of the process by which station personnel use this information handling system to determine that an

' activity is safety-related and what procedures for main-tenance, surveillance, parts replacement and other activitics defined in the introduction to 10 CFR 50, Appendix B, apply to safety-related components.

l 4. A description of the management controls utilized to verify

! that the procedures for preparation, validation and routine I utilization of the information handling system have been l followed.

5. A demonstration that appropriate design verification and .

qualification testing is specified for procurement of safety-related components. The specifications shall include quali-fication testing for expected safety service conditions and provide support for the licensees' receipt of testing documen-tation to support the limits of life recommended by the supplier.

• 5af ety-related structures, systems, and components are those that are relied

.upon to remain functional during and following design basis events to ensure:

(1) the integrity of the reactor coolant boundary, (2) the capability to shut down the reactor and maintain it in a safe shutdown condition, and (3) the capability to prevent or mitigate the consequences of accidents that could result in potential offsite exposures ccmparable to the guidelines of 10 CFR Part 100.

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6. Licensees and applicants need only to submit for staff review the equipment classification program for safety-related components'. Although not required to be submitted for staff review, your equipment classification program should also include the broader class of structures, systems, and components important to safety required by GDC-1 (defined irl 10 CFR Part 50, Appendix A, " General Design Criteria, Introduction").

2. For vendor interface, licensees and applicants shall establish, implement and maintain a continuing program to ensure that vendor infonnation for safety-related components is complete, current and controlled throughout the life of their plants, and appropriately referenced or incorporated in plant instructions and procedures.

Vendors of safety-related equipment should be contacted and an interface established. Where vendors cannot be identified, have gone out of business, or will not supply information, the licensee or applicant shall assure that sufficient attention is paid to equipment maintenance, replacement, and repair, to compensate for the lack of vendor backup, to assure reliability commensurate with its safety function (GDC-1). The program shall be closely coupled with action 2.2.1 above (equipment qualification). The program shall include periodic communication with vendors to assure that all applicable information has been received. The program should use a system of positive feedback with vendors for mailings containing technical information. This could be accomplished by licensee acknowledgment for receipt of technical ma'ilings. It shall also define the interface and division of responsibilities among the licensee and the nuclear and nonnuclear divisions of their vendors that provide service on safety-related equipment to assure that requisite control of and applicable instructions for maintenance work on safety-related equipment are provided.

Applicability l This action applies to all licensees and OL applicants.

Type of Review For licensees, a post-implementation review will be conducted. NRR will l perfonn the review and issue a Safety Evaluation.

For OL applicants, the NRR review will be perfonned consistent with the licensing schedule.

j Documentation Required Licensees and applicants should submit a report that describes the equipment classification and vendor interface programs outlined the position above, l

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U UMMQ Technical Specification Changes Required No changes to the Technical Specifications are required.

References -

Section 2.3.1 of NUREG-1000.

Section 2.3.2 of NUREG-1000.

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f9R Ef. M Ni 3.1 POST-MAINTENANCE TESTING (REACTOR TRIP SYSTEM COMPONENTS)

Position, The following actions are applicable to post-maintenance testing:

1. Licensees and applicants shall submit the results of their review of test and maintenance procedures and Technical Specifications to assure that post-maintenance operability testing of safety-related components in the reactor trip system is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

2. Licensees and applicants shall submit the results of their check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications, where required.

3. Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing Technical Specifications which can be demonstrated to degrade rather than enhance safety.

Appropriate changes to'these test requirements, with supporting justification, shall be submitted for staff approval. (Note that action 4.5 discusses on-line system functional testing.)

Applicability This action applies to all licensees and OL applicants.

Type of Review For licensees, a post-implementation review will be conducted for actions 3.1.1 and 3.1.2 above. The Regions will perform these licensing reviews and issue Safety Evaluations. Proposed Technical Specification changes resulting from action 3.1.3 above will receive a pre-implementation review by NRR.

For OL applicants, the review will be perfonned consistent with the l licensing schedule.

Documentation Required Licensees and applicants should submit a statement confirming that actions 3.1.1 and 3.1.2 of the above position have been implemented.

