ML18102A830

Forwards Response to Request for Info,Per 10CFR50.54(f) Re Adequacy & Availability of Design Bases Info for Salem Generating Station,Units 1 & 2

ML18102A830
Person / Time
Site:	Salem
Issue date:	02/11/1997
From:	Eric Simpson Public Service Enterprise Group
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
LR-N970074, NUDOCS 9702140145
Download: ML18102A830 (144)

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' Public Service Electric and Gas

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E. C. Simpson Public Service Electric and Gas Company P.o: Box 236, Hancocks Bridge, NJ 08038 609-339-1700 Senior Vice President - Nuclear Engineering February 11 , 1997 LR-N970074 United States Nuclear Regulatory Commission

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Document Control Desk I Washington, DC 20555 "REQUEST FOR INFORMATION PURSUANT TO 10CFR50.54(f) REGARDING ADEQUACY AND AVAILABILITY OF DESIGN BASES INFORMATION" SALEM GENERATING STATION UNITS 1 AND 2 DOCKET NOS. 50-272 AND 50-311 Gentlemen:

By letter dated October 9, 1996, the Nuclear Regulatory Commission requested PSE&G to submit information that will provide the NRC with added confidence and assurance that our nuclear facilities are operated and maintained within the design bases and any deviations are reconciled in a timely manner. As requested, PSE&G is submitting our response to the 10CFR50.54(f) letter within 120 days of its receipt. The two attachments contain PSE&G's response to the subject letter for Salem Generating Station Units 1 & 2.

PSE&G agrees it is essential to have programs in place that ensure the Salem Units are configured and operated in accordance with their design bases. Further, PSE&G is committed to assuring that there are programs and procedures in place that adequately support the maintenance of the design bases. PSE&G has undertaken a number of specific validation efforts to provide reasonable assurance that the design bases are maintained and plans to conduct additional reviews to provide further assurance. In addition, the new Nuclear Business Unit (NBU) culture, which focuses on the identification of problems through the use of critical self assessments and questioning attitude, will ensure problems are resolved in an effective and timely manner in accordance with our revamped Corrective Action Program.

The detailed responses to each of the five specific information requests are provided in Attachment 1. Attachment 1 also includes a detailed Table of Contents and Executive Summary for your convenience. It should be noted that while the discussion of processes and programs contained herein provide an accurate description of their present state, PSE&G will continue to revise them in accordance with approved 97021~"0145 970211** -- .... *- .. -

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• 2 LR-N970074 revision processes as further enhancements are identified, without modifying this response. In order to eliminate any ambiguity with regard to commitments contained in this response, Attachment 2 describes the PSE&G specific commitments related to this request.

Based on the detailed information provided herein, PSE&G has concluded that there is reasonable assurance that Salem Units 1 and 2 will be operated in accordance with their design bases. The future actions discussed in Attachment 2 will provide additional confirmation of compliance with design bases. If there are any questions regarding this information, we will be pleased to discuss these with you.

Attachments(2) c Mr. Hubert J. Miller, Administrator - Region I U. S. Nuclear Regulatory Commission 475 Allendale Road King of Prussia, PA 19406 Mr ..L. N. Olshan, Licensing Project Manager U. S. Nuclear Regulatory Commission One White Flint North 11555 Rockville Pike Rockville, MD 20852 Mr. C. Marschall - Salem (X24)

USNRC Senior Resident Inspector Mr. K. Tosch, Manager, IV Bureau of-Nuclear Engineering 33 Arctic Parkway CN 415 Trenton, NJ 08625

REF: LR-N97007 4 STATE OF NEW JERSEY

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COUNTY OF SALEM E. C. Simpson, being duly sworn according to law deposes and says:

I am Senior Vice President - Nuclear Engineering of Public Service Electric and Gas Company, and as such, I find the matters set forth in the above referenced letter, concerning the Salem Generating Station, are true to the best of my knowledge, information and belief.

Subscribed and Sworn to before me

\his I I J ,_ day of ~Al.!.(!L*r 1997

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Notary Public r/( New Jersey *

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SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 ATTACHMENT 1

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Table of Contents

• EXECUTIVE

SUMMARY

..................................................................................................................... 1 INTRODUCTION ................................................................................................................................... 5 ADEQUACY OF DESIGN BASES INFORMATION ........................................................................... 8 RESPONSE TO QUESTION A ............................................................................................................ 11 I. OVERVIEW ............................................................................................................................................ 11 IL APPENDIXB TO 10CFR50 .................................................................................................................... ll III. DESCRIPTION OF DESIGN AND CONFIGURATION CHANGE PROCESSES .................................................... 12 A. Control ofDesign and Configuration Change, Tests and Experiments (NC.NA-AP.ZZ-OOOB(Q)) .... 13 B. Minor Modification Process (NC.NA-AP.ZZ-0017(Q)) ................................................................... 19 C. Temporary Modification Process (NC.NA-AP.ZZ-0013(Q)) ........................................................... 21 D. Nuclear Fuel Management Program (NC.NA-AP.ZZ-0072(Q)) ....................................................... 22 IV. DESCRIPTION OF CONFIGURATION CONTROL PROCESSES ...................................................................... 23 A. Nuclear Procedure System (NC.NA-AP.ZZ-OOOl(Q)) ...................................................................... 23 B. Managed Maintenance Information System (ND.DE-TS.ZZ-5409(Q)) ............................................. 25 C. Work Control, Safety Tagging and Post Maintenance Testing Programs ........................................ 25 D. Engineering Analysis ..................................................................................................................... 28 E. Design Drawings Process (NC.DE-AP.ZZ-0004(Q)) ....................................................................... 29 F Software Control (NC.DE-AP.ZZ-0052(Q)) .................................................................................... 30 G. Procurement Process (NC.NA-AP.ZZ-0019(Q)) .............................................................................. 31 H. Vendor Technical Document Control Program (NC.DE-AP.ZZ-0006(Q)) ....................................... 32 L Technical Specification Surveillance Program (NC.NA-AP.ZZ-0012(Q)) ......................................... 32 V. DOCUMENT CONTROL PROCESS (NC.NA-AP.ZZ-0003(Q)) ................................................................... 33 VI. 10CFR50.59 PROCESS (NC.NA-AP.ZZ-0059(Q)) ............................................................................... 34 VII. TECHNICAL SPECIFICATION AMENDMENT PROCESS (NC.LR-AP.ZZ-OOOS(Z)) ..................................... 35 VIII. 10 CPR 50.7l(E) UFSAR UPDATE PROCESS (NC.LR-AP.ZZ-OOB(Z)) ............................................. 36 IX. SPECIFIC TECHNICAL PROGRAMS ........................................................................................................ 3 8 A. Equipment Environmental Qualification Program (DE-PS.ZZ-0002(Q)) ......................................... 38 B. Motor Operated Valve Program (NC.DE-PS.ZZ-0033(Q)) .............................................................. 39 C. Fire Protection Program (NC.DE-PS.ZZ-0001(Q)) ......................................................................... 40 D. Inservice Testing & Inservice Inspection Programs (NC.NA-AP.ZZ-0027(Q) and NC.NA-AP.ZZ-0070(Q)) ............................................................................................................................................. 41 X. INDEPENDENT OVERSIGHT ................................................................................................................... 42 A. Station Operations Review Committee (NC.NA-AP.ZZ-0004(Q)) ................................................... .42 B. Off-Site Safety Review (ND.SN-AP.ZZ-0001(Q)) ............................................................................. 42 C. On-Site Safety Review (ND.SN-AP.ZZ-0001 (Q)) ............................................................................. 43 D. Quality Assurance.......................................................................................................................... 43 XI. OTHER PROCESSES WITH THE POTENTIAL TO AFFECT CONFIGURATION CONTROL ................................. 43 A. Technical Specification Interpretations (SC.OP-DD.ZZ-AD45(Z)) .................................................. 44 B. Department Night Orders (NC.NA-AP.ZZ-0005(Q)) ....................................................................... 44 C. Control of Operator Aids (NC.NA-AP.ZZ-0044(Q)) ........................................................................ 44 D. On-the-Spot Changes (NC.NA-AP.ZZ-OOOl(Q)) ............................................................................ 45 E. Equipment Control Program (SC.OP-AP.ZZ-0006(Q) and SC.OP-AP.ZZ-OlOB(Q)) ....................... .45 XII. AVAILABILITY OF DESIGN BASES DOCUMENTATION .......................................................................... .45 XIII. PERSONNEL PERFORMANCE IMPROVEMENT ....................................................................................... 47

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Table of Contents XIV.

SUMMARY

....................................................................................................................................... 48 RESPONSE TO QUESTION B ............................................................................................................ 50 I. OVERVIEW ............................................................................................................................................ 50 II. PROCESS CONTROLS ............................................................................................................................ 51 III. vALIDATION REVIEWS ........................................................................................................................ 53 A. Procedure Upgrade Project (PUP) ................................................................................................. 53 B. Emergency Operating Procedure (EOP) Upgrade Project.............................................................. 54 C. Technical Specification Surveillance Improvement Project ............................................................ 55 D. Salem System Readiness Review .................................................................................................... 56 E. Integrated Test Program ................................................................................................................ 58 F. FSARProject ................................................................................................................................. 59 G. Instrument Valve Lineup Procedures.............................................................................................. 67 IV. ASSESSMENT OF EFFECTIVENESS ............................................................................................. 68 A. Design/Configuration Change ........................................................................................................ 68 B. Configuration Control .................................................................................................................... 69 C. Technical Specification (J'S) Implementation ................................................................................. 75 D. Specific Technical Programs .......................................................................................................... 76 E. 10CFR50.59 Process (NC.NA-AP.ZZ-0059(Q): NAP-59) ................................................................ 77 F UFSARMaintenance ...................................................................................................................... 80 G. Procedure Control ......................................................................................................................... 82 V. CONCLUSION ................................................................................................................................... 83 RESPONSE TO QUESTION C ............................................................................................................ 85 I. OVERVIEW ............................................................................................................................................ 85 II. PROCESS CONTROLS ............................................................................................................................ 87 A. Engineering Controls ..................................................................................................................... 87 III. VALIDATION EFFORTS ........................................................................................................................ 89 A. Salem Generating Station (SGS) Process for Achieving Restart...................................................... 89 B. Salem Configuration Walkdowns .................................................................................................... 90 C. Salem Generating Station (SGS) System Readiness Review Process ............................................... 93 D. Design Bases Review I FSAR Project ........................................................................................... 101 IV. ASSESSMENT OF EFFECTIVENESS ....................................................................................................... 111 A. INDEPENDENT ASSESSMENTS ................................................................................................. 111 V. IMPROVEMENTS IMPLEMENTED OR PLANNED ...................................................................................... 116 VI. CONCLUSION ................................................................................................................................... 117 RESPONSE TO QUESTION D .......................................................................................................... 118 I. CORRECTIVE ACTION PROGRAM .......................................................................................................... 118 A. Salem Action Request Process ...................................................................................................... 118 B. Action Request Process Operation ............................................................................................... 119 C. Conditions Adverse to Quality ..................................................................................................... 120 D. Operability Determinations ......................................................................................................... 121 E. Reportability 10CFR50.72 & 10CFR50.73 ................................................................................... 122 F Action Closure ............................................................................................................................. . 122 G. Training and Responsibility Identification ................................................................................... 123 II. OPERATING EXPERIENCE PROGRAM .................................................................................................... 123 III. 10CFRPART 21 PROCESS ............................................................................................................... 124 ii

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Table of Contents

• IV. EMPLOYEE CONCERNS PROGRAM .................................................................................................. ~ ... 125 V. EFFECTIVENESS ................................................................................................................................. 125 A. Program Requirements................................................................................................................. 125 B. Results ........................................................................................................................................ . 126 C. Inspection Results ............................................-............................................................................ 127 VI. DISCUSSION OF RECENT FINDINGS .................................................................................................... 129 VII. CONCLUSION********************* ............................................... ****************************** ................................ 130 RESPONSE TO QUESTION E .......................................................................................................... 131 iii

SALEM GENERATING ;STATION Response to 10 CFR 50.54(f) Letter

• • Dated October 9, 1996 EXECUTIVE

SUMMARY

This attachment provides PSE&G's response to the NRC letter of October 9, 1996, in which information was requested pursuant to 10CFR50.54(f) regarding the adequacy and availability of design bases information at the. Salem Generating Station. Specifically, the NRC's letter requested the following information:

A. Description of engineering design and configuration control processes, including those that implement 10CFR50.59, 10CFR50.71(e), and Appendix B to 10CFR Part 50; B. Rationale for concluding that design bases requirements are translated into operating, maintenance, and testing procedures; C. Rationale for concluding that system, structure, and component configuration and performance are consistent with the design bases; D. Processes for identification of problems and implementation of corrective actions, including actions to determine the extent of problems, action to prevent recurrence, and reporting to the NRC; and*

E. The overall effectiveness of current processes and programs in concluding that the configuration of the plant is consistent with the design bases.

The response was prepared by a team of several dedicated, experienced personnel. The team focused on review of (1) processes that control the design and configuration of the plant (2) numerous validation activities and various independent assessments, both internal and external, to assess the translation of design bases into plant configuration and operations and (3) the plant corrective action process. In addition to the requested information, a separate discussion is provided that gives reasonable assurance that Salem's current design bases are adequate and accurate.

Salem Generating Station was designed in the late 1960s and early 1970s. The Salem design was baselined in the early 1990s, when Configuration Baseline Documents (CBDs) were prepared for 44 safety significant and risk significant systems. Consequently, the response team focused primarily on review activities undertaken since preparation of the CBDs and especially on activities conducted during the extended outages beginning in mid-1995 .

• The "Response to Question A" section describes the major processes currently used to maintain plant design and operations consistent with the design bases.

1 Executive Summary

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 These processes include the configuration control process, design and configuration change process, document control process, 10CFR50.59 safety evaluation process, Technical Specification amendment process, and the 10CFR50.71(e) UFSAR update process. "Response to Question A" also describes human performance initiatives to improve personnel performance relative to understanding and sensitivity to design bases issues.

The "Response to Question B" section describes the various validation reviews and independent assessments performed over the past several years which provide us reasonable assurance that design bases information has been adequately translated into plant procedures. These reviews include the procedure upgrade project, the emergency operating procedure upgrade project, the Technical Specification surveillance improvement project, the Salem readiness review program and the FSAR Project (including 7 vertical-slice reviews). The "Response to Question B" section also summarizes key independent assessment findings, both strengths and weaknesses, to evaluate the effectiveness of process controls discussed in "Response to Question A".

The "Response to Question B" section concludes, based on recent plant procedure upgrades and the results of a large, representative sample of design bases information, there is reasonable assurance that design bases requirements have been adequately translated into Salem procedures. As additional issues are discovered through continuing review activities, they will be addressed and resolved in accordance with our Corrective Action Program and plant Technical Specifications.

The "Response to Question C" section describes validation efforts and independent assessments which provide reasonable assurance of consistency between the design bases and plant configurations. These activities include the Salem FSAR Project, the System Readiness Review Process, the Configuration Walkdown Project, and the Integrated Test Program. The "Response to Question C" section also summarizes key assessment findings, both strengths

• and weaknesses, to evaluate the effectiveness of process controls described in "Response to Question A".

The "Response to Question C" section concludes that, upon restart, PSE&G has reasonable assurance that structure, system, and component configuration and performance are consistent with the plant design bases. As additional design bases or configuration issues are discovered through continuing review activities, they will be addressed and resolved in accordance with our Corrective Action Program and plant Technical Specifications.

2 Executive Summary

SALEM G_ENERATIN_G_,STATION Response to 10 CFR 50.54(f) Letter

• Dated October 9, 1996 The Salem Corrective Action Program is described in "Response to Question D",

including the Action Request process, operating experience feedback process, 10CFR, Part 21 process, and the employee concerns program. The Corrective Action Program was substantially revised in the 1994/1995 time frame as a result of inspection findings which questioned the effectiveness of the previous system. Throughout 1996, numerous assessments and inspections were conducted by internal and external organizations, including the USNRC. The new corrective action program, implemented in July 1995, was demonstrated to be generally effective, including the ability to identify and correct design related issues. While improvements are still needed relative to the timeliness and effectiveness of corrective actions, the new Corrective Action Program has been effective in resolving safety significant issues that have arisen.

In "Response to Question E", PSE&G states that sufficient information has been presented to conclude that, upon plant restart, there is reasonable assurance the Salem Generating Station will be configured, operated and maintained within its design bases. As additional design or design bases related issues are discovered through continuing review activities, they will be addressed and resolved in accordance with our Corrective Action Program and plant Technical Specifications.

In general, PSE&G believes its corrective action program, document control and retrievability, and operating procedures are strong, but that UFSAR consistency can continue to be improved, CBDs should be updated to reflect recent plant changes, and engineering-staff design bases knowledge can be improved. To provide additional assurance on compliance with the design bases, a comprehensive design bases review project will be undertaken over the next two years, as further discussed in Attachment 2.

In Attachment 2, PSE&G outlines our program to continue correcting deviations or deficiencies in design bases documentation, as they are identified, and improving the consistency of the information with plant configuration and procedures. A formal submittal will be placed on the docket within 60 days of this letter, providing program details and schedules.

We intend to conduct the proposed activity in system teams comprised of our system managers, system design engineers, system SROs, and other support personnel as required. The system teams will* conduct further adequacy reviews of design bases information and documentation, and the translation of the design bases into operations and procedures, in order to both gain knowledge and the ability to revise and update the UFSAR and CBDs. As in the case of our 1996 FSAR Project, systems will be classified as safety analysis, risk significant,

3. Executive Summary

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 risk important, or other. The level and extent of reviews will vary depending upon system classification.

4 Executive Summary

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 INTRODUCTION Salem Generating Station was designed in the late 1960's and early 1970's with receipt of the construction permit on September 25, 1968. Salem Unit 1 began operation on August 13, 1976, with the issuance of Facility Operating License No. DPR-70, which authorized fuel loading and low power testing. Salem Unit 2 began operation on April 18, 1980, with the issuance of Facility Operating License No. DPR-75, which authorized fuel loading and low power testing.

Commercial operation for Salem Units 1 and 2 began on June 30, 1977 and October 13, 1981, respectively. For the most part, the units are identical, except that some post-TMI actions, which were not back-fit into Unit 1, were implemented in the Unit 2 design. The Final Safety Analysis Report (FSAR) is common to both units and both units pre-date the NRC's Standard Review Plan (SRP).

The Salem Units are 4-loop Westinghouse Pressurized Water Reactor (PWR) units. Public Service Electric and Gas Company was its own architect-engineer firm for the design and construction of those portions of the units not directly supplied by Westinghouse. Consequently, a considerable volume of original design bases information has always been held by PSE&G, with Nuclear Steam Supply System (NSSS) design bases information held either by PSE&G or Westinghouse.

By letter dated October 9, 1996, the NRC issued to licensees a request for information pursuant to 10CFR50.54(f) regarding the adequacy and availability of design bases information. The NRC specifically requested licensees to provide the following information:

1. Description of engineering design and configuration control processes, including those that implement 10CFR50.59, 10CFR50.71(e), and Appendix B to 10CFR Part 50;

2. Rationale for concluding that design bases requirements are translated into operating, maintenance, and testing procedures;

3. Rationale for concluding that system, structure, and component configuration and performance are consistent with the design bases;

4. Processes for identification of problems and implementation of corrective actions, including actions to determine the extent of problems, action to

• prevent recurrence, and reporting to NRC, and 5 Introduction

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996

5. The overall effectiveness of your current processes and programs in concluding that the configuration of the plant is consistent with the design bases.

The NRC further requested that, in responding to items (a) through (e),

licensees indicate whether any design review or reconstitution programs have been undertaken, and if not, provide a rationale for not implementing such a program.

This attachment provides Salem Generating Station's response to the request for additional information. The response was prepared by a team of several dedicated, experienced personnel who interfaced with several organizations in the Nuclear Business Unit (NBU). The team focused on (1) design, configuration control and corrective action processes (2) effective translation of design bases information into the plant configuration and operations and (3) independent assessments of configuration or operations in accordance with the design bases.

The organization of this response follows the order shown above, and is further detailed in the Table of Contents. In addition to the requested information, a section entitled "Adequacy of Design Bases" was added to provide reasonable assurance that Salem's current design bases are adequate and accurate. This section provides the cornerstone for the rest of the response. The validation reviews presented in response to questions (b) and (c) are focused on activities conducted in the past two years that were conducted to assess and support the restart efforts of the Salem Units. These efforts involved a number of process changes and enhancements to the configuration control process.,

The "Response to Question A" section provides a description of the major processes currently used to maintain plant design and operations consistent with the design bases. These descriptions are summaries of our processes intended to convey a sense of how they work, and are not intended to provide all procedural details. These processes are not static and may change in the future as enhancements are identified. Included in these descriptions are process and organizational interfaces, and the elements intended to satisfy 10CFR50.59, 10CFR50.71(e) and support 10CFR50, Appendix B, requirements.

The "Response to Question B" section provides the rationale for our reasonable assurance that the design bases are adequately translated into the operation of the facility. This section will discuss the fidelity of the implementing processes used to control and maintain the design bases, such that it will address the procedure control process and other processes such as work control. This section will present major validation efforts used to provide reasonable 6 Introduction

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 assurance that the design bases have been adequately translated and maintained into the implementing controls. In addition, this section will build upon the question (a) process discussion by highlighting the effectiveness of the processes used to translate the design bases into operation of the facility.

The "Response to Question C" section describes validation efforts and independent assessments associated with design bases consistency with plant configuration and operation. The purpose of this section is to provide the rationale for our reasonable assurance that system, structure, and component (SSC) configuration and performance are consistent with the design bases. As with the response to question (b), this section will expand on the processes described in "Response to Question A" by identifying and demonstrating how key attributes of configuration control are fulfilled at Salem Generating Station (SGS).

The Salem corrective action program is described in the section entitled "Response to Question D" including a discussion on the Action Request Process, Operating Experience Program, 10CFR21 process, and the Employee Concerns Program. This section provides a brief overview of the operation of the Action Request Process (issue identification and prioritization process) and the Corrective Action Program including conditions adverse to quality, operability determinations, reportability determinations and action closure. A summary of how the Operating Experience Program adds external information to the process is included. The important role of the Employee Concerns Program for identifying internal issues for entry into the Action Request Process is also described. In addition, information is provided on the effectiveness of PSE&G's implementation of these programs.

The "Response to Question E" section provides an overall conclusion to the response. The effectiveness of the processes are more appropriately addressed in the individual responses to questions (b), (c), and (d).

The requirement to address the performance of a design review or reconstitution is addressed in the "Response to Question C" section by describing the Configuration Baseline Document project that was conducted in the 1990 timeframe. Future activities are outlined in Attachment 2 .

• 7 Introduction

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 ADEQUACY OF DESIGN BASES INFORMATION The purpose of this section is to provide a brief overview of the rationale for reasonable assurance that design bases information is adequate. The NRC's request for information pursuant to 10CFR50.54(f) principally focuses on translation of design bases information into the actual plant configurations and plant procedures. As further described in this section of the response, PSE&G's rationale for translation adequacy relies upon results from various reviews and independent assessments of the Salem configuration control process. Many of these reviews and assessments not only provide reasonable assurance of translation of design basis information into the configuration and procedures but also tend to provide reasonable assurance of design bases adequacy, as further discussed below.