Technical Specification Changes Required Changes to Technical Specifications, as a result of action 3.1.3, are to be determined by the licensee or applicant and submitted for staff approval, as n;cessary.

Reference Section 2.3.4 of NUREG-1000. ..

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3.2'. POST-MAINTENANCE TESTING (ALL.0THER SAFETY-RELATED COMPONENTS)

Position

• The following actions are applicable to post-maintenance testing:

1. Licensees and applicants shall submit a report documenting the

extending of test and maintenance procedures and Technical

. Specifications review to assure that post-maintenance operability 4

testing of all safety-related equipment is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service.

2. Licensees and applicants shall submit the results of their check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or the Technical Specifications where required.

3. Licensees and applicants shall identify, if applicable, any post-maintenance test requirements in existing Technical Specifications which are perceived to degrade rather than enhance safety. Appropriate changes to these test requirements, with supporting justification, shall be submitted for staff approval.

Applicability This action applies to all licensees and OL applicants.

Type of Review For licens'ees, a post-implementation review will be conducted for actions 3.2.1 and 3.2.2 above. The Regions will perform these licensing reviews and issue Safety Evaluations. Proposed Technical Specification changes resulting from action 3.2.3 above will receive a pre-implementation review by NRR.

For 0L applicants, the review will be performed consistent with the

-licensing schedule.

Documentation Required Licensees and applicants should submit a statement confirming that actions 3.2.1 and 3.2.2 of the above position have been implemented.

. Technical Specification Changes Required Changes to Technical Specifications, as a result of action 3.2.3, are to be determined by the licensee or applicant for staff approval, as necessa ry.

Reference Section 2.3.4 of NUREG-1000.

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r W WMW 4.1 REACTOR TRIP SYSTEM RELIABILITY (VENDOR-RELATED MODIFICATI0HS)

Position All vendor-recommended reactor trip breaker modifications shall be reviewed to verify that either: (1) each modification has, in fact, been implemented; or (2) a written evaluation of the technical reasons for not implementing a modification exists.

For example, the modifications recommended by Westinghouse in NCD-Elec-18 for the DB-50 breakers and a March 31, 1983, letter for the DS-416 breakers shall be implemented or a justification for not implementing shall be made available. Modifications not pr eviously made shall be incorporated or a written evaluation shall be provided.

' Applicability This action applies to all PWR licensees and OL applicants.

Type of Review For licensees, a post-implementation review will be conducted. The Regions will perform these licensing reviews and issue Safety Evaluations.

For OL applicants, the NRR review will be performed consistent with the licensing schedule.

Documentation Required Licensees and applicants should submit a statement confirming that this action has been implemented.

- Technical Specifications Required l No changes to Technical Specifications are required.

Reference Section 3 of NUREG-1000.

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r E M 4.2 REACTOR TRIP SYSTEM RELIABILITY (PREVENTATIVE MAINTENANCE AND SURVEILLANCE PROGRAM FOR REACTOR TRIP BREAKERS)

Position Licensees and applicants shall describe their preventative maintenance

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and suryeillance program to ensure reliable reactor trip breaker operation.

The program shall include the following:

1. A planned program of periodic maintenance, including lubrication, housekeeping, and other items recommended by the equipment supplier.

2. Trending of parameters affecting operation and measured during testing to forecast degradation of operability.

3. Life testing of the breakers (including the trip attachments) on an acceptable sample size.

4. Periodic r,epl9 cement of breakers or components consistent with demonstrated life cycles.

( Applicability This action applies to all PWR licensees and OL applicants.

Type of Review L Actions 4.2.1 and 4.2.2 will receive a post-implementation review by l

NRR. A pre-implementation review will be performed by NRR for actions l

4.2.3 and 4.2.4 (the circuit breaker life testing program and the com-ponent testing / replacement requirements based upon the life testing results). A Safety Evaluation will be issued.

, For OL applicants, NRR will perform the reviews for actions 4.2.1 and l' 4.2.2 on a schedule consistent with the licensing schedule. NRR will.

perfonn a pre-implementation review for actions 4.2.3 and 4.2.4 (the i circuit breaker life testing program and the component testing / replace-j ment requirements based upon the life testing results). Safety

• l Evaluations will be issued.