Design bases information is defined in 10CFR50.2 as that information which identifies the specific functions to be performed by a structure, system or component of a facility, and the specific values or ranges of values chosen for controlling parameters as reference bounds for design. These values are typically derived in analyses or calculations or may be restraints derived from generally accepted practices for achieving functional goals, such as limits imposed by specifications, industry standards, manufacturer's technical manuals or data sheets.

From 1991 to 1994, 44 Configuration Baseline Documents (CBDs) were developed for Salem Station, covering systems and structures which were considered safety related, technical specification related, or important to safety.

The CBDs identify the system or structure design bases. Equally as important as the specific design values themselves, the CBDs also provide reference to the various source documents (calculations, analyses, specifications vendor drawings, etc.) from which the design bases values are derived. At the time they were prepared, the CBDs were peer reviewed and approved, such that they met ANSI Standard design input document requirements.

The CBD project was an extensive effort to compile design bases information for Salem. Design bases information was collected, compared to the as-built configuration, and recorded. Where discrepancies were noted, Discrepancy Evaluation Forms (DEFs) were prepared, evaluated, and resolved or scheduled for subsequent resolution. As such, the CBD project provided a baseline for the Salem design, a baseline which has been (1) generally preserved since that time by the current Salem configuration control processes and (2) has been evaluated by numerous reviews, inspections, and assessments.

8 Adq of Design Bases

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Where design calculations were not available to document a particular system design feature or parameter, DEFs were generally prepared which subsequently resulted in calculations or other engineering activities to document or otherwise confirm the design requirement. Where existing calculations were available, the CBD project identified these calculations and determined that they had been verified and approved in accordance with existing procedures. The project then used calculation output, where appropriate, in preparing the CBD. The project did not review these calculations for technical adequacy nor did it attempt to verify all input and assumptions as being appropriate.

Consequently, while the CBD project provides additional assurance that system design bases meet requirements and commitments, the type of comprehensive design reviews typically associated with safety system functional inspections or similar design reviews or inspections were not conducted. However, when combined with the results of independent inspections and other assessments which do further validate design bases technical adequacy, the CBD project fills an important role in that (1) the source of necessary design information was located, (2) missing information was identified and obtained or reconstituted.and (3) consistency with design commitments was evaluated .

• In view of the above, PSE&G has reasonable assurance that the Salem design bases are adequate to assure systems, structures, and components can perform their intended safety functions. This reasonable assurance is based upon:

• The CBD project identified and verified design bases information.

Deficiencies were addressed or are being resolved through the DEF closure process. A sample DEF closures were reviewed and found acceptable by the Salem FSAR Project in 1996.

• The configuration control processes, including the corrective action process, have been demonstrated effective by various reviews and assessments including the FSAR Project and two NRC special team inspections, all in 1996.

• A large number of in-depth reviews and assessments in the 1995/1996 time frame, generally demonstrated the Salem design bases to be adequate.

Further discussion of these reviews and assessments (listed below) are given in "Response to Question B" and "Response to Question C".

1. FSAR Project Seven Vertical Slice Reviews Chapter 15 Input and Assumption Validation

• DEF Closure Reviews

2. Configuration Walkdown Program 9* Adq of Design Bases

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996

3. System Readiness Review

4. Integrated Test Program for Restart

5. CCW System SSFI

• Where design related problems were identified by these reviews (such as CFCU response time or CCW pump runout), they were more related to implementation of the design than design basis adequacy. In addition, these items were promptly resolved or corrected by the Salem corrective action program. For example, the CFCU issue identified by the FSAR Project dealt with response time testing, while CCW pump runout was subsequently shown to be a non-problem.

The remainder of this section of PSE&G's response to the NRC's information request will address translation of design bases information into plant configuration and plant procedures.

10 Adq of Design Bases

)(

SALEM GENERATING.STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 RESPONSE TO QUESTION A Description of the engineering design and configuration control processes, including those that implement 10CFR50.59, "Changes, Tests and Experiments,"

10CFR50.71(e) "Maintenance of Records, Making of Reports" and Appendix B to 10CFR50, "Quality Assurance Criteria for Nuclear Power Plants .... ".

I. Overview Public Service Electric and Gas (PSE&G) has many design and configuration control processes, including processes that implement 10CFR50.59, 10CFR50.71(e) and Appendix B to 10CFR50. These processes are integrated into the daily operations and maintenance of the Salem Generating Station.

Changes to plant design and configuration are controlled by use of approved procedures and involve trained and qualified personnel at various levels of the organization. The majority of these procedures are common to both the Hope Creek and Salem Stations.

The processes described below are currently used to maintain plant design and operations consistent with the design bases. These descriptions are summaries of our processes intended to convey a sense of how they work and are not intended to provide procedural details; and the processes are not static and may change in the future. Included in these descriptions are process and organizational interfaces and the elements intended to satisfy 10CFR50.59, 10CFR50.71 (e) and support 10CFR50, Appendix B, requirements.

II. Appendix B to 10CFR50 Appendix B to 10CFR50 establishes quality assurance requirements for the design, construction and operation of those structures, systems and components (SSCs) that prevent or mitigate the consequences of postulated accidents that could cause undue risk to the health and safety of the public.

The Quality Assurance Program (QAP) is an essential part of the PSE&G commitment to safe and reliable operation of the Salem Nuclear Power Plant.

The QAP is described in the Salem Final Safety Analysis Report and meets the requirements of 10 CFR 50 Appendix B. The QAP provides control over independent and combined activities affecting. the quality of identified SS Cs consistent with their importance to safety.

11 . Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 10CFR50, Appendix B, is divided into 18 criteria relative to Quality Assurance which are to be applied to the design, fabrication, construction and testing of the SSCs. PSE&G has implemented processes to meet each of these criteria.

These processes are driven by PSE&G Nuclear Business Unit (NBU) procedures. The design and configuration control processes described below incorporate the requirements of Appendix B to 10CFR50 applicable to those activities.

• Specific criteria from 10CFRSO, Appendix 8, which apply to the design and configuration control processes include: Design Control; Procurement Document Control; Instructions, Procedures and Drawings; Document Control; Test Control; and Quality Assurance Records. These processes are integrated by procedure into the activities related to support the facilities. Other 10CFR50, Appendix 8, requirements related to program implementation, including those related to materials and equipment control and usage, and the Corrective Action Program are discussed in the "Response to Question D" below.

A comprehensive system of planned and periodic audits ensures the continuing effectiveness and adequacy of the design and configuration control processes .

These audits review program compliance with 10CFR50, Appendix 8, as well as effective implementation of the program elements.

Ill. Description of Design and Configuration Change Processes PSE&G has established processes for design and configuration changes to the Salem Station. These processes are intended to ensure that:

1. Design and configuration change activities receive the required analyses, evaluations and reviews,

2. Design information is correctly incorporated into the final design, and

3. Configuration control is maintained during the change activity.

Proper implementation will ensure that design and configuration changes are in compliance with the plant design and licensing bases and that documentation is appropriately reviewed to reflect design and configuration changes.

The processes that govern permanent and temporary changes to plant *design and configuration at the Salem Station are:

12 Question A

SALEM GENERATING. STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 A. Control of Design and Configuration Change, Tests and Experiments (NC.NA-AP.ZZ-0008(0)),

B. Minor Modification Process (NC.NA-AP.ZZ-0017(0)),

C. Control of Temporary Modifications (NC.NA-AP.ZZ-0013(0)), and D. Nuclear Fuel Procurement and Core Control (NC.NA-AP.ZZ-0072(0)).

A brief description of each of the above processes follows.

A. Control of Design and Configuration Change, Tests and Experiments (NC.NA-AP.ZZ-0008(Q))

Procedure NC.NA-AP.ZZ-0008 (0), "Control of Design and Configuration Change, Tests and Experiments," (NAP-8) establishes a uniform method for controlling design and configuration changes, test and experiments. The process is common to the Hope Creek and Salem Stations.

The NAP-8 process has the specific purpose of ensuring that proposed design changes are made in accordance with assumptions in the plant design and licensing bases and that impacted documents (including drawings, calculations, procedures and databases) are identified and revised in an appropriate manner.

Supporting procedures referenced throughout this process are designated to ensure that the appropriate requirements are understood by those developing, implementing and closing out a design or configuration change, test or

• experiment.

Changes controlled by the design and configuration change process fall into the three major categories of design changes, configuration changes and tests and experiments as discussed below.

Design Change A design change is a change to the system design that affects the functional performance or parameters .of plant SSCs.

• Design changes, once implemented, require an update of the fundamental configuration documentation to reflect the revised as-built condition. Design changes are classified as:

1. A "standard design change" which implements a new design for such reasons as improved safety, increased capacity, or to meet new regulatory

• requirements.

13: Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996

2. An "as-built" design change, which updates existing configuration documentation, including the UFSAR, to be consistent with actual plant design following a documented design analysis.

Configuration Change A configuration change is a change to the fundamental configuration documentation that does not affect the design bases functional performance or parameters of plant SSCs, or is typographical in nature. The four types of configuration changes are:

1. The "equivalent replacement" configuration change which involves replacement of parts or components with ones of equivalent functions, or non-design computer software changes. These component replacements may be due to obsolescence, unavailability, or lack of reliability of specific components. This change type documents the equivalency of form, fit and function.

2. The "generic equivalent replacement" configuration change pre-engineers the acceptability of whole components or piece parts as alternative equivalent replacements. Installation is performed under the work control

• process. This change type permits configuration control of databases as each alternative equivalent replacement is installed.

3. The "document only" configuration change is used to correct drafting or editorial errors, correct errors of omission, correct discrepancies between documents and to revise component bill of materials (BOMs) and vendor documents to identify parts number changes due to obsolescence.

4. The "engineering change authorization" configuration change implements changes to SSCs that do not affect the functional performance or parameters described in the design bases.

Tests and Experiments Tests and experiments are not permanent design or configuration changes to the plant, although they may involve installation of temporary equipment for their conduct. A test is a controlled set of plant operations intended to verify that systems or components function in accordance with predetermined specifications. An experiment .is a controlled set of plant operations intended to establish system or component characteristics or values not previously known.

14 Question A

SALEM GENERATING-STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Tests and experiments affecting configuration are controlled and documented by a change package. Design change package control procedures require incorporation of post-modification test procedures to confirm the achievement of the design change intent and identification of any changes to facility surveillance or operating procedures necessitated by the change. If necessary, experiments, e.g., core reload physics parameter determinations, are controlled by operational procedures reviewed and approved in accordance with procedure NC. NA-AP .ZZ-0001 (Q), "Nuclear Procedure System," (NAP-1 ).

Change Package Details Plant configuration control is maintained during implementation of a design or configuration change through issuance of change packages, namely Design Change Packages (DCPs) and Configuration Change Packages (CCPs). The format for change packages is prescribed for each type of change in the applicable Design Engineering procedures. The change package content varies according to the type of change and its scope.

From the perspective of design controls, development of a standard DCP is an important part of the process. Typically, the process will involve consideration of the information discussed below. Other types of change packages, e.g., as-built design change, equivalent replacement configuration change, generic equivalent configuration change, document only configuration change, or engineering change authorization, will contain some or all of the described elements, as required by the applicable Design Engineering procedures as they are applicable to that specific class of change. The format of a standard DCP, which is the most comprehensive class of change, includes:

1. General The Forms Section of the package includes documentation of approvals and revisions, an Interface Record, an Executive Summary, a Station Department Change Package Checklist and a turnover to operations checklist.

The Interface Record documents that the discipline or specialty areas and associated interfaces have been adequately addressed and mandatory station department and installer reviews have been obtained. The Station Department Change Package Checklist identifies affected station procedures or databases and update requirements for these items.

Change packages are subject to the procedurally controlled peer review process. Procedure NC.DE-AP.ZZ-0009(0), "Peer Review," (DEAP-9),

establishes controlled and standard guidelines for performing a peer review of

. 15 .: Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 design documents in accordance with Regulatory Guide 1.64. Peer review is the process of ensuring that each design output document satisfies the specific design inputs and source criteria and it serves to identify and correct deficiencies in the design documents. The review is performed, at a minimum, using a line-by-line check.

2. Engineering The Engineering section includes documentation of the supporting engineering and design analysis. This section includes:

Design Bases/Input (NC.DE-AP.ZZ-0001(Q))

Procedure NC.DE-AP.ZZ-0001 (Q), "Design Bases/Input," (DEAP-1) establishes a method for identifying design considerations and design input used in the preparation of design documents. This procedure is used by the change package preparer to identify the significant design inputs used in the preparation of the design document as required by ANSI N45.2.11, "Quality Assurance Requirements for the Design of Nuclear Power Plants," under our commitment to Regulatory Guide 1.64. The procedure ensures a complete record and an efficient means of retrieval of design document information.

Specialty Review (NC.DE-AP.ZZ-0007(Q))

Procedure NC.DE-AP.ZZ-0007(Q), "Specialty Reviews," (DEAP-7) establishes guidelines for performing a review of design documentation to ensure that effects of the change on specific technical and engineering programs are considered.

The Specialty Review procedure (DEAP-7) uses checklists to determine potential impact of the change on specific technical and engineering programs.

The checklists include fire protection, environmental qualification of electrical equipmenUdevices, pipe stress analysis, seismic qualification and the lnservice Inspection Program. In addition, security, Cold Shutdown From Outside Control Room (GDC-19), lnservice Testing and Valve Programs, Regulatory Guide 1.97 variables (Post Accident Monitoring), environmental design criteria, nuclear fuel design impact, Probabilistic Risk Assessment (PRA), Station Blackout, setpoint calculation and the Erosion/Corrosion Monitoring Program are also included.

The change package preparer determines the specialty reviews that are required for the design change. The programmatic specialist reviews the change package to ensure that applicable programmatic requirements are met and that sufficient information is available to update the applicable program.

16 Question A

SALEM GENERATING.STATION Response to 10 CFR 50.54(f) Letter.

Dated October 9, 1996 Design Analysis As defined in procedure NC.DE-WB.ZZ-0001 (Q), "Standard Design Change Workbook One," the Design Analysis section summarizes methods of analyses and discusses how the change meets the requirements and design impacts identified in the Design Bases/Input section. It includes discussions related to system performance requirements, component design conditions and constraints and how safety requirements are satisfied. The design analysis addresses exceptions or deviations from industry/vendor standards and assumptions.

10CFR50.59 Applicability Review or Safety Evaluation Design Change Packages require screening for 10CFR50.59 applicability as described in procedure NC.NA-AP.ZZ-0059(Q), "10CFR50.59 Reviews and Safety Evaluations," (NAP-59). If the change is within the scope of 10CFR50.59, a safety evaluation is performed prior to implementation of the design or configuration change. The safety evaluation determines if an unreviewed safety question (USQ) exists.

The 10CFR50.59 Applicability Review(s) and Safety Evaluation(s), if required, are included in this section of the change package. The Station Operations Review Committee (SORG) is required to approve the change package if a, 10CFR50.59 Safety Evaluation is required. These reviews are performed in accordance with the NAP-59 process, which is discussed in section VI below.

Design Verification (NC.DE-AP.ZZ-001 O(Q))

Procedure NC.DE-AP.ZZ-001 O(Q), "Desigri Verification," (DEAP-10) establishes guidelines for performing independent design verifications of design documentation to meet the requirements of Regulatory Guide 1.64. Design verification is performed on documentation for Q-listed items and non-Q-listed items that are considered to be important to safety, resulting from Technical Specification (TS) changes, or are changes to previously verified designs.* The design verifier determines the method, extent and depth of verification; performs verification of design documentation, specialty reviews and safety reviews. The design verifier also documents the acceptability of the design documentation.

3. Modification and Testing As defined in procedure NC.DE-WB.ZZ-0001 (Q) "Standard Design Change Workbook One," the Modification and Testing section contains the information Question A

SALEM GENERATING STATION .I Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 required by the installer in the field to complete the change. This section includes: a Modification Details section which provides the requirements to assure proper modification installation; a Testing section which includes operational, functional and special testing requirements; a Modification Documents section which lists the documents impacted by the change that are required to support installation; and the Equipment, Services and Materials section which lists material required to install the modification. If required, a Code ~ob Package is also included in this section.

4. Close-out The Close-out section contains items required to ensure that the change has been completed in accordance with the proposed design and that documents/databases have been updated to maintain current configuration control. This section documents acceptance by Plant Engineering, the System Manager and Station Operations. This section includes: a Materials Disposition section providing disposition for those materials left over at the end of the modification; a Change Document List section which identifies documents impacted by the change that are not required for installation; an Exception List section which identifies work not completed; and a Close-out Check List which

• identifies actions to be completed prior to system restart and prior to change package close-out.

Walkdowns may be performed throughout the various stages of change package development and implementation using procedure NC.DE-AP.ZZ-0003(0),

"Modification Walkdown Program," (DEAP-3). Walkdowns are performed to verify existing configuration, determine causes of planUdocument discrepancies and determine the feasibility, operability, maintainability or testability of a proposed design change.

Change Package Review and Approval The change package is reviewed throughout development, implementation and close-out. Requests for design changes are reviewed by a review board. The problem that the DCP is intended to correct and the proposed design change, are presented to the review board by the System Manager and Design Engineer.

The board then determines if the modification is necessary and whether the proposed change will adequately address the problem.

Reviews from the technical disciplines, specialty areas, mandatory station department and the installer are documented on th*e Interface Record. The design change package requires approval from a PSE&G Engineering Supervisor, SORC review (for change packages requiring 10 CFR 50.59 Safety 18 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Evaluation) and station General Manager approval prior to implementation. The design change packages are peer reviewed and design verified if required.

Documents required for system operation are revised following installation in accordance with the applicable controlled processes. The System Manager accepts the package for the station and confirms the system is ready for turnover to Operations. The Senior Nuclear Shift Supervisor (SNSS) then accepts the system for operability testing. Final close-out of the package occurs when impacted documents and databases have been updated or placed in a tracking system in accordance with their own controlled process.

The design change process interfaces with many processes and site organizations. The key supporting engineering processes include design bases/input, design calculations and analyses, Modification Walkdown Program, design drawings, peer review, Design Verification, Procurement Classification Guidelines, modification concerns and resolution and engineering evaluation ..

Higher tier processes include the Document Management Program, SORG, Work Control Process, Preventive Maintenance Program, Corrective Action Program, Safety Tagging Program, Code Job Package, nuclear licensing and reporting, chemical control and 10CFR50.59 applicability reviews and safety evaluations. The Change Process requires support from Operations, Maintenance, Planning, Radiation Protection, Chemistry, Licensing, Fuels, Quality Assurance, Engineering, Business Support and Nuclear Training.

Training on the NAP-8 change process is required as part of the Engineering, Support Personnel (ESP) Training Job Qualification Guide (JQG) for PSE&G ..

engineers. Staff augmentation contractors and contractors who perform work off-site also are trained either on-site, or as a part of their contract if they are to perform design or design change work.

The NAP-8 procedure receives a biennial review as required by NAP-1. The review includes confirmation of the technical accuracy of the procedure, level of information provided by the procedure for its intended use, legibility of procedure pages, compliance with the procedural hierarchy requirements and compatibility of the procedure with referenced procedures and processes. The procedure may also be revised in response to conditions -adverse* to quality identified in.

accordance with the Corrective Action Program. Revision requests may also be generated by any procedure user. NAP-8 revisions require screening in accordance with the 10CFR50.59 program.

B. Minor Modification Process (NC.NA-AP.ZZ-0017(Q))

• Procedure NC.NA-AP.ZZ-0017(Q), "Minor Modification Process," (NAP-17) describes the process for preparing, reviewing and installing minor modifications 19 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 at the Salem Station. A minor modification is a change that does not impact the safe operation of the station, affect the design bases of any SSC, require a change to the TS, or create an USQ.

The Minor Modification Process may be used provided the following criteria are met:

1. The modification is a result of a deficient condition in the field or is needed to enhance the reliability of performance of an SSC, and

2. The change does not impact the design bases information described in the UFSAR or other design bases documents, and

3. The change requires no complex calculations or engineering design, and

4. Complicated or specialty engineering analysis o.r unique equipment qualification is not required, and

5. The change can be implemented with minor engineering direction.

If a minor modification is appropriate, an engineer is assigned to develop a Minor Modification Package (MMP). The MMP addresses the purpose, description, quality classification, drawings affected and special instructions or conditions for installation and return to service. The MMP contains a Design Inputs Evaluation which documents review of the design inputs. Items considered include consistency of the minor modification with design parameters, environmental and seismic qualification, electrical system loading, ASME Code requirements, performance characteristics of safety-related SSCs, effect on safety actuation systems, separation criteria and common mode failure, fire hazards analysis, personnel injury or equipment damage and radiological concerns. Specialty reviews are performed as necessary in accordance with DEAP-7.

A 10CFR50.59 applicability review is also performed for minor modifications. If the minor modification is within the scope of 10CFR50.59, a safety evaluation is performed in accordance with NAP-59.

An independent design verification review is performed if the minor modification affects Q-listed SSCs, non-safety-related SSCs considered important to safety, or if required by management. The MMP is reviewed by an engineering supervisor and if important to safety, by station Quality Assurance. If the minor modification is within the scope of 10CFR50.59, it is presented to SORC for review and recommendation for approval by the station General Manager.

20 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 C. Temporary Modification Process (NC.NA-AP.ZZ-0013(Q))

A Temporary Modification (T-Mod) is defined as a modification to an operable SSC that temporarily alters the approved designed configuration. This includes lifted leads, jumpers and electrical and mechanical installations. Procedure NC.NA-AP.ZZ-0013(0), "Control of Temporary Modifications," (NAP-13) is common to both the Hope Creek and Salem Stations This procedure ensures that T-Mods are controlled in a manner that ensures operator awareness, conformance with design intent and operability requirements and preservation of plant safety and reliability. Proper use of the NAP-13 process ensures that the

. design process for permanent plant modifications will not be circumvented.

It is intended that temporary changes be minor in scope, of short duration and few in number. Activities not considered to be T-Mods are defined in NAP-13.

Examples of activities not considered to be T-Mods include installation of a pressure gauge on an instrument tap, installation of a hose on a station air/breathing air system hose station, connection of sample tubing and connection of hoses to service outlets specifically designed for that purpose.

• Under conditions defined in NAP-13, temporary modifications may be made without a Temporary Modification Package (TMP). These temporary

• modifications are controlled by approved procedures.

The need for a TMP is reviewed by a System Engineering or Nuclear .

Engineering representative. If a TMP is required, an engineering representative is assigned to develop a TMP which includes the following information:

1. Review and approval signatures.

2. Description of the T-Mod, including purpose, SSC affected, locations, quality classification, 10CFRS0.59 evaluation, affected drawings and procedures, special instructions for installation and removal, action required to remove, and expected removal date.

3. Design input evaluation, which addresses design characteristics such as pressure, temperature, fluid, chemistry, voltage, current, material compatibility, seismic, wind, thermal and dynamic loading, environmental or seismic qualification, electrical system loading, effects on ASME Code Class components, effects on safety-related SSCs, separation concerns and common mode failure potential. . Fire hazards analysis, potential personnel injury and equipment damage resulting from the T-Mod, effect

• • on the operators ability to control or monitor the plant and radiological effects are also considered.

21 Question A

SALEM GENERATING STATION*

Response to 10 CFR 50.54(f) Letter Dated October 9, 1996

4. Record of installation and removal, including procedural controls to assure return to normal configuration and retest requirements.

A 10CFR50.59 Applicability Review is performed for T-Mods. If the T-Mod is within the scope of 10CFR50.59, a safety evaluation is performed and processed in accordance with NAP-59. This review process is completed prior to T-Mod implementation.

• Following TMP development, a review is performed by a Nuclear Engineering Supervisor. SORC review is required if 10CFR50.59 is applicable, the T-Mod affects nuclear safety, or if requested by the engineering representative, supervisor, manager or SORC member. TMPs which require SORC review, also require approval by the station General Manager.

Installation of the T-Mod is authorized by the Senior Nuclear Shift Supervisor (SNSS) or Nuclear Shift Supervisor (NSS). The T-Mod is installed in accordance with the instructions contained in the TMP and an independent installation verification is performed. The SNSS confirms that affected Operational Working Drawings have been updated upon installation of the T-Mod. The engineering representative verifies the applicable procedure revisions have been incorporated. Removal of the T-Mod requires SNSS/NSS authorization.

The Technical Document Room (TOR) notifies copy holders of operational working drawings affected by the installation and removal of T-Mods. The operating shifts review the status of installed T-Mods prior to changing modes and as part of shift routine. Installed T-Mods, if accessible, are reviewed quarterly to ensure that they are still required and that each is properly installed and the tags are correctly in place.

• D. Nuclear Fuel Management Program (NC.NA-AP.ZZ-0072(Q))

Procedure NC.NA:*AP.ZZ-0072(Q), "Nuclear Fuel Program," (NAP-72),. defines an integrated fuel management program for the procurement, use and disposal of nuclear fuel at Salem. The Nuclear Fuels Section of the Nuclear Engineering Department is responsible for developing and maintaining the design bases for fuel and control blades and ensuring that fuel designs loaded into the Salem reactor cores and spent fuel pools conform to the design descriptions and constraints as described in the Salem UFSAR and TS.

• At the start of core reload design activities, design initialization _discussions are held with the fuel vendors to address operational and design bases constraints that could impact core design. The Nuclear Fuels Section designs or oversees 22 Question A

SALEM GENERATING.STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 the core reload design to ensure that required analyses are completed and meet pertinent regulatory requirements and commitments. Fuel fabrication techniques, methods, processes and procedures are reviewed by PSE&G personnel to ensure design intent is met and that fabrication processes are appropriately controlled. Quality Assurance performs technical audits of the fuel vendors periodically, in accordance with this program. Reviews ensure that a Reload Analysis is performed using NRC approved computer codes, methods and procedures.

A 10CFRS0.59 Safety Evaluation is performed for each core reload. In-house reviews are performed to validate the reload design analysis provided by the fuel vendors. For new fuel designs, additional operating margins are designed into the reload analysis. Operating experience with the new design is factored into determining the appropriate design margins to use in future cycles.

To ensure nuclear fuel design compliance, the Nuclear Fuels Group specifies required testing following core alterations, modifications, or instances when the core is not within analyzed design parameters. The technology is maintained to simulate the physical plant for safety, design, performance and transient thermal hydraulic evaluations. The Nuclear Fuels Engineers provide engineering support of station operations pertaining to nuclear fuel management, core performance and nuclear fuel performance.

IV. Description of Configuration Control Processes PSE&G has the following processes in place to maintain plant configuration in accordance with the design bases. These processes are used in day-to-day station operations, maintenance and testing to control plant configuration and interface with the design and configuration change processes described in Section 111 above.

A. Nuclear Procedure System (NC.NA-AP.ZZ-0001(Q))

Procedures govern the conduct of Station Operations, Department Operations, Corrective and Preventive Maintenance, Design and Configuration Changes, Testing, Procurement, Engineering Activities, Corrective Action Program and Oversight Activities. Procedures used at the Salem Station are available to

• station personnel electronically via the Document Management System (OMS).

Working procedures may be reproduced from the OMS for use by workers.

The requirements related to development, revision and use of procedures are contained in NAP-1, which is common to the Hope Creek and Salem Stations.

Each procedure has a sponsor organization responsible for ensuring that new 23 Question A

SALEM GENERATING STATION*

Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 procedures and revisions receive the appropriate levels of review and approval.

New procedures and revisions are subject to the 10CFR50.59 process as described in NAP-59. The 10CFR50.59 review process is the means by which procedure changes are evaluated for impact on the Salem Station UFSAR .

and TS.

Development and revision of upper tier Nuclear Administrative Procedures (NAPs) governing activities within the scope of Regulatory Guide 1.33 (February, 1978), are reviewed by SORC, as required by Section 6 of the TS.

SORC is also responsible for reviewing new procedures and revisions that are within the scope of 10CFR50.59.

As defined in NAP-1, Station Qualified Reviewers (SQRs) perform a

• standardized independent technical review of defined categories of procedures.

The SQR review focuses on the basis for the change and is intended to provide reasonable assurance that the procedure is technically correct and does not compromise nuclear safety. The need for a cross-discipline review is determined during this review. SQRs are required to meet or exceed the requirements of Section 4.4 of ANSI 18.1-1971 for the discipline(s) in which they are qualified, must complete SQR training and maintain an active status through biennial recertification subject to the approval of the SORC Chairman.

Any person on site may recommend procedure changes by submitting a revision request via the Action Request Process. In addition, a revision request maybe initiated as a result of a corrective action for a condition adverse to quality.

Procedure revisions resulting from changes to the TS are identified as part of the license amendment process. Revisions or new procedures resulting from design change packages are initiated and controlled as defined in NAP-8.

The need for Verification and Validation (V&V) of new procedures and revisions is determined by the procedure sponsor organization in accordance with procedure NC.NA-WG.ZZ-0001 (Q), "Procedure Writer's Guide.. " The V&V is performed in as near-to-real circumstances as possible, without actually manipulating installed plant equipment. During the V&V, the criteria applied include adequacy of meeting the intent of the procedure,. conformance of the procedure with plant design and satisfaction of commitments.

Administrative hold is used to place a procedure in the inactive status when it is determined that the procedure does not comply with station, regulatory, or licensing commitments. If an administrative hold is the result of a condition adverse to quality, the condition is addressed in accordance with the Corrective Action Program.

24* Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Biennial reviews are performed by the procedure sponsor organization for active and current procedures subject to the Quality Assurance Program in accordance with NAP-1. In defined cases, normal use of the entire procedure within the biennial review period may satisfy the biennial review requirement.

Training requirements for new procedures and revisions are determined by the procedure sponsor organization. The sponsor organization receives input regarding training requirements from procedure reviewers.

B. Managed Maintenance Information System (ND.DE-TS.ZZ-5409(Q))

The Managed Maintenance Information System (MMIS) is an important tool to assist site personnel in configuration control activities. Several databases reside in the MMIS system related to configuration control, including design and configuration change tracking, resource data, maintenance planning and the Corrective Action Program (NAP-6). The system is used for both the Hope Creek and Salem Stations.

The MMIS contains component nameplate data, maintenance planning, inventory control and purchasing information. The MMIS database is the basic tool to identify and maintain current information on SSCs and identify SSCs that are subject to the QAP.

The MMIS Resource Data Module contains the necessary entry screens, inquiry screens, data files and report definitions to enable site personnel to obtain data related to plant SSCs. Procedure NC.DE-AP.ZZ-0015(0), "MMIS Resource Data Module," (DEAP-15), defines the MMIS and provides directions for the control and use of the MMIS Resource Data Module data base.

Changes to data in the MMIS database are made through the change package process in accordance with NAP-8. Significant data are those data derived directly from design documents and include system, component identification, BOM information, manufacturer identification and model number and component quality classification. Procedures require that data base input be independently verified.

C. Work Control, Safety Tagging and Post Maintenance Testing Programs Procedure NC.NA-AP~ZZ-0009(Q), "Work Control Practices," (NAP-9), provides administrative controls for identifying, planning, scheduling, reviewing,

• • performing, testing and post review of preventative and corrective maintenance.

The work control process is designed to maintain configuration control during 25 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 maintenance activities. It provides for proper planning of work activities, authorizations for removal of SSCs from service, controlled replacement of parts to prevent unauthorized design changes, post-maintenance testing to ensure equipment operability and restoration of SSCs following maintenance activities.

Work Control (NC.NA-AP.ZZ-0009(Q))

Procedure NAP-9 describes the method for controlling work at the Salem Station. Work at the station includes corrective maintenance, preventive maintenance, modifications, testing, experiments, inservice inspections, TS surveillance tests, refueling and nondestructive examination activities.

Work activities are performed using procedures that specify how the activity is to be performed, plant conditions required, methods to be employed, equipment and materials to be used and the sequence of the work activity. Work packages are developed for each maintenance activity and are the official documentation of the work performed. The work package assures that work history is maintained for tracking and trending purposes.

The MMIS system is the primary tool used for the work control process. Work is identified, planned, scheduled and tracked on-line using the MMIS system.

Component history data is stored in MMIS to provide access to records of previous work for planning and trending purposes. When MMIS is not available, manual work requests are used to identify and document work.

Work is identified using the Action Request process. Maintenance activities are prioritized using station specific criteria. These criteria consider whether the activity is corrective or preventative maintenance, the impact of the malfunction on safe plant operation and TS Limiting Conditions for Operation (LCO) requirements.

Guidance is provided for planning maintenance in a manner that minimizes the potential for any compromise to plant safety. The planning and scheduling process considers the possible safety consequences of concurrent or sequential maintenance, testing or operations activities. The work package development process includes review of radiation protection, testing and administrative requirements; required permits; spare parts; engineering support; and protective tagging requirements.

Safety Tagging Program (NC.NA-AP.ZZ-0015(Q))

Protective tagging requirements are determined using procedure NC.NA-AP.ZZ-0015(0), "Safety Tagging Program," (NAP-15). This procedure controls and 26 Question A

SALEM GENERATING STATION*

Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 provides information regarding equipment status to protect personnel and equipment from unexpected energizing, startup or release of stored energy.

This program is applicable to SSCs at both Salem and Hope Creek. Equipment requiring safety tagging is identified based on the work to be performed, equipment operatjonal restrictions, or procedural requirements. Personnel with site access receive specific training related to the safety tagging program.

Station Testing Program (NC.NA-AP.ZZ-0050(Q))

Post-maintenance testing requirements are determined by procedure NC.NA-AP.ZZ-0050(Q), "Station Testing Program," (NAP-50), applicable to both Salem and Hope Creek. Post maintenance testing verifies that the affected SSC is capable of performing its intended function, that the original malfunction, if applicable, has been corrected and that no new or related problems have been created by the maintenance activity. Testing guidelines for electrical and .

controls, mechanical maintenance, valve maintenance and Motor Operated .*.

Valves (MOVs) are provided in Attachments to NAP-50. Other sources of information for determining post maintenance testing requirements include TS Surveillance requirements, procedures, INPO Good Practice, the lnservice Inspection Program, the lnservice Testing Program, the Environmental Qualification Program, the Fire Protection Program, maintenance history, the UFSAR and vendor manuals.

Work Package Review and Approval In accordance with NAP-9, work packages are reviewed by the Work Control Center/Operating Shift prior to commencing the work activity. Considerations for this review include TS requirements, system interactions, safety tagging needs; possible changes in radiological conditions and inadvertent Engineered Safety Feature (ESF) actuations. If needed, the tagging request is reviewed, approved and authorized for implementation. Qualified operators are briefed and any tagging requests are implemented prior to the start of removal of equipment from service.

Permission to commence the work activity is obtained from the SNSS or the NSS. Upon completion of the work activity, work is closed and any tags are removed. When operaj)ility retesting is satisfactorily completed and reviewed by the NSS or SNSS, th.e equipment is returned to service. The work package is reviewed by the Job Supervisor and Planning Group to ensure applicable requirements were met.

- 27 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 D. Engineering Analysis The plant processes governing engineering analyses include the design Calculation and Analyses process, the Dose Analysis and the Engineering Evaluations. These processes are established to ensure that the design bases are considered during routine and modification related engineering activities.

Design Calculation and Analysis (NC.DE-AP.ZZ-0002(Q))

Procedure NC.DE-AP.ZZ-0002(Q), "Design Calculations and Analyses," (DEAP-2), establishes the technical and administrative requirements for development, maintenance and control of design calculations in accordance with Regulatory Guide 1.64. This procedure is common to the Hope Creek and Salem Stations.

The DEAP-2 procedure specifies that calculations include or reference:

1. Purpose or objective,

2. Design inputs or design bases,

3. Computer programs used as input,

4. Assumptions including those requiring future confirmation,

5. Conclusions, and

6. Actions necessary to support the conclusions.

DEAP-2 recommends that calculation preparers use the design bases/input procedure (DEAP-1) for design considerations and design input. The Probabilistic Risk Assessment (PRA) program is considered as a method to assess the potential risk impacts.

Dose Analysis*

The Dose Analysis process is used to maintain dose analyses that are consistent with the design bases. Dose analyses are used to determine plant and offsite radiation dose rates and integrated radiation doses, during normal operation and following design bases accidents. These analyses are performed for design and configuration changes, as necessary.

Dose analyses are documented in design calculations or vendor technical documents. Design calculations are controlled by DEAP-2 and are approved at the supervisor level. Vendor technical documents are controlled in accordance with the Vendor Technical Document Control Program and are also approved at the supervisor level.

.~

28 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 The Specialty Review process ensures that the effects of Change Packages or Temporary Modifications on dose analyses are evaluated. Dose analysis changes are reviewed for 10CFR50.59 applicability and Safety Evaluations are performed when necessary, using the guidance provided in NAP-59. Conditions adverse to quality that involve dose analysis are dispositioned and corrected in accordance with the Corrective Action Program and the Action Request Process.

Engineering Evaluation (NC.DE-AP.ZZ-0026(Q))

Procedure NC.DE-AP.ZZ-0026(0), "Engineering Evaluations," (DEAP-26),

authorizes, defines and controls the use of engineering evaluations. This procedure applies to formalized evaluations performed to document reviews, analyses, conclusions, or recommendations, on topics such as root cause analysis, problem analysis, engineering alternatives, safety concerns, or economic considerations. The engineering evaluation process is common to the Hope Creek and Salem Stations.

Design considerations and design inputs to the engineering evaluation are identified using DEAP-1. Potential impact on specific engineering programs is also evaluated using the Specialty Review procedure. Evaluations that change the design bases for SSCs important to safety or the bases of analyses or conclusions stated in the SAR, receive a 10CFR50.59 applicability review or safety evaluation, if required, in accordance with NAP-59. The preparer is responsible for identifying updates to documents impacted by the engineering evaluation.

Engineering evaluations are approved by the preparers functional supervisor, manager and supervisors or managers responsible for commitments identified in the evaluation.

E. Design Drawings Process (NC.DE-AP.ZZ-0004(Q))

Design drawings are important in ensuring that plant configuration is maintained in accordance with the design bases because they depict or describe the designed SSCs as they actually exist. The design drawing control process, procedure NC.DE-AP.ZZ-0004(0), "Desig*n Drawings," (DEAP-4), provides direction for preparation, review and approval of new design drawings and permanent revisions to design drawings. This procedure applies to both the Hope Creek and Salem Stations.

~.

The Design Drawings procedure (DEAP-4) outlines the process for preparing, reviewing and approving design drawings. Existing drawings are revised, or new drawings are created, in order to reflect changes or additions. Before the 29 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 drawings become part of the approved permanent design record, a change package is developed in accordance with the NAP-8 process. The change package may include an interim revision document. After the change package is implemented, the interim revision drawings are used to update the original permanent drawings to reflect the approved installed condition.

During drawing preparation activities, plant walkdowns and necessary calculations are performed. The walkdowns are performed in accordance with DEAP-3. Calculations are performed in accordance with DEAP-2. The new or revised drawing is then prepared and peer reviewed. Final drawing approval is by a PSE&G supervisor or his designee.

Design drawings are retained as part of the permanent design record and are microfilmed and stored in accordance with the document control and records management program. Drawings are available to site personnel electronically from the Document Management System (OMS).

F. Software Control (NC.DE-AP.ZZ-0052(Q))

Procedure (NC.NA-AP.ZZ-0064(0), "Software and Micro-processor Based Systems (Digital Systems)," defines requirements for computer software control for the Nuclear Business Unit. Procedure NC.DE-AP.ZZ-0052(0), "Software Control," (DEAP-52), provides the controls applicable to critical software used by the Engineering Department to support or perform safety-related functions, or provide data upon which decisions can be solely made regarding plant operation, design or emergency response. This process is common to the Hope Creek and Salem Stations.

The use of software to produce, or assist in producing calculations that are critical to the design of components important to safety can have a direct effect on safety functions in the plant. As a result, software used in these cases is subject to the same OAP controls and requirements as other facets of nuclear plant design, construction and operation.

DEAP-52 provides guidance for evaluation of existing software, development of new software and revision of existing software. Guidance is also provided regarding procurement, validation and verification, turnover, maintenance, voiding and security of critical software. Critical software used to support a change package or temporary modification requires a review for 10CFR50.59 applicability in accordance with NAP-59. Computer errors identified in critical software are evaluated for applicability in accordance with the reporting requirements of 10CFR Part 21.

• 30 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Procedure NC.DE-AP.ZZ-0054(Q), "Process Computer Maintenance and Modification Control Program," (DEAP-54), used by the Digital Systems Group, documents the method by which maintenance and modifications to existing process computer systems and installations of new process computer systems, shall be performed. It describes the methods of test performance, how errors will be documented and corrected and the degree of subsequent testing.

G. Procurement Process (NC.NA-AP.ZZ-0019(Q))

Procedure, NC. NA-AP .ZZ-0019(Q), "Procurement of Materials and Services,"

(NAP-19), details the requirements for procurement of materials and services, inventory control and receipt of materials and equipment to meet the Quality Assurance Program requirements. The process is essential in the configuration control process because it ensures that replacement parts and new equipment installed in the plant meet appropriate technical, design bases and quality requirements. This process is common to the Hope Creek and Salem Stations.

Procurement activities related to design bases maintenance and configuration control include those that ensure:

1. Design bases requirements are translated into purchase specifications,

2. Appropriate quality and technical reviews and approvals are obtained for procurements,

3. Technical and quality assurance requirements for spare parts and materials are met, and

4. Materials subject to the QA Program are evaluated upon receipt.

Requirements for materials and services are incorporated into procurement documents. These requirements are on design bases considerations for the item or service involved. Procedure NC.DE-AP.ZZ-0013(Q), "Control of Purchased Material, Equipment and Services Program," (DEAP-13), defines a method for preparation, issuance and control of material, equipment and Q-listed service specifications.

Procedure ND.QA-AP.ZZ-0015(Q), "Material Evaluation Nonconformances,"

specifies th.at incoming parts.and materials are receipt inspected to ensure that the procurement document requirements are met. Material not conforming with applicable requirements is:

1. Clearly identified and segregated, if possible, to prevent inadvertent use,

2. Documented, 31 ' Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996

3. Evaluated for impact of the non-conformance and reportability under 10CFR21 as required, and

4. Reported to Licensing for reporting to the NRC, if required, under 10CFR Part 21.

Final disposition of the non-conformance can include repair, rework, use-as-is, return to vendor or scrap.

H. Vendor Technical Document Control Program (NC.DE-AP.ZZ-0006(Q))

Procedure NC.DE-AP.ZZ-0006(0), "Vendor Technical Document Control Program," (DEAP-6) is designed to ensure proper review, control and maintenance of vendor technical documents (VTDs) from the time of receipt through final disposition. The purpose is to ensure that only the proper and current VTDs are available for the operation and maintenance of Salem and Hope Creek SSCs. Accurate vendor information is essential in ensuring that plant design bases are maintained during operations, maintenance* and design change activities.

DEAP-6 applies to the processing and control of VTDs which relate to specific SSCs and is applicable at both Hope Creek and Salem Stations. Typical VTDs include drawings, calculations, technical manuals, seismic qualification documents and Environmental Qualification (EQ) documents.

I. Technical Specifjcation Surveillance Program (NC.NA-AP.ZZ-0012(Q))

Procedure NC.NA-AP.ZZ-0012(0), "Technical Specification Surveillance Program," (NAP-12) establishes the requirements pertaining to the scheduling

• and implementation of TS surveillance requirements for the Salem Station. The NAP-12 procedure provides direction for:

1. Maintaining the TS cross reference matrix and the MMIS Recurring Tasks related to surveillance testing activities,

2. Schedulirig, initiating and implementation of routine, non-routine and compensatory surveillance testing activities,

3. Applying the TS 4.0.2 maximum allowable extension .of 25% to the normal surveillance interval,

4. Review and approval of surveillance test results, and 32 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996

5. Extending surveillance test completion beyond the TS due date.

Procedures which define programs that implement the TS surveillance program are prepared and processed in accordance with NAP-1. The Salem Station maintains a TS cross reference matrix that relates each TS surveillance requirement to the implementing document, responsible department, mode requirements, surveillance interval, MMIS recurring task numbers, TS condition .

or event for non-routine surveillance tests and initiating document for non-routine surveillance tests. This cross reference is updated when changes are made to the program resulting from TS Amendments, new or revised implementing procedures and other changes.

Routine surveillance requirements with intervals of greater than seven days are scheduled in MMIS. Recurring tasks associated with MMIS scheduled surveillance tests are maintained, scheduled and performed in accordance with NAP-9. Changes to surveillance testing procedures are reviewed for 10CFR50.59 applicability and safety evaluations are performed, if required, in*

accordance with NAP-59.

V. Document Control Process (NC.NA-AP.ZZ-0003(Q))

Procedure NC.NA-AP.ZZ-0003(0), "Document Management Program," (NAP-3),

which includes both Document Control and Records Management functions, includes controls to assure that required documents are controlled and retained.

The program assures that quality documents, including procedures, vendor ,

technical documents, design change packages, design drawings, calculations:

and other engineering documents, are reviewed and approved by authorized personnel prior to issuance and use and are retained in accordance with regulatory requirements.

The Document Control System (DCS), Design/Configuration Change Management System (DCCMS) and Document Management System (OMS),

identify the status, history and current revision of controlled documents. When changes are made to quality documents, users can verify pending or approved changes in these systems. The OMS includes an electronic system for retrieval of effective and superseded versions of procedures, drawings and documents.

This permits ready verification of the applicable version of documents and tracking of changes made in each revision of a document.

Each site department coordinates with the Document Management Group to establish a Records Type List (RTL) for their department. The RTL includes a list of retention requirements for current and anticipated record types maintained by the Document Management Group. When documents listed in the RTL are 33.: Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 deleted, superseded or completed and required data is provided, documented and authenticated, they are transferred to Records Management. A copy of each completed record is either scanned into the OMS in accordance with NRG Generic Letter 88-18, converted to a microfilm, or retained as a hard copy original.