Documentation Required Licensees and applicants should submit descriptions of their programs to ensure compliance with this action.

i Technical Specification Changes Required l No changes to Technical Specifications are required.

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! Reference

Section 3 of NUREG-1000.

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'4.3 REACTOR TRIP SYSTEM RELIABILITY ( AUTOMATIC ACTUATION OF SHUNT TRIP ATTACHMENT FOR WESTINGHOUSE AND B&W PLANTS)

Position Westinghouse and B&W reactors shall be modified by providing automatic reactor trip system actuation of the breaker shunt trip attachments.

The shunt trip attachment shall be considered safety related (Class IE).

Applicability This action applies to all Westinghouse and B&W licensees and OL applicants.

• Type of Review For licensees, a pre-implementation review shall be performed for the

- design modifications by NRR. A Safety Evaluation will be issued.

For OL applicants, the NRR review will be perfomed consistent with the lice sing schedule.

Technical Specification changes, if required, will be reviewed prior to implementation.

Documentation Required Licensees and applicants should submit a report describing the modifications.

Technical Specification Changes Required Licensees are to submit any needed Technical Specification change requests prior to declaring the modified system operable.

Reference Section 3 of NUREG-1000.

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g. pABEORMATMN 4.4 REACTOR TRIP SYSTEM RELIABILITY (IMPROVEMENTS IN MAINTENANCE AND TEST PROCEDURES- FOR B&W PLANTS)

Position Licensees and applicants with B&W reactors shall apply safety-related maintenance and ' test procedures to the diverse reactor trip feature provided by interrupting power to control rods through the silicon controlled rectifiers.

This action shall not be interpreted to require hardware changes or additional environmental or seismic qualification of these components.

Applicability This action applies to B&W licensees and OL applicants only.

Type of Review For licensees, a post-implementation review will be conducted. The Regions will conduct the licensing review and i~ssue a Safety Evaluation.

For OL applicants, the review will be perfomed consistent with the licensing schedule.

Documentation Required Licensees and applicants should submit a statement confirming that this action has been implemented.

Technical Specification Changes Required

! Include the silicon controlled rectifers in the appropriate surveillance and test sections of the Technical Specifications.

! Reference -

Section 3 of NUREG-1000.

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4.5 REACTOR TRIP SYSTEM RELIABILITY (SYSTEM FUNCTIONAL TESTING)

Position On-line functional testing of the reactor trip system, including independent testing of the diverse trip featuret, shall be perfonned on all plants. -

, 1. The diverse trip features to be tested include the breaker undervoltage and shunt trip features on Westinghouse, B&W (see Action 4.3 above) and CE plants; the circuitry used for power interruption with the silicon controlled rectifiers on B&W plants (see Action 4.4 above); and the scram pilot valve and backup scram valves (including all initiating circuitry) on GE plants.

2. Plants not currently designed to pemit periodic on-line testing shall justify not making modifications to permit such testing.

Alternatives to on-line testing proposed by licensees will be considered where special circumstances exist and where the objective of high reliability can be met in another way.

3. Existing intervals for on-line functional testing required by

. Technical- Specifications shall be reviewed to determi-- . .at

• the intervals are consistent with achieving high react.- trip

system availability _when accounting for considerations such '

, as:

1. uncertainties in component failure rate's

2. uncertainty in common mode failure rates

3. reduced redundancy during testing

4. operator errors during testing

5. component " wear-out". caused by the testing Licensees currently not perfoming periodic on-line testing shall determine appropriate test intervals as described above. Changes to l existing required intervals for on-line testing as well as the l intervals to be detemined by licensees currently not performing l on-line testing shall be justified by information on the sensitivity -

l of reactor trip system availability to parameters such as the test l intervals, component failure rates, and common mode failure rates.

Applicablity '

This action applies to all licensees and OL applicants.

Type of Review

( For licensees, a post-implementation review will be conducted for action 4.5.1. The Regions will perfom these ifcensing reviews and issue

! Safety Evalu6tions. Actions 4.5.2 and 4.5.3 will require a pre-implemen-

.tation review by NRR. Results will be issued in a Safety Evaluation.

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