The Document Control and Records Management Programs interface with several plant programs. These include the Nuclear Procedure System for review, approval and issuance of procedures, the Document Management Program for issuance, distribution and retention of quality documents, the SORG for review of upper tier procedures and review of quality documents when an evaluation under 10CFR50.59 is required, the DCP procedure for review and approval of design changes, the Vendor Information and Processing Program for control of vendor documents, the 10CFR50.59 process for changes to the facility, test or experiments or changes to procedures and the Software Control Process.

The overall responsibility for the Document Control and Records Management Program resides with PSE&G Nuclear Business Support. Department Managers are responsible for ensuring that documents are approved and records are transferred to the Document Management Group, in accordance with interfacing procedures.

VI. 10CFR50.59 Process (NC.NA-AP.ZZ-0059(Q))

10CFR50.59 provides a mechanism for licensees to change the design of the facility as described in the FSAR, revise procedures described in the FSAR and conduct tests and experiments not described in the FSAR, without prior NRG approval. This regulation requires a determination and written documentation providing the basis for concluding that the change or test does not involve an USQ, or a change to the TS. If a USQ or change to the TS is involved, the activity must be modified so that an USQ or TS change does not exist, or NRG approval must be obtained prior to implementing the activity.

The 10CFR50.59 Applicability Reviews and Safety Evaluations Procedure (NAP-59) is common to both the Hope Creek and Salem Stations. This process applies to design and configuration changes, tests or experiments, FSAR changes, procedural changes, temporary modifications, changes to specific technical programs and certain dispositions of degraded and non-conforming conditions as documented by the Corrective Action Program.

As directed by appropriate procedures, proposed activhies are screened to determine if they are within the scope of 10CFR50.59 by performing an 34 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 applicability review. An activity is considered to be a change to the facility, as described in the FSAR, if it affects the design bases, operation, or functions of SSCs as described in the FSAR. If the proposed activity involves an SSC not described in the FSAR that could affect the design or operation of SSCs described in the FSAR, an evaluation under 10CFR50.59 is performed. In determining applicability under 10CFR50.59, the preparer is required to review the proposed activity against the UFSAR, NRC Safety Evaluation Reports (SER) and TS. The applicability review is formally documented.

If the activity is determined to be within the scope of 10CFR50.59, a Safety Evaluation is performed. The purpose of the Safety Evaluation is to determine if the proposed activity involves an USQ or a TS change. Guidance for answering the required 10CFR50.59 questions is provided in a Nuclear Administrative Standard NC.NA-AS.ZZ-0059(Q), "1 OCFR50.59 Program Guidance,"(NAS-59).

The Safety Evaluation undergoes a peer review. It is then reviewed by the SORC and approved by the station General Manager prior to implementation. If the proposed activity involves a USQ or a TS change, NRC approval is obtained prior to implementing the activity. An independent review is performed by the Offsite Safety Review (OSR) staff to verify that the changes implemented under the 10CFR50.59 process did not constitute an USQ.

The responsibility for NAP-59 and required reports rests with the Nuclear Licensing Department. Each site organization performing applicability reviews or safety evaluations is responsible for implementation of the program requirements.

The 10CFR50.59 Program requires that only qualified personnel perform applicability reviews and safety evaluations. PSE&G has established uniform training and qualification requirements for preparers, peer reviewers and approvers. Required training of preparers and reviewers is performed by the Nuclear Training Department.

The 10CFR50.59 process interfaces with site processes and programs that have the ability to change the facility and procedures as described in the UFSAR.

Examples of these processes include the Procedure Control, Corrective Action Program, Design Change, Temporary Modification, Minor Modification, UFSAR and TS Bases Change and Design Calculation and Analysis.

VII. Technical Specification Amendment Process (NC.LR-AP.ZZ-OOOB(Z))

• • Procedure NC.LR-AP.ZZ-OOOB(Z), "Operating License and Technical Specification Change Processes," (LRAP-8), describes Technical Specification 35 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 License Change Request (LCR) development and submittal, approved amendment receipt and implementation, and changes to the TS Bases. The process is common to the Hope Creek and Salem Stations.

• Any Site department may initiate an LCR by submitting a request per procedure NC.NA-AP.ZZ-0035(0), "Nuclear Licensing and Reporting," (NAP-35). UFSAR and TS Bases changes related to the LCR are submitted in this manner as well.

The LCR is developed by the Licensing and Regulation Department staff using guidance contained in LRAP-8. Factors considered during LCR development include supporting evaluations and analyses, similar amendments developed by other utilities, generic guidance such as Improved Standard TS, the Standard Review Plan, the Code of Federal Regulations, the UFSAR and SERs, previous TS amendments and Utility Owners Group information. The LCR is developed using a standard template and writers guide included in LRAP-8. Following review and concurrence by the affected departments, the LCR is reviewed by SORC and the Offsite Safety Review committee. The LCR is approved by a company officer prior to submittal to the NRC.

An Implementation Plan is normally developed by Licensing and .affected site departments after LCR development and just prior to submittal to the NRC. The Implementation Plan considers items including procedure changes, UFSAR changes, design changes, setpoint changes and training. Tracking items are assigned for each identified action. The Implementation Plan is finalized when the approved License Amendment is received. The License Amendment is issued following completion of necessary implementing actions and the approved TS amendment is given to Document Control for distribution. The TS, TS bases and NRC amendment Safety Evaluation are available to site personnel on the OMS system.

VIII. 10 CFR 50.71(e) UFSAR Update Process (NC.LR-AP.ZZ-0013(Z))

10CFR50.71(e) requires that the UFSAR be revised to include the effects of changes made to the facility or procedures as described in the UFSAR, safety evaluations performed by the licensee either in support of license amendments or in support of conclusions that changes did not involve an unreviewed safety question and all analysis of new safety issues performed by or on the behalf of the licensee at Commission request, since the prior revision of the .UFSAR.

Changes to the UFSAR are reported to the NRC either annually or 6 months after each refueling outage, not to exceed 24 months*, in accordance with 10CFR50.71.

.~

36 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Procedure NC.LR-AP.ZZ-0013(Z), "UFSAR Maintenance Process," (LRAP-13),

maintains the accuracy of the Salem UFSAR through regular revisions to incorporate changes identified through the design change process, the Corrective Action Program and other internal processes and reviews. NAP-59 specifically requires that the UFSAR be reviewed and the Safety Evaluation Form provides a space for documenting any UFSAR Change Notices generated.

Conditions adverse to quality affecting the UFSAR are documented and dispositioned in accordance with the Corrective Action Program. Design and configuration changes are reviewed for effects on the UFSAR. A change notice is prepared for required UFSAR changes and is included as a "change document" in the Change Package.

The SORG approval dates are used for tracking the update requirements contained in 10CFR50.71(e). Changes to the plant are reflected in the UFSAR following installation so that the UFSAR reflects the plant "as-built" condition. *'

The UFSAR is provided to. plant personnel electronically on the OMS computer.

Controlled copy distribution of the UFSAR and change notice lists and scanning of the UFSAR and change notice lists onto OMS, is performed through the Document Management Program.

The responsibility for UFSAR maintenance rests with the Licensing and Regulation Department which manages the process, develops UFSAR revisions and approves UFSAR change notices. The Engineering Department is responsible for maintaining an accurate description of systems, structures and components in the UFSAR through the preparation of UFSAR change notices.

The SORG reviews and approves safety evaluations related to UFSAR change notices. The OSR Group reviews related safety evaluations for unreviewed safety questions. The Document Management Group performs controlled copy distribution of the UFSAR and change notice lists and scans these into OMS.

Site organizations are responsible for maintaining the UFSAR description for their organization, processes and procedures and preparation and review of UFSAR change notices related to the organization's area of responsibility.

UFSAR Change Notices are tracked using a OMS database. Approved change notice lists and copies of the UFSAR mark-up books are provided to station personnel via the OMS system. The station SNSS has access to hard copies of the UFSAR mark-up books and approved change notice lists for pending UFSAR changes. * ,

• 3.7. Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter IX.

Dated October 9, 1996 Specific Technical Programs

• Procedure ND.DE-AP.ZZ-0038(Q), "Engineering Programmatic and Technical Standards," (DEAP-38), establishes requirements for preparation, identification, review and approval, issuance and maintenance of Programmatic and Technical Standards. A Programmatic Standard* is a design document developed for a specialized area where a history of licensing compliance needs to be developed and maintained.

The following programs are examples of specialized areas that are governed by Programmatic Standards:

A. Equipment Environmental Qualification Program (DE-PS.ZZ-0002(Q))

The Salem Environmental Qualification (EQ) Program provides assurance that electrical equipment important to safety located in harsh environments will perform its safety function when called upon to do so during normal, anticipated operational occurrences and accident conditions. Equipment important to safety includes safety-related Class 1E electrical equipment, non safety-related *.

electrical equipment whose failure under postulated environmental conditions could prevent the satisfactory accomplishment of required safety functions by safety-related equipment and certain post-accident monitoring equipment.

A description of the EQ program is outlined in UFSAR Section 3.11. In addition, the EQ Programmatic Standard DE-PS.ZZ-0002(Q}, "Environmental Equipment Qualification Program," (DEPS-2) and Administrative Procedure NC.NA-AP.ZZ-0062(Q}, "Environmental Qualification Program," (NAP-62), provide information on the implementation and interface process for the EQ Program.

The Salem EQ Program meets the intent of NRC requirement 10CFRS0.49 and assures the qualification is consistent with Salem's licensing commitments. In accordance with Salem's licensing commitments, the EQ Program was evaluated against Division of Operating Reactor Guidelines for Unit 1 and NUREG 0588 Category II requirements for Unit 2. Details of the differences in regulatory commitments between Unit 1 and Unit 2 are included in the UFSAR and section 7.2 of DEPS-2.

Administrative changes to the NAP-62 program are performed in accordance with NAP-1. Reviews and approvals are required of the EQ Program sponsors, SORC and the respective station General Manager. Design Change Packages

• affecting the EQ Program are subject to the Specialty Review Process.

38 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 The EQ specialty review checklist assures that changes to the EQ Program are captured. Within the EQ Program, programmatic controls including Equipment Qualification Maintenance and Surveillance Instruction Sheets (EQMSIS) provide maintenance requirements for the EQ equipment. Environmental Qualification related maintenance or inspection tasks are uniquely identified in the MMIS system to ensure that the task is implemented by the due date.

Deferral of maintenance tasks for EQ equipment is not permitted without prior evaluation from the EQ Engineer. The EQ Maintenance Engineer performs an independent verification to ensure that the EQ preventive maintenance actions are implemented.

B. Motor Operated Valve Program (NC.DE-PS.ZZ-0033(0))

This program applies to the population of Salem MOVs that perform a specific fluid system design bases function to assure that concerns of NRG Generic Letter 89-1 Oare appropriately addressed. The purpose of this program is to , .

verify MOV operability under operating, abnormal or emergency operating design bases conditions to assure Generic Letter 89-10 issues are appropriately addressed. The program also addresses the maintenance of MOV data and calculations to assure continued MOV operability for the life of the plant.

The MOVs included in the scope of the program are determined in accordance with the screening criteria contained in the MOV Program Programmatic Standard NC.DE-PS.ZZ-0033(Q), "Motor Operated Valve Program," (DEPS-33).

For each valve determined to be within the scope of the program, an operating conditions evaluation is performed. This evaluation determines the appropriate design bases parameters present during valve operation under the worst case operating conditions. Typical documents reviewed to determine the valve operating design bases include the UFSAR, TS, CBDs, normal, abnormal and emergency operating procedures, plant drawings, design calculations and component test data and vendor information. An electrical and mechanical capability review is then performed and available diagnostic test data is reviewed. The results of these reviews are compiled into an MOV Capability Assessment which is a summary of the conclusions developed from the design bases review and documentation of the assumed values, calculations and limitations for each MOV.

DEPS-33 establishes the general guidelines for programmatic valves. Exclusion of valves from the MOV Program is accomplished by use of the change package peer review process and an independent valve specialist review for concurrence. Design verification of changes and calculation verifications are performed using the Design Engineering Procedures.

39 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 The preventative maintenance repetitive tasks process includes valve testing and maintenance. The Corrective Action Program documents and evaluates conditions adverse to quality relative to the MOV Program. Corrective Maintenance is identified through the Action Request process.

The engineering and maintenance organizations have responsibility for implementation of the MOV Program. Responsibilities for the design bases, programmatic development and evaluations reside within the Specialty Engineering Branch of Systems Engineer"ing.

C. Fire Protection Program (NC.DE-PS.ZZ-0001(Q))

The Fire Protection Program as outlined in the Salem UFSAR and in procedure NC.DE-PS.ZZ-0001 (Q), "Fire Protection," (DEPS-1 ), provides a program to identify and control design features unique to fire protection which forms an integral part of the PSE&G Fire Protection Program. The Fire Protection Program includes the identification of fire protection requirements for safe shutdown equipment and systems, fire detection and suppression systems, fire resistive structures, administrative controls and procedures to assure equipment reliability and to minimize fire hazards, training of personnel and maintenance of a dedicated fire brigade. The Fire Protection Program has been unified to serve both Hope Creek and Salem. Although both stations have different license and design bases commitments, the fire protection program is designed to address and accommodate both plants.

The programmatic standards, design change process and the document configuration control processes assure that the fire protection design features are maintained. The level of fire protection review for design changes not involving a fire protection feature is performed in accordance with DEAP-7. If any items on the Fire Protection Specialty Review Check Sheet are identified by the Project Team Member or a Peer Reviewer as impacting fire protection the design change is reviewed to determine any impact on the Fire Protection Program.

A detailed evaluation is performed for activities that could potentially impact the fire protection program. The Fire Protection Program assures compliance with 10CFRS0.48 and 10CFRSO, Appendix R, as well as commitments made to the guidance of Branch Technical Position APCSB-9.5-1. ,.

In accordance with NAP-1, the level of review for fire proteQtion procedure changes includes an originator from the Loss Prevention* Group, (formerly the Site Fire Protection Group) with department review, and a department manager's approval. Changes to fire protection procedures are also subject to a 40 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 10CFR50.59 applicability review. If a safety evaluation is performed, SORG review and the Station General Manager's approval are also required.

D. lnservice Testing & lnservice Inspection Programs (NC.NA-AP.ZZ-0027(Q) and NC.NA-AP.ZZ-0070(Q))

Procedures NC.NA-AP.ZZ-0070(0), "lnservice Testing," and NC.NA-AP.ZZ-0075(0), "Valve Programs," apply to lnservice Testing (IST) of Nuclear Class 1, 2 and 3 pumps and valves required to perform a specific function in shutting down the reactor to the cold shutdown condition or in mitigating the consequences of an accident. Procedure NC.NA-AP.ZZ-0027(0), "lnservice Inspection Program," applies to Nuclear Class 1, 2 and 3 pressure retaining components and their supports. The requirements for these programs are defined in 10CFR50.55a(g) and the Salem UFSAR.

The IST and ISi Programs are subject to PSE&G management review and reviews by the Authorized Nuclear Inspector. The IST and ISi Programs interface with the Design Control Process, the 10CFR50.59 Process, the Work Control Process and the Corrective Action Process. The specialty review process requires that modifications proposed to Nuclear Class 1, 2 and 3 components and their supports will be processed through the Specialty Engineering (ISi or IST Group, as applicable). Also current regulations require updating of the ISi and IST Programs once every ten year interval in accordance with 10CFR50.55a.

The IST Group develops and maintains the IST Program. Tests required by the program are performed by the Operations Department, Maintenance Department and the lnservice Inspection and Test Group. Data evaluation is performed as required by the IST Program and results are trended.

The ISi Group develops, maintains and implements the required examinations in accordance with the ISi Program Long Term Plan. To maintain the ISi Program accurate and updated, plant design changes are reviewed for ISi applicability in accordance with the specialty review process. To maintain control of repairs and replacements under ASME Section XI, Code Job Packages and their revisions are routed to Specialty Engineering (ISi Group) for review and acceptance. The Code Job Packages control applicable post maintenance testing and preservice Inspections.

Paragraph 10CFR50.55a(3) provides guidance for submitting proposed

• alternatives (relief requests) to the ASME Code. The need for an ASME Code relief request is identified by the IST or ISi Group and developed and processed 41 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 by the Licensing and Regulation Department. Typically, the relief request consists of a cover letter and attachment(s) which demonstrate that proposed alternative testing or examination would provide an acceptable level of quality and safety, or compliance with the specified requirement of the ASME Code would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety. NRC approval of the relief request must be obtained prior to implementation.

X.

• Independent Oversight A. Station Operations Review Committee (NC.NA-AP.ZZ-0004(Q))

The Station Operations Review Committee (SORC) advises the station General Manager on matters related to nuclear safety. As defined in procedure NC.NA-AP.ZZ-0004(0), "Station Operations Review Committee," (NAP-4). SORC reviews plant operations, modifications, procedures, tests, temporary modifications, 10CFR50.59 Safety Evaluations, unit trip reports, and other issues as delineated in TS 6.5.1. The TS review requirements are reflected in NAP-4. This procedure is common for both Salem and Hope Creek.

SORC reviews specifically defined issues regarding changes to the design bases, licensing bases and accident analyses in Chapter 15 of the UFSAR. In reviewing safety evaluations required by 10CFR50.59, SORC determines if the conclusion whether an USQ is involved was properly made. The SORC review also considers the radiological safety effect on personnel and any effect on safety-related systems/components the change may cause. SORC ensures that the preparer of the change has outlined the relationship of the proposal to both safety related and important to safety systems, structures and components and all issues raised have been addressed. Issues identified by SORC are documented and tracked by the Corrective Action Process.

SORC is comprised of members from station staffs who act as members and alternates as approved by the General Manager. They must meet the minimum qualifications listed in ANSI N18.1-1971 for the Salem Station. The membership of SORC is identified by position title in the Technical Specifications. The qualifications of each individual comprising SORC require documentation. Also, management expectations have been established for SORC members and presenters.

B. Off-Site Safety Review (ND.SN-AP.~-0001(Q))

The Off-Site Safety Review (OSR) staff performs independent review activities consistent with the requirements of TS in the areas of Operations, Engineering, 42 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Chemistry, Metallurgy, Instrumentation and Controls, Radiological Safety, Quality Assurance, Non-destructive Testing, and Emergency Preparedness.

Documented reviews are performed in accordance with procedure ND.SN-AP.ZZ-0001 (Q}, "Independent Safety Review Program." Areas reviewed include 10CFR50.59 safety evaluations, proposed changes to procedures or equipment and tests/experiments that involve an USQ, proposed changes to the facility Operating License and the TS, NRC violation responses and events warranting a review to verify that root causes and safety significance have been determined and that proposed corrective actions appear effective and have or will be taken. This procedure is common for both Salem and Hope Creek. The results of reviews are reported to the SORC.

C. On-Site Safety Review (ND.SN-AP.ZZ-0001(Q))

The On-Site Safety Review Group (SRG) staff performs reviews in accordance with procedure ND.SN-AP.ZZ-0001 (Q), "Independent Safety Review Program;"

of selected plant operating characteristics, NRC requirements and guidance, industry advisories and other appropriate sources of plant design and operating experience information that may indicate areas for improving plant safety. Also reviewed are selected facility features, equipment and systems, as well as procedures and plant activities including maintenance, modifications, operational problems and operational analysis. This procedure is common for both Salem and Hope Creek. The SRG staff members function as voting members of the SORC.

D. Quality Assurance I Quality Assurance (QA) performs planned and periodic audits and surveillances of the ongoing design change and configuration process. QA also samples design and configuration changes to ensure Quality Assurance requirements such as inspection and test requirements, acceptance requirements, and test results documentation are incorporated into the changes. Specifications for Q-listed materials, equipment and services are reviewed and approved by QA.

XI. Other Processes With the Potential to Affect Configuration Control In addition to the processes discussed above, there are other operational practices by which traditional configuration control methods might be circumvented; however, appropriate procedural controls are applied to assure that the design bases is adequately considered and protected. These practices include such activities as Technical Specification Interpretations, Operator Aids, 43 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 and procedures for removing and restoring systems to service following maintenance.

A. Technical Specification Interpretations (SC.OP-DD.ZZ-AD45(Z))

In cases where TS requirements are considered ambiguous, formally documented interpretations may be generated to provide management clarification of the intent. Procedure SC.OP-DD.ZZ-AD45(Z), Technical Specification Interpretation Program" (historical), defined controls for use at Salem.

When generated, these interpretations were reviewed and approved for consistency with the design bases. The interpretations were documented, reviewed for applicability of 10CFRS0.59, reviewed by SORG and approved by station management. In addition, the interpretations were periodically reviewed for continued applicability and need. Technical Specification interpretations did not result in changes to the design bases for the facility.

Technical Specification Interpretations will be phased out at Salem by incorporating their substance into operating procedures; therefore the historical procedure, AD45, may not be utilized in the future.

8. Department Night Orders (NC.NA-AP.ZZ-0005(Q))

Night Orders as defined by procedure NC.NA-AP.ZZ-OOOS(Q), "Station Operating Practices," are a mechanism for issuing management instructions to Operations Department personnel. Such instructions may encompass daily schedule matters, special events or activities, housekeeping matters, non-routine plotting of process parameters, notices about personnel actions, requirements to read certain publications or documents, or similar matters. The procedure specifies that Night Orders should neither conflict with existing procedures, nor should they be used in place of procedures.

C. Control of Operator Aids (NC.NA-AP.ZZ-0044(Q))

Operator Aids as defined by procedure NC.NA-AP.ZZ-0044(0), "Stat.ion Aids and Labels," are a category of locally mounted information, including sketches, notes, graphs, instructions, drawings or other documents. Aids are used to provide warnings to plant personnel performing work or testing in accordance with approved procedures, or to provide information, e.g., component name, or number to assist station personnel. They are not intended to be used to provide instructions for maintenance or testing. Operator Aids are controlled by each operating department for aids applicable to their department. Each department 44 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996

• maintains a log of each Aid initiated by that department, controls the posting of the Aid, periodically reviews the continued need for the Aid and assures that it is removed when no longer necessary. Operator Aids are hot used to change the configuration or affect the plant design bases.

D. On-the-Spot Changes (NC.NA-AP.ZZ-0001(Q))

On-the-spot changes may be made to previously approved procedures, as authorized by procedure NAP-1, to permit completion of the activity prescribed by the procedure as long as the intent of the procedure is not changed and the change does not alter a commitment. These changes typically involve correction of editorial or typographical errors identified during procedure execution. Procedure NAP-1 requires that the procedure change be documented and reviewed by the job supervisor or department management and approved by the SNSS or NSS before performance. Subsequent review and .

approval of on-the-spot changes is required to assure 10CFR 50.59 review and to provide for permanent incorporation into the procedure. Copies of procedures including on-the-spot changes are incorporated into the document control system pending document revision.

E. Equipment Control Program (SC.OP-AP.ZZ-0006(Q) and SC.OP-AP.ZZ-0108(Q))

Operability of SSCs is determined and tracked by the Operations department for Salem by procedures SC.OP-AP.ZZ-0006(0), "Operability Determination," and SC.OP-AP.ZZ-01.0B(Q), "Removal/Return of Nuclear Safety Equipment and .

Surveillance Completion Tracking." Equipment declared inoperable is evaluated against the action statements of the ts and the design bases for the SSC and necessary action taken. If appropriate, an Action Request is processed for repair, modification, or replacement of the inoperable equipment. Inoperable equipment is formally tracked along with the necessary compensatory actions prescribed by the TS action statements.

XII. Availability of Design Bases Documentation In Section V. above, the Document Control Process is c;iescribed. This process assures that documentation is available for use in evaluating potential changes to the facilities design bases. Procedure NAP-3 contains an overview of the document management system for the NBU. Record systems which PSE&G utilizes for retention of records include provision for easy retrieval. The primary systems for record storage and retrieval include the Document Management System, Microfilm Record Files and the Managed Maintenance Information System.

• 45 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 Document Management System: The OMS is a computerized system for retention and retrieval of current configuration documents, including drawings, licensing manuals such as Technical Specifications and UFSAR, vendor documents, engineering documents such as calculations, drawings and design changes (change documents and modification documents); Configuration Baseline Documentation manuals; administrative and operating procedures; correspondence; and other similar classes of documentation. These classes of documentation must be widely available and controlled since several classes of these records are subject to revision. In addition to current configuration documentation, limited historical documentation is also retained for reference in this system. Access terminals and printers are strategically located in the plant and in the Nuclear Administrative Building (NAB), permitting easy verification of revision status and retrieval of working documents.

Microfilm Records: Historical documentation, including design change records, maintenance work orders, documentation furnished by architect engineers and vendors, etc., is filmed and made available from microfilm libraries located throughout the NBU, including the Technical Document Rooms for both Salem and Hope Creek and the NAB. Libraries are equipped with film readers with reproduction capability. The microfilm libraries include records which originated before the OMS system was fully operational.

Managed Maintenance Information System: A computerized system is used on the central mainframe computer to control, track and permit retrieval of equipment configuration information, maintenance work orders, Action Requests (ARs), and recurring maintenance work order tasks. The Document Control System portion of MMIS is also used to verify the current revision and revision status of documents. The MMIS system is accessible through the facility computer network through use of individual personal computers available to plant staff who have access needs.

To assure ready availability of these records to plant personnel, PSE&G has provided Technical Document Rooms, and a record system for licensing manuals and correspondence.

• Technical Document Rooms: Salem, Hope Creek, and the NAB have reference rooms which retain copies of documentation related to ongoing activities, a microfilm library, and copies of applicable reference manuals.

Licensing Files: Copies of correspondence related to requests for license change, responses to NRC reports and other submissions to the NRC are .

retained in the NAB and in a computerized Licensing Data Base, which is 46 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 available for access through the full text search option of the local area computer network. The full text search has the capability to search a large computer based file of documents for specified words or phrases. A library of correspondence between PSE&G and regulatory agencies, along with reference materials is also maintained in the NAB.

The above systems assure the availability of reference documentation necessary for those who prepare, review or approve design change packages, minor modification packages, procedures and other processes that could impact the design bases for the Salem and Hope Creek facilities. These processes are also available for those personnel with operational, maintenance, planning and licensing responsibilities.

During two external reviews in 1995 conducted by Ogden Environmental and Energy Services Company, it was concluded that "the design bases was readily retrievable (largely due to the OMS system, which facilitates easy access to design documents)," and "Document control and retrieval was excellent."

XIII. Personnel Performance Improvement The above sections have discussed the various NBU programmatic and procedural systems and controls which contribute to maintenance of the design bases. This section discusses initiatives taken by PSE&G to provide training and documented qualifications for those staff who utilize these systems. The training has the potential to improve awareness and decrease the potential for compromising the integrity of the design bases. An increase in the level of awareness of personnel performance, as it relates to the design bases has been initiated by a strong policy statement . These NBU expectations were expressed by the Senior Vice President--Nuclear Operations regarding conformance with the design bases. Additionally, the training and qualification programs for engineering support personnel have been upgraded.

On July 25, 1996, the Senior Vice President--Nuclear Operations, issued to licensed operators, with distribution to other managers in the NBU, a letter entitled "Expectations for the Knowledge Level of the Licensing and Design Bases." The purpose of that memorandum was to communicate his expectations regarding understanding and applying licensing and design bases information during the performance of licensed activities.

Also, procedure NC.NP-PO.ZZ-0012(0), "Training, Qualification and, Certification," (NPP0-12), was revised to reflect the new cultural behavior. This

• procedure provides guidance regarding the performance of training, qualification activities, documentation of training, qualification and certification to satisfy Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 regulatory requirements, industry standards, job requirements and other competencies. The implementing procedures apply to the NBU organizations.

One of these implementing procedures, NC.TQ-TC.ZZ-0905(Z), "Engineering Support Personnel Training Program Description," (TP-0905), provides training requirements for initial and continuing training for Engineering Support Personnel (ESP). The program is designed to train PSE&G engineers, engineering supervisors and staff augmentation contractors. These include:

Reactor Engineers, System and Maintenance Engineers, In-service Inspection and In-service Testing Engineers, Design Engineers, Plant Engineers, Licensing Engineers, Nuclear Safety Review Engineers, Nuclear Fuel Engineers, Procurement Engineers, Digital Systems Engineers, and Quality Assurance Engineers. An additional training module has been added to those defined in TP-0905 entitled "Design Basis Fundamentals." This training is currently being given to ESP in the above categories. The scope of the course includes definition of the plant design bases, identification of documents which constitute the design bases, demonstration of techniques to access design bases documents, discussion of regulatory requirements and examples of previous failures to control design configuration.

The Nuclear Design Engineering Department has also recently initiated a Design Engineering Review Board to evaluate the ongoing competency of engineers. This process utilizes selected Licensing and Engineering managers and supervisors, sitting as a review board, to interview and orally examine the knowledge and competence of individual engineers. The oral examination includes questioning on various aspects of the design bases, assessing design bases information in support of the answers to the questions, and the procedures which are designed to assure their maintenance.

Continuing maintenance of the design bases for the station is expected to result from effective implementation of these personnel training and quafification programs, coupled with the existing procedural and programmatic programs.

XIV. Summary The above administrative and procedural controls are intended to prevent unreviewed or unauthorized alteration of the design bases for the plant. These controls provide reasonable assurance that facility operations are consistent with the TS and their bases, as well as the design bases described in the UFSAR and other design documents. Provisions have been defined to permit emergency changes to these procedures and controls when operational conditions require, or errors are discovered; however, procedures for review, approval and implementation include controls to assure that these changes are 48 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 identified, reviewed and subsequently incorporated into the approved design packages, procedures and training programs.

PSE&G is confident that these procedures are adequate in scope to address 10CFR50, Appendix B, 10CFR50.59 and 10CFR50.71(e) and they provide adequate control over the design and configuration. Should implementation problems or deficiencies in the program be detected in the future, they will be reported to the Commission as warranted in accord with 10CFR50.72 or 50. 73.

Appropriate corrective actions will also be taken .

• 49 Question A

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 RESPONSE TO QUESTION B Rationale for concluding that design bases requirements are translated into operating, maintenance and testing procedures.

I. Overview This "Response to Question B" section presented herein provides the rationale for our reasonable assurance that the design bases requirements are appropriately translated into the operating, maintenance and testing procedures.

The primary engineering design and configuration control processes are discussed in the "Response to Question A" section. Key attributes of the "Response to Question A" section processes applicable to the translation of design bases requirements into operating, maintenance and testing procedures are as follows:

• that implementing procedures governing operations, maintenance and testing activities are controlled and readily available for plant uses;

• that implementing procedures accurately reflect design requirements; and

• that overarching processes (e.g., processes that control changes to design requirements, plant design, maintenance and operations activities) are defined and appropriately implemented.

• The fidelity of the Salem design bases was described above under the heading of "Adequacy of Design Bases Information," of this response. Availability of procedures is described in the "Response to Question A" section. The "Response to Question B" section will concentrate on the translation of the design bases into plant procedures. Where appropriate, the discussions identify known weaknesses and detail actions taken and/or planned to resolve these matters.

The "Process Controls" section of this response discusses key elements of the processes described in the "Response to Question A" section as they affect the relationship between the design bases and operations, maintenance and testing procedures. The "Validation Reviews" section *of this response describes actions to provide reasonable assurance that procedures are consistent with the plant design bases and identify potential process issues or improvements. The "Assessment Of Effectiveness" section provides an assessment of the success with which the process controls assure proper translation of the design bases into operating, maintenance and testing procedures, with consideration given to the validation reviews, independent assessment findings and recent process 50 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 improvements. The "Conclusion" section provides the overall rationale for the SGS "Response to Question B."

II. Process Controls The "Response to Question A" section within this document provides details of the processes which may affect operating, maintenance or testing procedures, and the processes by which procedures are controlled. The "Response to Question D" section within this document describes the corrective action program and provides an assessment of its effectiveness. The corrective action program establishes the means by which conditions adverse to quality, including those related to procedures, are identified, evaluated and corrected. The "Response to Question B" section presented herein, describes the processes as they relate to the translation of design bases requirements into operating, testing and maintenance procedures.

The operating license for the SGS units is based on the Final Safety Analysis Report which contains design bases information for the SGS units. As part of the original licensing of SGS, documented in the NRC Safety Evaluatic:m Report for the SGS dated October 11, 1974, plant procedures and their administrative controls were reviewed and found to be in conformance with Atomic Energy Commission (AEC) Regulatory Guide 1.33, "Standard for Administrative Controls for Nuclear Power Plants." However, given the age of the original procedures and licensing reviews, more recent activities are taken as the starting point for the SGS response to question (b ).

The SGS Unit 1 and 2 Technical Specifications (TS) comprise Appendix A to the respective unit's operating license. The TS contain safety limits and limiting conditions of operation (LCOs). The TS action statements prescribe the measures to be taken in the event the LCOs are not met, in order to maintain the plant in safe condition. The TS prescribe periodic surveillance requirements to demonstrate the operability of structures, systems and components (SSCs) subject to the LCOs.

Implementation of TS limiting conditions of operation (LCOs) and surveillance test requirements is a critical aspect of maintaining plant operations, maintenance and testing activities consistent with the design bases. The Work Control Process is used to schedule periodic TS surveillances with a nominal scheduled interval of one week or greater. For non-routine TS surveillances (i.e., those which are initiated by a specific condition or event) and periodic TS surveillances with a nominal interval less than one week, the activities may be tracked to completion using Action Requests (ARs) within the Managed

• Maintenance Information System (MMIS) and/or logs and checklists used by the responsible implementing departments (e.g., Operations, Maintenance). The 51*.* Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 license amendment process includes the development of implementation plans to identify the procedures affected by proposed amendments, and coordination of the necessary changes upon NRG approval and implementation.

The design and configuration change processes can have a significant impact on station procedures. Design Change Packages (DCPs) maintain configuration

• control over procedures affected by the change. DCPs include testing requirements necessary to verify correct installation, conformance with the design change and proper operation. The close-out section of the change package includes the identification and tracking of procedure revisions requiring implementation prior to returning the SSCs to service, including surveillance testing, maintenance or system operating procedures. Controls placed ori the affected procedures via the change package ensure that configuration control is maintained during and after change package implementation.

Temporary modifications (T-mods) may also affect operating, maintenance and testing procedures. The engineering sponsor for the T-mod coordinates the applicable procedure revisions and verifies the procedure revisions are incorporated to reflect the configuration of the affected SSCs.

The work control and safety tagging processes are necessary to maintain plant

.configuration consistent with the design bases during the performance of work on plant SSCs. Work activities associated with modifications, maintenance, testing and inspections are planned, reviewed, scheduled and implemented such that plant safety is maintained. Removal of equipment from service is performed with consideration given to TS req~irements and potential adverse consequences of rendering the affected equipment unavailable. Operability retest and/or post-maintenance .test requirements are specified and performed as necessary to return the SSCs to service.

The UFSAR update process implements 10CFR50.71(e), which requires that the UFSAR accurately reflect the effects of changes to the facility and procedures.

The 10CFR50.59 process is an integral part of the UFSAR update process, in that 10CFR50.59 provides the mechanism for evaluating proposed activities against the design bases in the Safety Analysis Report (SAR), as defined in administrative procedure NC.NA~AP.ZZ-0059(0), "1 OCFR50.59 Applicability Reviews and Safety Evaluations." The UFSAR is an integral part of the SAR.

The Nuclear Procedure Sys.tern establishes requirements for procedure revision and use as described in the "Response to Question A".section. Regardless of the source of a change to a procedure (e.g., design change package (DCP),

license amendment, stand-alone procedure change to facilitate a plant activity),

the procedure is subject to the procedure control process, which invokes the requirements of fOCFR50.59 .. The UFSAR update process, 10CFR50.59 implementation and the Nuclear Procedure System are interdependent 52 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 processes which collectively ensure the translation of the plant design bases into operating, maintenance and testing procedures.

Inconsistencies between procedures and the design bases may be identified during routine activities (e.g., preparation of a 10CFR50.59 evaluation for a design change), by internal Verification and Validation (V&V) activities, or by independent assessments. In any case, the corrective action program (described in the "Response to Question D" section) provides a mechanism for the identification, evaluation, trending and correction of conditions adverse to quality, including those related to operating, maintenance and testing procedures. In addition to identifying and correcting individual discrepancies between the design bases and procedures, the corrective action program enables PSE&G to address potential programmatic deficiencies through the use of trending, root cause analysis and monitoring the effectiveness of corrective actions.

Ill. Validation Reviews This section describes various review activities undertaken by PSE&G which include validation of elements of the plant design bases against operating, maintenance and testing procedures.

A. Procedure Upgrade Project (PUP)

The Salem Procedure Upgrade Project (PUP) was initiated in 1989, in response to concerns raised by the NRC and the Institute of Nuclear Power Operations (INPO) related to the overall quality of Salem implementing procedures. The project scope included implementing procedures used by Operations (except emergency operating procedures), Maintenance and Chemistry departments, as well as a limited number of Reactor Engineering procedures. The PUP resulted in over 3400 procedures upgraded to current format, content, technical adequacy and human factors standards, and included reviews of the procedures against design bases requirements.

The technical bases for procedures were researched by reviewing sources of design bases information, including the UFSAR, design drawings and vendor documents. Technical guidelines were used to provide consistent guidance to the PUP team regarding issues with generic impact on procedures, resulting from corrective actions associated with assessments, audits and inspections.

The procedures issued by the PUP were subjected to 10CFR50.59 applicability reviews to determine the need for a full 10CFR50.59 evaluation .

• Key elements of the PUP were solicitation of user input to achieve procedure usability and independent technical reviews to address consistency of the 53 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 revised procedures with the plant design bases. Station Qualified Reviewers (SQRs) were dedicated to the PUP to achieve consistency in the approach to performing the independent technical reviews. Verification and Validation (V&V) was performed by the implementing departments, to determine technical accuracy and usability. The PUP established procedure file configuration controls to facilitate the issuance of controlled procedure revisions. Upon completion of the project in 1993, the PUP transferred responsibility for procedure maintenance to the station line organization, subject to the administrative controls (e.g., 10CFR50.59, SQR reviews) necessary to maintain procedures consistent with the design bases.

The PUP was the first major effort at SGS to validate implementing procedures against elements of the plant design bases subsequent to NRC issuance of the plant operating licenses. The PUP helps provide reasonable assurance that the design bases are effectively translated into operating, maintenance and testing procedures.

B. Emergency Operating Procedure (EOP) Upgrade Project Based on recognized weaknesses in the SGS EOPs, the Salem Operations department established a project team in January, 1996, to achieve consistency of the EOPs with industry standards and validate the ability of the operators to use them effectively in response to potential plant emergencies. Actions taken to improve the overall quality of the EOPs include the elimination of flowchart logic errors and revisions to address conformance with the generic Emergency Response Guidelines (ERGs). EOP basis documents were revised to reflect the comparison with the ERGs, including justification for Salem-specific deviations from the ERGs.

EOP setpoints were identified. Setpoints were calculated by Westinghouse with channel accuracy data provided by PSE&G. The design bases for setpoints used in EOPs were reviewed and documented in a controlled EOP setpoint document. Setpoint and numerical value data were added to the EOP basis documents.

Verification and Validation (V&V) of the EOPs was accomplished via licensed operator training and a review by a multi-discipline validation team. Four experienced operating crews exercised EOPs during the validation period.

Operating crews provided feedback during EOP simulator training. Licensed operator restart training was successfully completed using the upgraded EOPs.

Technical review and validation was performed by Westinghouse ERG specialists, with the involvement of a SGS-licensed SRO, station quality assurance (QA), a human factors expert and the EOP Upgrade Project manager.

Technical verification evaluation criteria included conformance of the EOPs to

• 54 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter

• Dated October 9, 1996 setpoints and numerical data, and compatibility of the EOPs with plant configuration. A Westinghouse V&V letter for the EOP Upgrade Project indicates the project has met or exceeded industry standards.

The EOP Upgrade Project helps to provide reasonable assurance that SGS operating procedures are consistent with the plant design bases (i.e., EOP consistency with design bases parameters such as setpoints).

C. Technical Specification Surveillance Improvement Project Unit 2 Licensee Event Report (LER) 311195-008, dated January 15, 1996, reported a condition in which a TS surveillance requirement for the containment purge and pressure vacuum relief manual initiation signal was not fully proceduralized. Investigations performed as part of the LER development identified several prior instances of ineffective TS surveillance implementation, including one additional case whereby a requirement was not adequately proceduralized. Therefore, in addition to the corrective actions specific to the identified condition, the LER included a commitment to implement the Technical Specification Surveillance Improvement Program (TSSIP).

The TSSIP involves a review of the surveillance testing program to enhance administrative controls, assure proper scheduling and tracking of surveillances, and validate the implementing procedures for TS Surveillance Requirements.

TSSIP is being performed in two phases. Phase 1, which will be completed prior to the restart of each Salem unit from its current outage, involves a general review of the TS surveillance requirements and implementing procedures, along with an evaluation of scheduling controls. This review involves a comparison of the TS surveillance requirements to their implementing procedures, in order to validate that the purpose statement of the implementing procedure accurately reflects and references the surveillance requirements, and that the stated acceptance criteria are consistent with the associated TS requirements. The scheduling review involved a comparison of surveillance-related Managed Maintenance Information System (MMIS) Recurring Tasks (RTs) against the TS requirements, and a detailed evaluation of operating mode transition procedures. Conditional, or event-driven, surveillance requirements were identified as part of the Phase 1 review. An AR was initiated to ensure conditional surveillance requirements have a formalized mechanism for

• recognizing the condition and initiating the appropriate response. Phase 1 TSSIP implementation provides assurance that surveillance requirements are scheduled at the appropriate periodicity and performed in the correct modes or conditions .

• TSSIP Phase 2, which is currently scheduled for completion by the end of 1997, is a detailed technical review of surveillance procedures (except inservice 55 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 testing and inspection procedures, and snubber inspection procedures, which are controlled under separate programs), development of a basis for compliance with the associated surveillance requirements, and evaluation of the methodology to ensure continued technical accuracy of TS implementing procedures following TSSIP.

The Phase 1 TSSIP review identified over 1,000 Unit 2 Technical Specification Surveillance Requirements. This resulted in the review of over 900 Unit 2 and common Technical Specification implementing procedures. The TSSIP has also reviewed over 1300 Unit 2 MMIS recurring tasks, as well as departmental mode transition procedures and conditional surveillance trigger mechanisms. As of December 1996, the TSSIP reviews resulted in approximately 55 ARs for potential conditions adverse to quality, and approximately 620 revision requests for procedure enhancements. The resulting process improvements are summarized below in Subsection IV.C of the "Assessment of Effectiveness" section.

Completion of the TSSIP Phase 1 reviews and process improvements prior to unit restart provides reasonable assurance that TS Surveillance Requirements will be consistently implemented by their associated surveillance procedures.

Completion of Phase 2 TSSIP will ensure identification and correction of potential design bases inconsistencies with implementing procedures. The completion schedule for TSSIP Phase 2 is the subject of discussion with NRC as part of the SGS restart action plans.

D. Salem System Readiness Review The Salem System Readiness Review Program (SRRP) was developed as part of the System Engineering Restart Action Plan to provide a process for evaluating plant systems' readiness to support long term safe and reliable operation. Forty-six systems were selected for evaluation under the SRRP based on the systems' impact on overall plant safety, high historical corrective maintenance, and impact on generation output.

The SRRP used a four phase review process to evaluate each of the forty-six systems.

Phase I: Initial System Readiness Review (ISRR)

Phase II: Restart Activities Monitoring (RAM)

Phase Ill: Final System Readiness Review (FSRR)

Phase IV: Startup and Power Ascension (SPA) 56- Question 8

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Phase I of the SRRP involved a general review of sources of system design bases documentation (e.g. Updated Final Safety Analysis Report (UFSAR), TS, Configuration Baseline Documents (CBD)), evaluation of open action items against restart screening criteria, and performance of a multi-disciplined system walkdown. Items that were identified as restart-required per the restart screening criteria include, but are not limited to, proposed actions to resolve safety, operability and design bases issues. The evaluation of open items against the restart screening criteria determined the work required for SGS

• restart and power ascension, and the post-restart work scope. Phase I was completed for both SGS units in October, 1995.

The open action items were captured in the Systems Indexed Database (SIDS).

SIDS allowed the System Managers to document their evaluation of the open actions associated with their respective systems in a central database. SIDS contains a compilation of design change packages, operability determinations, preventive maintenance, procedure revision requests, deficiency reports, station workarounds, temporary modifications, configuration control issues and other items for each of the Salem systems.

Phase II involved the scheduling of the restart work, continued evaluation of emerging action items, and the development of the System Startup Test Plans (SSTPs), described further in Subsection E, "Integrated Test Program." Phase II was completed for Salem Unit 2 in October 1996.

Phase Ill, FSRR, involved a System Readiness Review walkdown by the System Manager, initiation of the final review of SSTPs, presentation of the results of the UFSAR Macro-reviews to the System Readiness Review Board (SRRB) and Management Review Committee (MRC), and an evaluation of the aggregate impact of work not scheduled to be completed prior to restart of the unit. The UFSAR Macro-review is discussed below in Subsection F, "FSAR Project." The aggregate impact evaluation is performed to discuss the overall impact of post-restart work activities on the system and its capability to perform its design bases function. Phase Ill was completed in November 1996 for Salem Unit 2.

Phase IV, SPA, involves the completion of the final review of the SSTPs, completion of system walkdowns, performance of testing and review of system test results, continuous monitoring of system readiness, and System Readiness Affirmation. The System Readiness Affirmation will address current operator workarounds, temporary modifications, active operability determinations, and other aspects to affirm that the SGS systems are ready to support the restart of the Unit. Phase IV is currently in progress for Salem Unit 2. The SRRP will resume for Salem Unit 1 after completion of the Unit 2 SRRP.

57 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 E. Integrated Test Program In order to develop a systematic approach to the restart of the Salem units, PSE&G established a startup test organization in December 1995, accountable to the General Manager - Salem Operations. A new startup and power ascension program procedure was developed to define the responsibilities and processes to implement the integrated startup and power ascension test program.

The 46 systems in the System Readiness Review Program (SRRP) are the base for development of the Startup System Test Plans (SSTPs). SSTPs also include some systems outside the scope of the system readiness review program. The SSTPs consist of: .

• a matrix of design change packages vs. post-modification testing requirements

• a matrix of work orders vs. retest or post-maintenance test requirements

• industry experience and commitments with impact on te~ting

• recommendations for additional testing

• a summary of the rationale for concluding the system(s) is ready for restart based on the testing specified in the SSTP.

The system manager is responsible for development of the SSTP and obtains supervisory approval. The SSTPs are presented to a Test Review Board (TRB).

The TRB is a multi-disciplinary collegial review board consisting of senior level engineers, supervisors and managers. The TRB chairman is the startup testing manager. The purpose of the TRB review is to advise the chairman on the adequacy of the integrated test program with respect to demonstrating the ability of SSCs to satisfy design criteria. The system manager presents the SSTP to the SORG. The final approval authority of the SSTP is the General Manager -

Salem Operations.

Hold points are used throughout the startup test sequence. Plant performance hold points enable the shift plant manager to determine the acceptability of proceeding to the next set of activities in the schedule. In addition to plant response and test results, plant performance hold point decisions are based on resolution of open and emergent issues, considering completion of corrective maintenance activities, personnel performance and plant support activities.

Planned assessment hold points utilize self-assessments by the line organizations and independent assessments by the Quality Assessment department, to determine the readiness of the plant and the organization to 58 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 proceed with the startup plan. These assessments result in recommendations to the General Manager - Salem Operations relative to proceeding with the startup schedule.

10CFR50.59 reviews are performed for tests resulting fromdesign changes as well as stand-alone startup and power ascension tests. A startup and power ascension sequencing procedure has been developed. The 10CFR50.59 evaluation for the sequencing procedure addresses the aggregate impact of testing activities. The 10CFR50.59 process ensures the test performance is consistent with the plant design bases as described in the Safety Analysis Report (SAR).

The integrated test program establishes a systematic approach to demonstrating the performance of SSCs with respect to their design bases, following the extended SGS unit outages and associated maintenance and modification activities. This program therefore helps provide reasonable assurance that design bases requirements are translated into operations, maintenance and testing procedures.

F. FSAR Project

• The results of self assessments and a "Salem Licensing Bases Team Inspection" (Inspection Report 96-80), conducted by the NRC in several phases between May and October, 1996, identified a further need to provide reasonable assurance that, upon restart, Salem Unit 2 would be operated within its licensing and design bases. The NRC inspection identified instances of weakness in licensing and design bases documentation and incorporation of the licensing **

and design bases into plant procedures and configuration.

The FSAR Project integrated a series of new initiatives with several recently completed or on-going Salem improvement programs related to the plant licensing and design bases. The combination of new initiatives and the on-going programs contribute to reasonable assurance of operation within the licensing and design bases. These initiatives and on-going programs reviewed the design and licensing bases against current plant procedures and configuration control documentation. While the project activities focused, on Unit 2, the vast majority of items are common to both units.

The new initiatives implemented by the FSAR Project Plan were:

• UFSAR Macro-Review for Unit 2 SRRP Systems

• UFSAR Chapter 15 Safety Analysis Input Review

• UFSAR Vertical Slice Reviews of Select Systems 59 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996

• Discrepancy Evaluation Review Form Review

• Engineering Evaluation/Justification for Continued Operation Review

• SER Review The on-going programs (when the FSAR Project was being -conducted) were:

• Configuration Walk-downs

• Maintenance Rule Implementation

• Technical Specification Surveillance Improvement Program (TSSIP)

• Emergency Operating Procedure (EOP) Upgrades

• System Readiness Review Program

• System Startup Test Program (Integrated Test Program)

The new initiatives were conducted from June to September, 1996. These programs received substantial oversight and monitoring throughout the period, including independent technical experts who reviewed the initial plan, other independent technical experts who monitored in-progress reviews and technical activities, Quality Assurance/Nuclear Safety Review (QAfNSR) surveillance, and NRC inspections. At the end of the FSAR Project, an NRC team of three.

inspectors spent one week reviewing the program results and interviewing engineering and operating plant staff personnel.

Each of the specific initiatives and on-going programs identified above that are germane to the incorporation of design bases information into plant procedures and testing, are described in this section. "Response to Question C" addresses the attributes of these activities as they pertain to consistency of the design bases with plant configuration and equipment performance.

UFSAR Macro Review (FSAR Project)

The UFSAR Macro-review for each of the SRRP systems encompassed a review of select design bases parameters and system attributes (flow rates, temperatures, pressures, setpoints, automatic start features, etc.) from the UFSAR. These parameters and system attributes were reviewed against the appropriate TS, design output documents, and/or plant procedures to determine proper incorporation of the system attributes into these documents. If the system attributes were not properly incorporated or not consistent with plant procedures, TS, and design output documents, an AR was generated in accordance with the Corrective Action Program to resolve the issue.

60 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter

• Dated October 9, 1996 The UFSAR Macro-review resulted in the generation of a report for each of the 46 systems reviewed, identifying the parameters and attributes reviewed and validated. The UFSAR Macro-review validated approximately 2000 parameters and attributes in total, with 60 ARs initiated by the UFSAR Macro-review team during the review period. In accordance with the corrective action program, each AR was assigned a significance level. Significance levels are discussed in the corrective action program description in the "Response to Question D" section. Significance level 1 is the most safety significant and requires a detailed root cause evaluation whereas a level 4 AR is the least safety significant and requires resolution of the issue identified in the AR. The breakdown of UFSAR Macro-review ARs by significance level is as follows.

• 3 significance level 2 ARs

• 35 significance level 3 ARs

• 22 significance level 4 ARs Of the three significance level two ARs, one involved a discrepancy between the actual total service water system flow rate and the amount of water the service water system would draw from the estuary as stated in the NRG Safety

• Evaluation Report (SER) for the Salem Generating Station, dated October 11, 1974. This was determined to be an editorial error in the SER that does not impact the operation of the service water system. The other two significance level two ARs involve test procedures and are described below:

1. An UFSAR statement regarding leakage inspection of pressure-retaining portions of the containment spray system during operational testing was not found in either the Technical Specifications or the surveillance procedure.

Leakage inspections are performed for the subject piping, but at a lower frequency than stated in the UFSAR. Evaluation of the discrepancy concluded that the leakage inspection requirements in the inservice inspection (ISi) procedures were consistent with ASME XI requirements. The Unit 2 procedures have been revised to include the UFSAR inspection requirement. The Unit 1 procedures are scheduled to be revised prior to Mode 4 entry.

2. The UFSAR contained reactor trip breaker surveillance and reporting requirements which have been removed from the Salem Unit 1 and 2 TS via License Amendments 176 and 157, respectively. This discrepancy is the result of failure to properly update the UFSAR to reflect the NRG approval and implementation of a License Amendment. The UFSAR will be updated to be consistent with the License Amendment.

61 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 The UFSAR Macro-review validated roughly 2000 system parameters and attributes, with an AR generation rate of approximately 3%. The low percentage of ARs written compared to the population of parameters or system attributes reviewed indicates that the a large percentage of the information in the UFSAR is both valid and accurate. The three significance level 2 issues described above were not safety-significant. Completion of the UFSAR updates associated with resolving the 60 ARs identified by this review will enhance the thoroughness and accuracy of the UFSAR. Therefore, the UFSAR Macro-review helps provide reasonable assurance that design bases are properly translated into operations, maintenance and testing procedures.

Chapter 15 Safety Analysis (FSAR Project)

As part of the FSAR project, system and component-related inputs and assumptions to the Salem UFSAR Chapter 15 accident analyses were identified in a matrix with a reference to the licensing basis calculation file that demonstrates the safety limits are met for the analyzed accident. These inputs and assumptions were then reviewed against current plant testing procedures to validate that current testing of the parameters is consistent with the assumptions used in the accident analyses. Where testing is not performed to. demonstrate accident analysis inputs, calculations or other design documents were reviewed to validate the input assumptions used in the accident analyses. Where calculations demonstrated acceptability, the calculations were reviewed for inputs, assumptions, methodology and reasonableness of output. When an input assumption could not be verified using a test or surveillance in the Salem Unit 2 TS, or by reviewing a design calculation or other design document, an AR was written in accordance with the corrective action program. Approximately 400 input parameters were reviewed with the majority of the parameters being verified by test or surveillance. 25 ARs (5 -Level 2, 19 - Level 3, and 1 - Level 4) were either previously written or documented during this review to track the closure of open issues identified during the review. These ARs were screened for restart applicability in accordance with the restart screening criteria. Items that were identified as restart-required per the restad screening criteria include, but are not limited to, proposed actions to resolve safety, operability and design bases issues. The majority of the ARs written were due to a lack of source documentation and not a problem with the validity of the input assumption. As of January 9, 1997, all but 5 of the ARs had been closed out. Closure of the remaining 5 ARs is scheduled prior to Unit 2 restart.

UFSAR Vertical Slice Reviews (FSAR Project)

During the FSAR project, vertical slice reviews were conducted for a total of 7 safety-significant systems.

62 Question 8

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Fuel Pool Cooling Safety Injection Reactor Protection System Dated October 9, 1996

• Auxiliary Building Ventilation (ABV)

• Fuel Handling Ventilation (FHV)

• Containment Building Ventilation (CBV)

• Miscellaneous Ventilation Systems (Service Water Intake Structure, EDGs, Switchgear and Penetration Area) were added as additional scope.

These vertical slice reviews were multi-disciplinary team reviews starting with the licensing and design bases documents such as the UFSAR, NRC Safety Evaluation Reports, NRC regulations, docketed commitments, and Technical Specifications. Key descriptions and parameters such as operating and design limits, configuration descriptions, and descriptions of system operation were compared to the as-built configuration of the system and to the system operation (operating and emergency procedures, etc.) to assess whether the plant is built and operated in accordance with the licensing/design bases. Since the Configuration Walkdown Project verified the as-built plant piping, instrumentation, and electrical configurations with the piping and instrumentation drawings (P&IDs), licensing/design bases system descriptions were reviewed against the appropriate P&IDs. Various parameters (i.e., setpoints, response times, etc.) identified in the UFSAR were compared with the Technical Specifications for consistency and then reviewed against plant operating and surveillance procedures to ensure that the plant is being operated in accordance with the licensing and design bases.

In non-ventilation systems (Safety Injection, Spent Fuel Pool Cooling, and Reactor Protection System), various minor discrepancies were identified between the licensing and design bases information and the as-built plant configuration and procedures. 25 ARs (2 - Level 2, 15 - Level 3, and 8 - Level 4) were written to document the discrepancies. These discrepancies did not prevent the fulfillment of these systems' intended safety functions.

However, concerns were identified with ventilation systems such as:

• Containment Fan Cooler Unit (CFCU) response time testing did not ensure as-built configuration met the accident analysis

• Auxiliary building charcoal filter testing did not meet ANSI standard testing

• requirements committed to in the Technical Specifications 63 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996

• Assumptions for Auxiliary building charcoal filter alignment in the accident analysis were not support by the current EOPs.

Based on these findings and other issues identified during the ABV, FHV and CBV vertical slices, an additional vertical slice was added to the scope of the project to review the remaining safety related ventilation systems (except for Control Room Area Ventilation, which was undergoing major modifications and licensing basis changes during the concurrent outage). The issues identified for the ventilation systems from the vertical slice reviews were documented in the ARs and reviewed against the restart screening criteria to determine the work necessary to support restart of Salem Unit 2. Procedural and/or physical plant changes were determined to be required to resolve these issues to assure the systems will achieve their safety function. Resolution of the restart-required issues will provide reasonable assurance that the ventilation systems at Salem will be operated in accordance with their design and licensing basis.

Discrepancy Evaluation Form Review CFSAR Project)

The Discrepancy Evaluation Form (DEF) was the method to document and resolve engineering discrepancies. As of July 2, 1995, the DEF process is no longer used to document new engineering discrepancies but will be used to document the resolution of DEFs identified prior to July 2, 1995. Engineering discrepancies identified after July 2, 1995 are entered into the Corrective Action Program via ARs.

Engineering discrepancies are defined in procedure NC.DE-AP.ZZ-0018(Q) as errors, omissions, inconsistencies or conflicts in design documentation, engineering drawings, calculations, analyses, engineering evaluations, or vendor technical documents. The Configuration Baseline Documentation project was one key source of DEFs.

A technical review of closed DEFs was performed during June and July 1996, as part of the FSAR Project, to determine if licensing and design bases issues were properly addressed within the DEF closure. A total of 1752 closed DEFs related to 16 UFSAR Chapter 15 safety analysis systems were screened for licensing and design bases impact. Approximately 500 DEFs were determined to be related to the design and licensing basis. Of the 500 DEFs reviewed, there were 59 DEFs for which the review team could not reach closure. These 59 DEFs were documented in 7 ARs to further evaluate the basis for the closure of ttie DEFs. Of the 7 ARs issued, 4 ARs were significance level 2 and are described below:

• Possible RVLIS errors with containment flooding 64 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996

• Deferral of DCPs associated with RG 1.97 upgrades

• Manual valves isolating non-seismic lines from the RWST

• No documentation for procedure to calculate RCS inventory at low temperature.

Of these four issues, only one (isolation of non-seismic lines from the RWST) resulted in the identification of a significant design basis issue which is documented in LER 272/96-018-00. Further review of the remaining ARs under the corrective action program by the engineering department determined that other DEFs were properly addressed by plant modifications, procedure changes or document updates, however, the DEF closure package was not updated to reflect the proper resolution of these issues.

Based on the large sample of DEFs reviewed associated with safety analysis systems, the limited number of issues identified by the review of DEFs indicates that the DEF closure program was adequately addressing and maintaining the plant licensing and design bases.

Engineering Evaluation/Justification for Continued Operation (EE-JCO) Review (FSAR Project)

The EE-JCO review was initiated due to problems identified at Hope Creek which indicated that old Engineering Evaluations (EEs) which support past Justifications for Continued Operation (JCOs) might still be active and not properly removed from plant procedures. As part of the FSAR Project, a review of Salem EEs associated with the Safety Analysis systems was performed in two phases to assess the JCO-related EEs. The first phase of the EE reviews identified approximately 200 EEs related to the sixteen Safety Analysis systems.

15 EEs were identified as JC Os by title and an additional 100 were screened to determine if they were JCO related. This screening identified 13 additional EEs as JCOs for a total population of 28. An initial review of the 28 EE-JCOs identified that:

• 9 EE-JCOs should be revised and classified as inactive

• 7 required confirmation of completed DCPs and then could be classified as inactive

• 12 required additional action to resolve.

Based on the above findings a second phase of EE reviews was conducted .

• The remaining EEs associated with the sixteen Safety Analysis systems and an 65 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 additional 100 EEs associated with risk significant systems, as classified by the maintenance rule, were reviewed. The results of this review are as follows:

• 11 EEs are JCO-related

• 27 EEs are not JCO-related and require resolution for Unit 2 restart

• 49 EEs are not JCO-related and resolve the evaluated plant condition but recommend follow-on actions

• 49 EEs provide adequate resolution of the plant condition with no further action required.

Improvements have been made in the control and tracking of JCO information.

Procedure NC.DE-AP.ZZ-0058(Q) for control of JCOs was voided in September 1996. The process for identifying and controlling JCOs was incorporated into the Corrective Action Program and Operability Determination process.

Equipment operability issues are identified via the Corrective Action Program by generation of an AR. Upon being informed of an operability issue, the on-duty SNSS/NSS determines and directs any immediate actions needed to place the plant in a safe condition. The results of the initial operability screening for the affected SSC(s) are documented in accordance with Operations Department procedures. Based on the initial operability screening, the SNSS/NSS determines the need for a follow-up operability assessment by system engineering. If operability of an SSC is dependent upon a restriction on plant or system operation, or compensatory actions, then the SSC is operable but degraded or nonconforming. The SNSS/NSS is responsible for ensuring implementation of any such restrictions or compensatory actions necessary to maintain the affected equipment operable. Actions to restore degraded or nonconforming SSCs to full qualification status are tracked and implemented via the Corrective Action Program. The number of active operability determinations are tracked on the Nuclear Business Unit (NBU) performance indicator system.

This allows trending of the closure of operability determinations and progress made in eliminating active operability determinations.

Safety Evaluation Report (SER) Review (FSAR Project)

SER's associated with License Amendments issued through May 31, 1996, were reviewed to assure that the UFSAR was appropriately updated to reflect the changes implemented by the License Amendments. The review consisted of 181 Unit 1 and 162 Unit 2 License Amendments. As a result of.these reviews, a total of 28 ARs (classified as significance level 3 or 4) were written involving primarily minor errors in the UFSAR updates. Often times the failure was to update a table or section of the UFSAR where information is repeated. The deficiencies identified by the review indicated that the TS changes were correctly implemented but the companion UFSAR changes were not always 66 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 thorough or complete. The lack of TS issues and the minor nature of the failures to fully update the UFSAR are sufficient to provide reasonable assurance of plant operation within the plant licensing and design bases.

FSAR Project Overall Conclusion The FSAR project was conducted to provide reasonable assurance that Salem Unit 2 would be operated within its licensing and design bases. Approximately 200 ARs were written as a result of the initiatives summarized above. These initiatives compared licensing basis requirements and commitments against the UFSAR, Technical Specifications, Chapter 15 Safety Analysis, and plant engineering and design documents including drawings, calculations, test procedures, and specifications. The FSAR project review activities, and implementation of the resulting restart-required corrective actions, provide reasonable assurance of plant operation in accordance with its licensing and design bases.

G. Instrument Valve Lineup Procedures In March 1996, an instrument valve for a pressure switch in the Unit 2 service water system was found to be mis-positioned. Corrective actions for this condition included a review of past mis-positioning events, which prompted a root cause evaluation. The root cause was determined to be failure to require documentation of valve manipulations during the performance of instrument calibrations.

The Salem Maintenance Department led a project to develop instrument valve procedures with lineup sheets, and to label any unlabelled instrument valves.

The project was implemented for 24 of the 46 Unit 2 restart systems. The remaining 22 restart systems were determined not to require valve lineup.

procedures (e.g., the system is not a fluid system; the valves are included on P&IDs). The 24 Unit 2 systems in the project scope had procedures developed for valve lineup verification. The lineup procedures were developed by creating a list of instrument sensors and instrument valves for each system, specifying the valve identification number, location and proper position. The list was verified by physical plant walkdown. Valves that were found to be unlabelled were provided with identification tags. This project was completed for Unit 2 via issuance of procedures, in January 1997. A similar project is planned for Unit 1 prior to Unit 1 restart.

Proper identification and control of instrument valve lineups via procedure helps ensure consistency with the design bases relative to the performance of instrumentation in systems important to safety.

67 Question B

SALEM GENERATING STATION -*.

Response to 10 CFR 50.54(f) letter Dated October 9, 1996 IV. ASSESSMENT OF EFFECTIVENESS This section provides an assessment of the effectiveness of key processes used to provide reasonable assurance that design bases requirements are translated into operating, maintenance and testing procedures.

A. Design/Configuration Change Two aspects of the DCP process with particular relevance to the translation of design bases requirements into operating, maintenance and testing procedures are:

1. post-modification testing, which is a special case of post-maintenance testing (PMT). PMT is the testing and/or inspection of a component to demonstrate conformance to design specifications following the performance of a work activity. Effectiveness of PMT in general is discussed in Subsection B.

2. the proper identification and incorporation of station procedure revisions necessitated by design changes.

Post-Modification Testing In October and November, 1995, a third party assessment of DCP quality was performed using vertical slice, performance-based review techniques. The review included verification that selected DCPs contained the post-modification testing requirements necessary to demonstrate physical integrity and functionality of affected equipment, and the objectives of the modification were satisfied. A total of seventeen SGS modification test-related open items were generated by the review team. The majority of these items were dispositioned by either providing clarification or technical justification for the test approach used, or by modifying the affected DCP using Modifications, Concerns and Resolutions (MCRs) in accordance with administrative procedure NC.DE-AP.ZZ-0017(Q), "Modification Concerns and Resolutions." Two of the seventeen issues are associated with ARs for specific design bases-related issues: one involving the assumptions associated with the spent fuel rerack DCP (significance level 2) and another involving auxiliary building ventilation system flow rates (significance level 1). The significance level 1 AR had been initiated prior to the commencement of the third party review activities. These issues have been evaluated and are being addressed in accordance with our corrective action program.

As part of the implementation of the Integrated Test Program (discussed above in the "Validation Reviews" section), the Test Review Board (TRB) identified several concerns with the quality of the test sections of DCPs associated with 68 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 System Startup Test Plans (SSTPs). A significance level 1 AR was initiated in May, 1996. The root cause evaluation performed for this AR concluded that there was no safety significance to the particular identified issues and that the issues were not generic in nature. There was no concern identified which cal.led into question the ability to demonstrate design bases capability via post-modification testing.

During an ongoing inspection of the startup test program, NRC inspectors identified concerns with the test sections of a DCP for the Control Area Ventilation System (CAV). A significance level 2 AR was initiated in December, 1996. A team of test engineers reviewed the subject CAV DCP and an additional sample of 10 DCPs, to evaluate the specified post-modification testing for each package. The review team subsequently evaluated the testing associated with 8 DCPs for the Auxiliary Building Ventilation System (ABV).

Based on these reviews, and evaluation of the specific issues identified by the NRC inspectors, there were no concerns identified relative to the ability to demonstrate design bases performance requirements. Process improvements were identified and are being implemented in accordance with our corrective action program.

Procedure Changes Resulting From DCPs A third party review of the DCP process was conducted from November, 1995 through April 1996. This review included evaluation of the adequacy with which procedure revisions were identified and incorporated as part of the DCP process. Each of the twelve SGS DCPs reviewed specifically for adequacy of identification and incorporation of procedure revisions, was determined to be acceptable (one of the twelve was considered questionable in the review team's final report, but confirmed to be adequate by PSE&G). The third party review team noted that the identification of procedure impacts by the organizations most capable of making such determinations, is a process strength.

• Conclusion

• The assessments and resulting actions summarized above help provide reasonable assurance that implementation of the DCP process effectively translates design bases requirements into operating, maintenance and testing procedures.

B. Configuration Control The effectiveness of the Work control, safety tagging and post-maintenance testing/operability re-testing (PMT/RT) processes are described in this section.

While these processes do not directly translate design bases requirements into

69. Question 8

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 procedures, they are relevant because they establish controls on implementation of work activities governed by station procedures. These processes help assure that the implementation of operating, maintenance and testing procedures is controlled such that the removal of equipment from service and restoration to operable status is done in a manner consistent with the plant design bases.

Operator workarounds and burdens are also described in this section.

Identification and removal of operator workarounds and burdens helps ensure that SSCs are operated and configured in accordance with their design bases.

Work Control The need for improvement in the Work Control Process to support SGS restart was identified by both PSE&G and the NRC. PSE&G recognized that excessive action requesUwork order corrective maintenance (AR/WO CM) backlogs, due to low staffing levels and process inefficiencies, inhibit the ability to effectively process work on a real-time basis. Backlogs at SGS included duplicative and low-value items which interfere with the identification and processing of more significant equipment issues, thereby challenging the ability of the station organization to maintain the plant consistent with the design bases.

The Salem Restart Action Plan addresses Work Control Process* issues and backlogs. A multi-department Work Control Team led by the Work Control Program Manager, was established to define and implement an effective Work Control Process. A Work Control Manual was developed to institutionalize process controls. Departmental Desk Guides (e.g., Operations, Maintenance, Planning) were developed to provide specific process implementation instructions.

Approximately 6500 AR/WO CMs were reviewed to determine if they were valid CM WOs, resulting in a reduction of approximately 31 % of the backlog due to duplicate or otherwise invalid CM WOs. Performance indicators were developed to provide consistent definition and tracking of backlogs. A cross-functional team has been established to address issues involving work order holds (i.e.,

work activities on hold for parts or engineering potentially resulting in excessive delays and process inefficiencies). Performance indicators measuring backlogs and activity holds are used to monitor the effectiveness of the Work Control Process improvements.

The Salem Planning and Maintenance Departments underwent an intervention effort in 1996 to address performance issues and bring work quality in line with industry standards. These intervention efforts were completed by November, 1996. The Planning and Maintenance Intervention improved areas such as communication, job knowledge and skills.

Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996_

A self-assessment of the Planning Department was completed in December, 1996. Ongoing corrective actions resulting from the Planning Department Self-assessment include improved communication between Planning and Scheduling groups and improved communication between the Engineering, Planning and Maintenance Departments.

The backlog reduction efforts and process improvements described above help to facilitate the planning, scheduling and completion of activities necessary to support implementation of maintenance and modification work on plant SSCs.

These efforts, combined with continuing assessment of process implementation, also help provide reasonable assurance that the plant is operated, maintained and tested consistent with its design bases.

Safety Tagging Although the safety tagging program is a personnel safety measure, it also provides control over SSC configuration (e.g., valve lineups) and protects plant *-

equipment. Its effectiveness is discussed because it is a configuration control process used in conjunction with work control planning and scheduling functions, to ensure equipment is operated, maintained and tested consistent with its design bases during the performance of work activities.

NRC Special Inspection Team (SIT) report number 50-272; 50-311/95-80, dated June 9, 1995, acknowledged process improvements and Sf:!lf-identification of negative trends relative to mis-positioning events and configuration discrepancies. However, the inspection team also noted that no plan had been .

developed to address the identified negative trends, and had several findings in the areas of definition and communication of process improvements, and training.

At PSE&G's request, Failure Prevention Incorporated (FPI) completed an independent common cause analysis report regarding tagging issues in August, 1995. PSE&G initiated a significance level 1 AR in January, 1996, to perform a root cause analysis of tagging-related events trended by the Operations department. The key findings of the FPI and PSE&G cause evaluations are similar:

• the complexity of the safety tagging process results in many skill-based and rule-based human ,_errors

• the human errors suggest programmatic errors of excessive implementation requirements and inadequate self-verification

• the errors have root causes in inadequate program design, self-verification process, training, supervisory methods, work planning and work practice.

71 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 As a result of the weaknesses identified by the root cause analysis the following corrective actions were implemented.

• Staffing levels and personnel qualifications for tagging functions were upgraded to provide. sufficient support of tagging functions with personnel qualified via training and qualification cards

• Self-checking and human error reduction techniques were incorporated into an operations training lesson plan and presented to Operations Department personnel

• Desk Guides for tagout preparation were issued to provide clear direction for the required elements of tagout preparation

• NC.NA-AP.ZZ-0015(0), "Safety Tagging Program" has been revised to simplify the program and incorporate guidance from night order book entries and memos. This revision became effective on January 15, 1997

• Operations will improve methods of monitoring tagging process effectiveness and review tagout performance during the current outage. Post-outage critiques to identify process improvements to eliminate tagging errors associated with modification activities will be performed subsequent to restart from the current outage

• The Quality Ass~ssment department will evaluate the effectiveness of '

tagging-related corrective actions, in accordance with the corrective action program.

Based on lessons learned from assessments of the tagging process, significant process and performance improvement actions were taken to improve the safety tagging program. These actions, combined with continued assessment and corrective action implementation, provide reasonable assurance that the safety tagging program, as a configuration control process to protect plant equipment, will help maintain the plant consistent with its design bases during the performance of operations, maintenance and testing activities.

Post-maintenance Testing/Operability Re-testing (PMT/RT)

Administrative procedure NC.NA-AP.ZZ-0050(0), "Station Testing Program,"

describes the process for Post-Maintenance Testing and Operability Retest (PMT/RT). PMT is the performance of tests or inspections to demonstrate conformance to design specifications. RT is the testing performed to demonstrate operability (i.e., ability to perform specified safety functions).

PMT/RT requirements are specified consistent with the maintenance or modification activity being performed.

72 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Based on self-identified programmatic weaknesses in the Station Testing Program for PMT/RT, a significance level 1 AR was initiated in May, 1996. A root cause evaluation was completed in June, 1996, identifying the following failed barriers:

• Work planners, who are responsible for specifying retest requirements, lack system training

• The principal procedures specifying retest requirements (NC.NA-AP.ZZ-0050(0) and M23B) had not been revised to update their technical data for several years

• There was no review and approval cycle for technical/operations review of specified retest activities

• There was lack of ownership of the maintenance testing and operability retest programs. Although the Salem Maintenance department was the NAP-50 sponsor, Maintenance performs the retests as specified on the work order.

INPO good practice (GP) MA-305 recommends ownership of retest programs by system engineering.

A multi-discipline team was established, with operations and system engineering support, to address post-maintenance testing and operability retest issues, including the development of a revision to NAP-50. NAP-50 was revised in December, 1996. The sponsor organization for NAP-50 is now Salem System Engineering. System manager responsibilities were expanded to include resolution of testing issues with job planners and supervisors, and specifying test requirements as necessary. NAP-50 was revised to improve communication to the planner, regarding changes in work scope with potential to impact post-maintenance or operability retest requirements. The technical guidance in the procedure was updated, including the identification of trigger values for specific TS conditional surveillance requirements. The work control center function of identifying or verifying the scope of testing, is to be performed by a licensed operator per revised NAP-50.

The station testing process for.PMT/RT provides the mechanism for testing SSCs affected by modification or maintenance activities to verify their conformance with design bases requirements. Based on the improvements made in response to identified process deficiencies, and continued monitoring of performance via the corrective action program, PMT/RT helps provide reasonable assurance that SSCs are operated, maintained and tested consistent with design bases requirements.

73. Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Operator Workarounds:

Dated October 9, 1996 Operator workarounds are defined as equipment malfunctions or design deficiencies that can cause a plant transient or reduce the effectiveness of plant transient or accident mitigation efforts. Operator burdens are defined as equipment malfunctions, design deficiencies or administrative deficiencies that require the operator to discontinue his normal watch station duties or prevent the operator from completing critical tasks in a timely manner.

In letter dated September 28, 1994, "NRC Performance Assessment of Salem,"

the NRC identified that weaknesses in PSE&G's resolution of long-standing concerns and the inadequate pursuit of problem resolution have apparently contributed to the over-reliance on operator workarounds. During this performance assessment the NRC identified that actions were taken to identify operator workarounds and that PSE&G had begun to implement a program to consolidate and prioritize outstanding work requests, design change requests, and operator workarounds/bypasses.

As part of the Salem Operations Restart Action Plan, additional actions were initiated to improve the identification and resolution of operator workarounds and operator burdens. To strengthen the process for identifying and resolving operator workarounds and operator burdens, procedure SC.OP-AP.ZZ-0030(Q) was issued on June 1, 1996. The procedure requires an AR to be written for identified operator workarounds or operator burdens and notification of the SNSS/NSS to establish any required compensatory actions until their resolution.

A separate AR is written to track any program or procedure revision change necessitated by the operator workaround or burden to ensure the removal of the program or procedure changes once the operator workaround or burden has been resolved. Periodic assessment of the operator workarounds is performed to determine the aggregate impact on the plant and operator capabilities.

An operations supervisor is dedicated to coordinate the reduction of existing operator workarounds and operator burdens. The operations supervisor interacts directly with maintenance management to prioritize and coordinate removal of the operator workarounds and burdens. Approximately 134 combined operator workarounds and burdens existed on Unit 2 during the first week in January 1996. The combined number of operator workarounds and burdens was below 20 by the end of December 1996. The reduction in operator workarounds and operator burdens provides additional assurance that SSCs will be operated and configured consistent with the design bases of the respective SSC.

74 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 C. Technical Specification (TS) Implementation Effective implementation of TS Surveillance Requirements, which prescribe the activities necessary to periodically verify the operability of TS-related SSCs, is critical to ensuring the plant is operated, maintained and tested in accordance with the plant design bases. NC.NA-AP.ZZ-0012(0) (NAP-12), "Technical Specifications Surveillance Program," specifies the requirements for scheduling and implementing TS Surveillance Requirements.

NRC Integrated Inspection Report 50-272/96-15; 50-311/96-15 (IR 96-15), dated December 3, 1996, includes a review of 21 License Event Reports (LERs) issued the last two years, each involving a failure to implement TS. This review was performed to assess the adequacy of the TS surveillance program. The NRC Inspector also reviewed the Technical Specification Surveillance Improvement Program (TSSIP) scope and documentation. It was noted by the NRC inspector that many problems were identified by TSSIP including: 1) TS surveillance requirements with no implementing procedure, 2) surveillance procedures had inadequate purpose and/or acceptance criteria, 3) procedure revisions inappropriately deleted TS Surveillance Requirements, and 4) surveillance requirements existed for features not installed in the plant. The NRC inspector concluded that taken collectively, the TS related problems described in the LERs were indicative of a programmatic failure.

The NRC inspector's review of the TSSIP charter and scoping document found the program provided a comprehensive and thorough approach to res-olving the problems identified. The NRC inspector noted that the review of LERs did not-identify new*problems or problems that would not be resolved by effective implementation of the TSSIP. However, since the review of LERs identified five examples of problems that the second stage of TSSIP is designed to identify and address (that is, they are examples of design information not adequately captured by-TS implementing documents) the NRC inspector concluded that the NRC should assess the acceptability of the basis for not completing the second phase of TSSIP prior to Salem Unit 2 restart in conjunction with NRC restart inspection item 111.3.

Historical occurrences of missed surveillances and other TS implementation issues described in LER 311/95-008 prompted the initiation of the TSSIP and the resulting improvements to administrative controls. Improvements are being made to the TS Surveillance Program (NAP-12) based on lessons learned from TSSIP and prior TS-related events. Revision 7 of NAP-12, effective February 14, 1997, includes the following improvements .

• TS surveillance responsibilities for the responsible implementing departments (e.g., Operations, Maintenance, Chemistry) are emphasized in the procedure.

75 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Responsible implementing departments include a TS Surveillance Coordinator, who is responsible for ensuring that routine surveillance are scheduled for completion on or before the due date. Department managers are responsible for developing and maintaining tracking mechanisms to ensure the required TS surveillances are completed prior to mode changes. Department managers also ensure design change packages (DCPs) are reviewed for impact on their departments' TS surveillances, and that personnel responsible for reviewing completed surveillance results have the necessary training and qualifications.

Another NAP-12 improvement is the standardization of the TS Administrator's role between SGS and Hope Creek. The TS Administrator uses the AR process to manage the creation and revision of recurring tasks used for routine TS surveillance implementation, including those resulting from approved license amendments. The Planning and Scheduling Department is responsible for processing such ARs assigned by the TS Administrator.

Completion of the TSSIP Phase 1 reviews and process improvements prior to unit restart provides reasonable assurance that TS Surveillance Requirements will be consistently implemented by their associated surveillance procedures.

Completion of Phase 2 TSSIP will ensure identification and correction of potential design bases inconsistencies with implementing procedures. The completion schedule for TSSIP Phase 2 is the subject of discussion with NRG as part of the SGS restart action plans.

D. Specific Technical Programs The lnservice Testing (IST) Program, described in the "Response to Question A" section, bears particular relevance to the translation of design bases requirements into operating, maintenance and testing procedures. This program provides a link between the design bases performance requirements and test criteria to demonstrate the capabilities of the SSCs in the program's scope.

NRC Inspection Report 50-272: 311-94-21, dated November 30, 1994, documented an NRC review of the SGS IST program. The inspection concluded that the program generally met regulatory requirements and had high quality procedures. However, a concern was expressed that several shortcomings previously identified by PSE&G quality assurance remained uncorrected. The NRC cited two violations in Inspection Report 94-21; one for specific check valve testing noncompliance with 10CFRSO.SS(a), and one for failure to take effective corrective action relative to pump test discrepancies. Inspection Report 94-21 expressed a concern that additional management oversight of the program was warranted.

76 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 The Quality Assessment department performed an audit of the IST program resulting in a July, 1995 stop work order for the IST program and a significance level 1 AR Corrective actions taken in response to the identified programmatic deficiencies include organizational changes and increased staffing levels to improve the overall effectiveness of the program. The IST program was benchmarked against nuclear facilities with recognized good performance in program implementation. A system-based re-validation of IST program scope was performed, resulting in updates to IST basis documents for over 1800 pumps and valves. A complete verification and validation of the IST processes for adherence to applicable requirements was performed by a third party.

Recent specific issues related to the effectiveness with which the design bases requirements are translated into the IST program have been identified and documented as ARs. In accordance with our corrective action program, these specific issues will be addressed, as will any programmatic issues identified via cause analysis and trending. These issues are also being evaluated by the NRC as part of their review of the IST program as a technical restart issue.

The significant recent program improvements and reviews, combined with continued oversight and correction of identified issues via the corrective action program, will provide reasonable assurance that testing of safety-related pumps and valves per the IST program reflects the proper translation of design bases requirements into operating, maintenance and testing procedures.

E. 10CFR50.59 Process (NC.NA-AP.ZZ-0059(Q): NAP-59)

Assessments of the 10CFR50.59 applicability reviews and safety evaluations were completed by Offsite Safety Review (OSR) group in November 1995. The assessment of 10CFR50.59 applicability reviews was performed in accordance with PSE&G's commitment to perform periodic assessments in response to weaknesses identified in NRC Integrated Plant Assessment Team (IPAT) inspection report 90-81 concerning the 10CFR50.59 applicability review process.

During the November 1995 assessment, applicability reviews were assessed against the following criteria:

• Was the applicability decision performed in accordance with the 50.59 procedure?

• Was there sufficient basis provided to support the conclusion?

• Did the document involve an Unreviewed Safety Question (USQ)?

• 77: Question 8

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 A sample of applicability reviews that involved safety related systems, non-safety related systems potentially impacting safety systems, or documents appearing to have incorrect 50.59 applicability conclusions were selected (48 documents were reviewed). Overall, the assessment of applicability reviews concluded that no USQs existed however, two applicability reviews associated with Temporary Modifications incorrectly assessed that a 50.59 safety evaluation did not apply and two procedure revisions did not receive an applicability review as required by procedure NC.NA-AP.ZZ-0001 (Q) (for procedure revision control).

The OSR assessment of safety evaluations was an analysis of OSR's review of approximately 600 safety evaluation performed between January 1, 1994, and June 30, 1995, to determine weaknesses in the 50.59 program. Ninety-two percent (92%) of the safety evaluations were determined to meet the requirements of 10CFR50.59 and PSE&G's safety evaluation program; however, the remaining eight percent (8%) contained documentation deficiencies. These 8% had insufficient bases contained in the safety evaluation to support the conclusions, or the quality/accuracy of the content of the safety evaluations were not in accordance with the requirements of NAP-59. No changes involving a significant safety issue or USQ were implemented.

A self assessment of the 10CFR50.59 program (NLR-PA-96-002) was performed by PSE&G from January 5 to February 17, 1996, to address concerns with the quality of the 10CFR50.59 safety evaluations identified by OSR and SORC and to assess the overall condition of the 10CFR50.59 program. The assessment evaluated the 10CFR50.59 program against industry standards and requirements (10CFR50.59, 10CFR50.71(e)) focusing on procedures and controls, program implementation, training and qualification, and a review of 50.59 programs at other facilities. Some of the results of the self assessment identified that: 1) one example in NAP-59 was misleading, 2) continuing problems with the quality of safety evaluations presented to SORC, 3) performance feedback to preparers, peer reviewers and approvers is not consistent, and 4) no formal training/qualification requirement for personnel performing safety evaluations. Consistent with the results of the self assessment, PSE&G issued Revision 4 of NAP-59 on May 2, 1996 to include some of the good practices observed in other facilities 50.59 programs. A subsequent revision to NAP-59 was issued on August 30, 1996, to establish formal training and requalification requirements for 10CFR50.59 preparers, peer reviewers, and approvers.

The NRC conducted a review of the 10CFR50.59 Safety Evaluation Program to assess the effectiveness of PSE&G's corrective actions for addressing programmatic deficiencies identified by the NRC and PSE&G, and documented this review in NRC Inspection Report (IR) 96-13 dated December 3, 1996. The NRC inspector reviewed NAP-59, Rev. 4, and noted that the procedure was 78 Question 8

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 largely based on NUMARC's NSAC-125 guidance which conflicted with the NRC Inspection Manual Part 9900 guidance for USQ determinations regarding the increase in the probability of occurrence or consequences of an accident.

PSE&G had evaluated the differences between the NRC Inspection Manual Chapter 9900 (issued on April 9, 1996) and the 10CFR50.59 procedure under the Correctiv_e Action Program. As a result of this review, PSE&G issued Revision 5 of NC.NA-AP.ZZ-0059(0), removing the existing guidance for performing 10CFR50.59 applicability reviews and safety evaluations from the 10CFR50.59 procedure and issued administrative standard NC.NA-AS.ZZ-0059(0). This administrative standard incorporated the NRC's interim guidance on 10CFR50.59 safety evaluations. The NRC inspector concluded that the Salem staff had improved the quality and effectiveness of the 10CFR50.59 program through self assessment, improved training, and program revision.

However, the NRC inspector also stated that they would continue to evaluate the program implementation due to the recently implemented program changes.

More recently, the NRC observed during a monthly inspection exit meeting for Hope Creek Inspection 96-09 that the implementation of 10CFR50.59 with respect to procedure changes was weak in comparison to the plant modification process. The NRC expressed concern that it may be too easy to conclude that 10CFR50.59 does not apply, thereby preventing SORC from reviewing the proposed procedure change. Although this issue was identified at Hope Creek, a similar issue was identified at Salem concerning the adequacy 10CFR50.59 applicability reviews for procedures. As a result of this issue, the NRC issued violation 50-272 &311/96-17-01 on January 8, 1997. PSE&G is currently conducting a root cause analysis to evaluate the 10CFR50.59 implementation for procedure revisions under the Corrective Action Program.

Effective implementation of the 10CFR50.59 program will ensure that changes to the physical plant and procedures will be properly evaluated against the design bases. In assessing the effectiveness of the 10CFR50.59 program, PSE&G concludes that a good framework has been established for an effective 10CFR50.59 program. Recent internal and external assessments have found the program implementing procedures to be well written with good guidance provided to the user. Training is adequate and recent improvements in the qualification program will ensure that personnel performing 10CFR50.59 review activities have an adequate understanding of the regulation. However, recent assessment activities have identified weaknesses in program implementation which are being addressed through the corrective action program. Future periodic assessment of the implementation of the 10CFR50.59 program will provide a means to measure the effectiveness of the 10CFR50.59 process to ensure that modifications to the plant and changes to plant procedures are properly evaluated against the plant design bases.

79 Question 8

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter F. UFSAR Maintenance Dated October 9, 1996 In August 1995, the NRC issued violation 50-354/95-10 as a result of Hope Creek UFSAR Revision 6 not reflecting all changes up to a maximum of six months prior to the date of filing as required by 10CFR50.71(e)(4). A root cause analysis was performed in response to this violation which determined that this issue was also applicable to Salem.

Corrective actions as a result of the root cause analysis were:

• Elimination of the change notice backlog

• Issuance of a letter to UFSAR copyholders identifying the backlog and outstanding UFSAR changes. This list was to be consulted when reviewing the UFSAR for 10CFR50.59 safety evaluations.

• Interim revisions to the Salem and Hope Creek UFSARs (which were issued in December 1995)

• Review and revision of procedures to correct deficiencies that contributed to the creation of the UFSAR change backlog.

The procedures controlling the maintenance of the UFSAR (NC.NA-AP.ZZ-0035(0)-(NAP-35) and NC.LR-AP.ZZ-0013(Z)-(LRAP-13))and those procedures that interface with the process (NC.NA-AP.ZZ-0059(0)-(NAP-59) and design modification workbooks) were revised to include some of the following changes:

• added requirement to process related UFSAR changes, technical specification changes and technical specification basis changes together

• notification of Licensing that a design change had been implemented was moved from the part B closure process (paper closure) to the part A closLJre process (system return to operation)

• added question on safety evaluation form "is a UFSARchange required?"

and a place to document the UFSAR Change number

• added requirement to process any related UFSAR change with the primary document being reviewed (i.e., procedure or DCP) *

• UFSAR change is to be fully prepared and included in the design change package.

In May 1996, the Offsite Safety Review (OSR) group perform*ed a review of the adequacy of the UFSAR Maintenance program. This assessment only addressed the adequacy of the UFSAR Maintenance program and did not attempt to address whether Salem Generating Station is being operated 80 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 consistent with UFSAR commitments. This assessment reviewed the program definition and program implementation and assessed the corrective actions from violation 50-354/95-10.

OSR identified that although the overall UFSAR Maintenance program responsibility lies with the Licensing Department, many other processes, and organizations, are relied upon to assure that the UFSAR is properly up-to-date (e.g., design change, procedure revision,). OSR noted that the design change package process (NAP-8), the safety evaluation process (NAP-59), and LRAP-13 were revised to reflect interfaces in initiating a UFSAR change and a procedure revision request was submitted to revise the procedure control process (NC.NA-AP.ZZ-0001 ). The awareness of approved changes to the UFSAR, not yet physically incorporated into the UFSAR, is also necessary to*

assure the adequacy 10CFR50.59 safety evaluation process. This concern was addressed by issuing a list of approved changes to be inserted in the front of the UFSAR controlled copies for reference when performing a safety evaluation. :

The review of program implementation by the OSR group identified that the UFSAR Maintenance process is supported by a full time coordinator. With the support of the full time coordinator and other resources, the UFSAR change backlog was eliminated and an interim revision to the Salem UFSAR was submitted in December 1995.

A review of three Salem DCPs by OSR identified cases where a completed UFSAR change notice was not included with the DCP (ARs were generated to

. document these specific cases). However, OSR concluded that these instances appear to be isolated performance issues, based upon an OSR review of over 200 documents. OSR also identified that as a result of the System Readiness Review Program (SRRP) performed for Salem, over 70 issues relative to the contents of the UFSAR were identified by the System Managers. These issues are being addressed in accordance with the corrective action program.

Based on the elimination of the change notice backlog, issuance of the interim revision of the Salem UFSAR in December 1995, and the improved procedure and programmatic controls as well as appointment of a full-time UFSAR coordinator, OSR concluded that the UFSAR Maintenance Program is adequate to support the restart of the plant.

Based on the above, PSE&G concludes that process improvements made in the UFSAR Maintenance program will ensure that changes to the plant, whether by design changes, procedure changes; etc., are adequately incorporated into the UFSAR. The improvements made to this process will ensure that the licensing and design_ bases information is continuously maintained. However, although the backlog of UFSAR changes was eliminated, thereby increasing the accuracy 81' Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 of the UFSAR, issues concerning the technical accuracy of the UFSAR were identified by the SRRP and FSAR Project (described in the "Validation Reviews" section) and may continue to be identified through implementation of ongoing review activities and the activities outlined in Attachment 2. As these issues are identified they will be resolved in accordance with the corrective action program and UFSAR Maintenance program. With the resolution of identified UFSAR discrepancies, the increased accuracy of the UFSAR will result in a more effective assessment of the impact of plant modifications and procedure changes against the design bases information contained in the UFSAR under the 10CFR50.59 program.

G. Procedure Control As discussed in the "Validation Reviews" section of this response, the PUP was the first major effort at SGS to validate implementing procedures against elements of the plant design bases subsequent to NRC issuance of the plant operating licenses.

Although the PUP was an extensive review and rewrite of the SGS implementing procedures that resulted in improvement of the implementing procedures, the NRC noted in the Systematic Assessment of Licensee Performance (SALP)

Report 93-99 (January 3, 1995) that procedure adequacy continued to be a problem. During 1996, the NRC issued violations in Inspection Reports 96-06 and 96-08 concerning the adequacy of procedure content. Resolution of the adequacy of procedures and the procedure revision backlog is being tracked by the NRC as NRC Programmatic Restart Issue P-3.

As part of the Salem Operations Restart Action Plan, Department Administrative Procedures, Emergency Operation Procedures (EOPs), Abnormal Operating Procedures and Alarm Response Procedures were reviewed and revised as appropriate. Outstanding procedure revision requests were reviewed as part of the System Readiness Review Program. Procedure revisions required to support the safe and reliable operation of the plants have been scheduled for implementation prior to restart. Approximately 200 normal operating procedures and 85 inservice testing procedures were also reviewed and revised. Abnormal Operating Procedures, Integrated Operating Procedures, and Safe Shutdown Procedures were validated by simulator exercise or other methods to determine the procedures' usability in the field. Emergency Operating Procedures (EOPs) were reviewed, revised and validated as discussed in the "Validation Reviews" section of this response. TS surveillance procedures were reviewed by TSSIP as discussed in the "Validation Reviews" section of this response.

82 Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 On December 6, 1996, revisions to NC.NA-AP.ZZ-0001 (Q), "Nuclear Procedure System" (NAP-1) and NC.NA-WG.ZZ-0001 (Q), "Procedure Writers Guide" (NWG-1) were issued. These procedure changes were issued to address concerns with the adequacy of procedures identified by both the NRC and PSE&G. Key elements of the procedure control process per NAP-1 and NWG-1 that affect the maintenance and incorporation of design bases information into procedures are as follows:

• New or revised implementing procedures and "Q"-designated (procedures for safety related activities in Regulatory Guide 1.33) administrative procedures are required to receive a 10CFRS0.59 applicability review

• New or revised "Q" designated implementing procedures and implementing procedures described in Technical Specification 6.8.2 are required to be reviewed by a Station Qualified Reviewer (SOR)

• The SOR qualification requirements were incorporated into NAP-1

• New sections were added to the NWG-1 that discuss procedure bases and general requirements and the Verification and Validation of procedures.

Since these procedures were recently revised, the effect of the changes to NAP-1 and NWG-1 on the content of SGS procedures can not be measured at this time. However, PSE&G believes that the process controlling changes to procedures and generation of new procedures has been enhanced to alleviate the concerns that have led to the inadequate procedures identified by the NRC and PSE&G.

Although weaknesses were recently identified by the NRC and PSE&G in the adequacy of procedures in the NBU as discussed above, the initiatives (i.e.,

PUP, EOP Upgrade, TSSIP, FSAR Project) undertaken by PSE&G as described in the "Validation Reviews" section of this response provide reasonable assurance that procedures are consistent with the design such that systems will be able to perform their intended safety function. To provide additional assurance that design bases requirements have been translated into procedures, PSE&G will be undertaking additional review activities as discussed in Attachment 2.

V. CONCLUSION As discussed above, PSE&G has performed review activities to assure the adequacy of plant procedures and to evaluate the incorporation of design and licensing basis information into these procedures. These activities consisted of

• the Procedure Upgrade Project, Emergency Operating Procedure Upgrade Project, Technical Specification Surveillance Improvement Project (TSSIP),

83' Question B

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 System Readiness Review Program, Integrated Test Program for restart of Salem, the FSAR Project and the Instrument Valve Lineup Procedure project.

These review activities, implementation of process improvements identified via internal and external assessments (as discussed in the previous section) and implementation of the corrective action program, collectively provide reasonable assurance that procedures are consistent with the design bases such that systems at Salem Generating Station will be able to perform their intended safety functions.

Although extensive reviews have been performed to verify incorporation of licensing and design bases information into plant procedures, additional review activities, as discussed in Attachment 2, will continue to further confirm the correct translation of design bases information into procedures.

84 Question 8

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 RESPONSE TO QUESTION C Rationale for concluding that structure, system, and component configuration and performance are consistent with the design bases.

I. Overview This response provides PSE&G's rationale for our reasonable assurance that structures, systems, and components (SSC) configuration and performance at Salem Generating Station (SGS) are consistent with the design bases. The discussion that follows expands on the processes described in "Response to Question A" by identifying and demonstrating how key attributes of configuration control are manifested at SGS, and in doing so provides the requested rationale.

The key attributes applicable to this request are as follows:

• "As built" configuration of plant SSCs are reviewed to be consistent with design bases requirements;

• Functional characteristics of plant SSCs are established through testing and surveillance activities; and

• Processes to control installation and maintenance are defined and appropriately implemented at SGS.

The fidelity of SGS's design bases is described above in the "Introduction" to the SGS response, under the heading of "Adequacy of Design Bases." As discussed, there is reasonable assurance that the design bases have been adequately developed, implemented, and maintained over time. As such, the remainder of this response to "Response to Question C" explains how the design bases have been and continue to be translated and incorporated into the "as built" configuration of the plant. Specifically, this response analyzes the effectiveness of the above described configuration control attributes. Where appropriate, the discussion identifies kn.own weaknesses and details actions taken and/or planned to resolve these matters.

In "Response to Question C," each major heading response addresses the following:

• Implementation of the processes that maintain plant configuration and performance in accordance with the design bases is summarized in the "Process Controls" section.

• The "Validation Efforts" section describes actions which provide reasonable assurance that SSC configuration and performance processes are consistent 85 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter improvements.

Dated October 9, 1996 with the plant design bases and identifies potential process issues or The "Assessment of Effectiveness" section, first, summarizes key independent oversight findings related to the consistency of plant SSC configuration and performance with the design bases. Assessments discussed in this section have been performed by either internal oversight (Quality Assurance and Nuclear Safety Review), third party industry personnel, or the NRG. The section then assesses the effectiveness of process controls in ensuring consistent translation of the design bases into SSC configuration and performance with consideration given to reviews and independent assessment findings.

• "Improvements Implemented or Planned" section summarizes recent process or program improvements.

• "Conclusions" section provides the overall rationale for our reasonable assurance that SSC configuration and performance at SGS are consistent with the design bases.

The discussion that follows the Salem Restart Plan is a comprehensive and systematic process to provide reasonable assurance that plant systems have been reviewed, modified, tested and maintained to support restart. Certain restart processes provide reasonable assurance that SSC configuration and performance are consistent with the design bases. In addition, restart processes establish a disciplined approach to system assessment, ownership, and accountability, thereby promoting the continued safe and reliable operation of the plant systems. PSE&G recognizes that it has a fundamental responsibility to ensure that the licensing and design bases are maintained, that the installed plant conforms with these bases, and that plant operations are consistent with the design bases.

Also discussed below is the FSAR Project, conducted between June and September 1996. It combined a series of new initiatives (for example, the safety analysis input and assumption reviews, and system vertical slice reviews) with the Salem improvement programs (such as configuration walkdowns, EOP upgrades, system readiness reviews, and the restart test program). The combination of these initiatives provides reasonable assurance that the plant will operate within its design bases. Furthermore, to review the effectiveness of these process, PSE&G has conducted evaluations on eleven design bases related processes such as the 10 CFR 50.59 Safety Evaluation process, the Corrective Action Program, UFSAR Maintenance process, Self Assessments.

These assessments provide reasonable assurance that SGS configuration and performance processes are consistent with the design bases.

86 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Activities and processes that ensure plant SSC configuration and performance are consistent with the design bases are of paramount importance because they ensure that:

1. Plant physical and functional characteristics are maintained consistent with design bases;

2. SSC perform their intended functions; and

3. The plant is operated in a manner consistent with the design bases.

In summary, PSE&G has reasonable assurance that Salem's programs and processes (as discussed in the "Response to Question A" section and this section) maintain configuration control at a level sufficient to ensure that plant physical and functional characteristics are consistent with and are being maintained in accordance with the design bases. PSE&G's processes also include activities that will continue to identify and correct deviations in plant configuration and performance from the design bases, as discussed later in th s response.

II. Process Controls The process controls discussed below, and in response to "Response to Question A", gives PSE&G reasonable assurance that processes are in place to ensure that SSC configuration and performance are consistent with the design bases. In addition several supporting procedures help maintain configuration control at the SGS, as described below.

A. Engineering Controls The response to "Response to Question A" describes, in detail, of the processes that maintain the design bases. The following engineering controls are intended to prevent discrepancies in SSC configuration and processes to ensure SSC configuration and processes remain consistent with the design bases.

NC.DE-AP.ZZ-0001 (Q), "Design Bases/Input" (DEAP-1) procedure establishes a method for identifying design considerations and design input used in the preparation of design documents.

NC.DE-AP.ZZ-0002(Q), "Design Calculations and Analyses" (DEAP-2) procedure establishes the technical and administrative requirements for the development, maint~nance, and control of design calculations.

NC.DE-AP.ZZ-0003(Q), "Modification Walkdown Prog~am" (DEAP-3) procedure provides guidance to personnel participating in modification walkdowns. As part 87* Question c

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 of the Design Change Package (DCP) process, walkdowns are performed to verify the existing configuration, to identify causes of plant and document discrepancies; and determine the feasibility, operability, maintainability and testability of the proposed design, or configuration change and, ultimately, conformance to the approved Change Package (CP).

NC.DE-AP.ZZ-0004(0), "Design Drawings" (DEAP-4) procedure outlines the process for preparing, reviewing, and approving design drawings. New drawings or revisions to existing drawings are implemented by a CP developed according to the CP process. Drawings and pending impacts to drawings from CPs are available to site personnel electronically by the Document Management System (OMS).

NC.DE-AP.ZZ-0013(0), "Processing Material, Equipment and 0-Listed Service Specifications" (DEAP-13) procedure defines the method for preparation, issuance, and control of material, equipment, and 0-listed service specifications within the Nuclear Business Unit (NBU). Design considerations and design input information required by DEAP-1, along with the CBD, and the SAR, are reviewed to ensure incorporation of design requirements and impacts. This process ensures that material, equipment and 0-listed services are procured such that the design bases requirements are maintained.

• NC.DE-AP.ZZ-0017(0), "Modification Concerns and Resolution" (DEAP-17) process establishes a consistent method of documenting and resolving "minor" concerns encountered prior to and during the installation, testing, and close-out of a CP.

NC.DE-AP.ZZ-0026(0), "Engineering Evaluation" (DEAP-26) procedure applies to evaluations performed to document reviews, analyses, conclusions, or recommendations on topics including root cause analysis/problem analysis, engineering alternatives, safety concerns, or economic considerations.

Evaluations that change the des.ign bases for SSCs important to safety, or the Bases of analyses, or conclusions stated in the SAR receive a 10 CFR 50.59 applicability review or safety evaluation.

NC.DE-PS.ZZ-0034(0), "Probabilistic Safety Assessment (PSA)" program is considered for use to assess the potential risk impacts.

SC.DE-AP.ZZ-0019(0), "Design Classification of SSCs" provides technical criteria for determining the design classification of SSCs. Any modification, change, or addition to an SSC is evaluated to determine whether it is required to perform a safety related function. Potential failure modes of the SSC are considered in process of this evaluation and must be documented in a design 88 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter

• Dated October 9, 1996 classification. The classification of SSCs is available to site personnel on Managed Maintenance Information System (MMIS).

Calculations prepared or revised independent of a DCP are reviewed for impact on station procedures and other design documents, and are reviewed for impact on the Safety Analysis Report (SAR). Refer to NC.NA-AP.ZZ-0059(0), "10 CFR 50.59, Applicability Reviews and Safety Evaluations" (NAP-59).

Calculations, drawings, Engineering Evaluations, component classifications, and equipment specifications are readily available through the site computer system.

These engineering process controls provide PSE&G reasonable assurance that SSCs are operated and maintained consistent with the design bases.

Ill. Validation Efforts This portion of the response focuses on programs, processes, and reviews that provide added assurance that SSC configuration and performance at SGS are consistent with the design bases.

A. Salem Generating Station (SGS) Process for Achieving Restart The Salem Restart Plan (SRP) describes and controls the activities needed to fundamentally improve the performance of the SGS and to support the controlled restart and sustained reliable operation of the Salem units. Salem Units 1 and 2 were removed from service by PSE&G on May 16, 1995, and June 7, 1995, respectively.

Prior to the unit shutdowns, PSE&G initiated a comprehensive assessment and a

improvement effort. PSE&G had developed tool for the overall management of improvement actions in the Nuclear Business Unit (NBU) called the IMPACT (Issue Management and Prioritized Action) Plan. This plan was based on an extensive analysis of recent internal and external performance reviews, audits, assessments, and inspections of both SGS units specifically, as well as the NBU in general. The IMPACT Plan included Intervention Actions, Near-Term Action Plans, and Long Term Actions.

Following the unit shutdowns, the IMPACT plan was reassessed to review its applicability to the shutdown status of the units and to ensure it addressed the underlying causal factors contributing to the decline in performance at Salem. It was then restructured into the SRP. The SRP consists of a comprehensive and systematic approach for the identification, review, approval, implementation

• assessment, and affirmation of activities needed to support the controlled restart and sustained reliable operation of the Salem units. This comprehensive plan 89 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 includes activities and process that provides reasonable assurance that plant SSC configuration and performance at SGS are consistent with the design bases and, as such, serves as part of our rationale in this area.

The Salem Restart Process (SRP) was submitted to the NRG on Nov. 24, 1995, letter LR-N95211. The SRP consists of five primary stages: Issue Discovery; Corrective Actions; Testing; Recommendation; and Concurrence.* Each of these stages, in turn, consists of several discrete processes or actions, some of which help provide reasonable assurance that plant SSC configuration and performance at SGS are consistent with the design bases.

In summary, as a result of the established process, PSE&G will be able to confidently affirm readiness to restart the SGS to the Chief Nuclear Officer and request NRG agreement to restart. At this point, it will have reasonable assurance that SSC configuration and performance at SGS are consistent with the design bases. Specifically, PSE&G also believes with reasonable assurance that upon completion of these restart efforts:

• "as built" configuration of plant SSCs are reviewed to be consistent with design bases requirements;

• Functional characteristics of the plant are established through testing and surveillance activities; and

• Processes to control installation and maintenance are defined and appropriately implemented.

B. Salem Configuration Walkdowns Another important review effort at SGS, by which PSE&G has developed reasonable assurance that SSC configuration and performance conform to design bases, is configuration walkdowns. Salem Configuration Walkdown Program processes and results are important to PSE&G. Their processes and recommendations began as a plant configuration baseline and review effort.

PSE&G believes that, the combination of these processes, the Design Bases Review I FSAR project, the System Readiness Review Process, .and the SGS Process for Achieving Restart provides reasonable assurance that Salem's SSC configuration and performance are consistent with the design bases.

Background

Configuration Control is a system of document management that allows a station's physical and operational condition to be known. Configuration Control

• Related (CCR) issues have been identified and documented by the Incident Report (IR) process. Special Inspection Team (SIT) Inspection Report 95-80 90 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 addressed the issue. In short, the team found that PSE&G had not taken the initiative, based on the IR trending information, to make an in-plant engineering scope assessment of the problems. However the IRs were generated as a result of an increased awareness and ownership of the material condition of the plants.

Thus, the IRs brought into question the process of how configuration control was maintained at Salem.

In June of 1995, the Collegial Self-Assessment (CSA) Team began an assessment of the Configuration Control Related Incident Reports as outlined in SA-95-013, "Root Cause Assessment of Configuration Control Related Incident Reports." The assessment was completed by Design Engineering in June 1995.

The purpose was to determine the causes of CCR issues that have been identified over the past year at SGS via Incident Reports (IRs). The CSA also identified specific and generic corrective actions that would correct deficiencies and preclude the reoccurrence of CCR issues. The objectives of the CSA Team assessment was to:

• Determine the apparent cause for the identified deficiency in the sample of CCR IRs;

• Determine the root causes of CCR issues that had been identified at Salem from the IRs; and

• Identify specific and generic corrective actions that will correct identified deficiencies and preclude CCR issues from recurring.

The method for completing the CSA objectives involved reviewing CRR documents, plant walkdowns, personnel interviews, categorization of CCR items, and assessments.

Based on the recommendations contained in the CSA report, a corrective action plan was developed. This plan was incorporated into the then NBU Impact Plan and later rolled into the Salem Restart Plan to include the performance of -

configuration control walkdowns and the initiation of corrective actions to resolve discrepancies.

Salem Configuration Walkdown As stated above, the Salem Configuration Walkdown Program was initiated on recommendations from CSA SA-95-013, "Root Cause Assessment of Configuration Control Related Incident Reports." The Salem Configuration Walkdowns were performed in accordance with PSPP5.2, "Conducting System Configuration Walkdowns." It was initiated to review as-built piping, -

91 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 instrumentation, and electrical configurations against plant piping and instrumentation drawings (P&IDs), electrical schematics, and one-line diagrams.

• Forty six (46) systems were selected for review based upon the systems selected by the Salem Readiness Program. The walkdowns were finally completed in August 1996.

• The concurrent effort for disposition and closure of backlogged CCR IRs, newly generated CCR IR/CRs, and CCR CRs generated as a result of he configuration walkdown effort:

1. Dispositioned greater than 2000 IR/CRs covering about 4700 configuration issues.

2. Closed greater than 1460 IR/CRs covering about 2750 configuration issues.

3. Over 400 Design Change Packages (DCPs) and 230 Corrective Maintenance Action Requests (CMARs) were generated to close the more than 1460 CRs. In the majority of cases, the configuration was found to be correct or acceptable, and the DCP was generated to correct the documentation. In two cases, neither the configuration nor the documentation were deemed correct and the DCPs provided remediation. With regard to the CMARs, minor corrections to bring plant configuration into conformance with.design documents were needed, such as adding pipe caps or hose connections.

• The following products were delivered by the Configuration Walkdown Program:

1. A set of marked-up P&IDs was provided for each Salem unit, identifying omissions and discrepancies.

2. A system summary sheet for the affected systems, at both Salem units, were developed to support the required formal processing and disposition of noted discrepancies.

3. Tags were placed on components requiring maintenance in accordance with procedure NC.NA-AP.ZZ-0009(Q).

4. Condition Reports (CRs) were initiated in accordance with procedure NC.NA-AP.ZZ-0006(Q) to track resolution of discrepancies. Over 870 CRs covering approximately 2600 configuration issues were .

identified during the course of the program.

The CSA and the Salem Configuration Walkdown Program processes and results are important to PSE&G. Their processes and recommendations led to the plant configuration review effort. PSE&G believes with the culmination of 92 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter

• Dated October 9, 1996 these processes, including the Design Bases Review I FSAR project, the System Readiness Review Process, and the SGS Process for Achieving Restart, provide reasonable assurance that Salem's:

• "As built" configuration of plant SSCs are reviewed to be consistent with design bases requirements;

• Functional characteristics of plant SSCs are established through testing and surveillance activities; and

• Processes to control installation and maintenance are defined and appropriately implemented.

C. Salem Generating Station (SGS) System Readiness Review Process Several issues associated with system and equipment reliability contributed to the decline in performance at SGS, including ineffective system engineering processes, lack of system ownership, and recurring equipment problems. In response to these issues, PSE&G implemented the System Readiness Review Program (SRRP). This program is a comprehensive and systematic process, separate and distinct from the work control process, that provides reasonable assurance that plant systems are evaluated, modified, and maintained, consistent with the design bases, to support restart. It also provides PSE&G with reasonable assurance that equipment operability and reliability issues are identified (including design bases fidelity) and that effective and timely corrective actions have been taken to address these issues.

The program also re-establishes an effective system engineering* process that restores ownership. It establishes a disciplined approach to system assessment, ownership, and accountability through the System Manager and system teams.

Using current processes and a team approach, it willbe evident through the presentation of the following phases that PSE&G has increased performance to effectively negate recurring equipment problems and ensure that SSC configuration and performance are consistent with the design bases.

As part of the SRRP, System Managers have systematically assessed forty six select critical systems and are affirming the readiness of other supporting systems. Some of the SRRP actions include processes that provide PSE&G reasonable assurance that SSC configuration and performance at SGS are consistent with the design bases. The critical systems were selected on the bases of a combination of attributes such as:

• Safe shutdown risk; 93 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Risk significance; Dated October 9, 1996 Historical power production impacts; and

• High historical corrective maintenance.

The SRRP, as discussed in the "Response to Question B", was developed as part of the System Engineering and Equipment Reliability Restart Action Plan to provide a process for evaluating plant systems' readiness to support safe and reliable operation.

The System Engineering Department Directive, SC.SE-DD.ZZ-0001 (Z) provides guidelines which assist the System Managers in researching, assessing, and documenting the readiness of critical plant systems.

The System Readiness Review Program evaluated open corrective action items against the Restart Screening Criteria. The restart screening criteria categorized items in two levels (Level 1 and Level 2) for restart applicability, if neither criteria applied, then the action item was not applicable to the restart.

Level 1 action items resolved a safety or operability issue and required resolution prior to plant startup. Level 2 actions items included issues that: 1) eliminate a component failure, deficiency or condition that could result in operation or entry into an LCO, or 2) restores licensing basis deficiencies to conforming conditions, or 3) corrects design bases deficiencies, i.e., deficiencies in safety related equipment or other TS equipment not in conformance with design bases documents such as UFSAR, or 4) resolves conditions that have resulted in repetitive safety system or power block equipment failures, and other criteria as described in procedure SC.SE-DD.ZZ-0001 (Z).

There are an additional forty-two supporting systems beirig reviewed using a thorough, but slightly less rigorous, process described in SC.SE-DD.ZZ-0002(Z).

The system readiness reviews involve:

• Identification of problems and corrective actions;

• Monitoring of corrective action implementation;

• Review of corrective action completion; and

• Continued monitoring of system performance.

The System Engineering Department Directive, SC.SE-DD.ZZ-0001 (Z) establishes the key attributes for PSE&G's performance criteria .. It is presented in a general outline in the following four phases.

94 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Phase I. the Initial Svstem Readiness Reviews OSRR.)

Upon shutdown of the Salem units, system teams were formed to conduct ISRRs. The ISRR included the following:

• A general review of system design bases documentation (e.g., Updated Final Safety Analysis Report (UFSAR), Technical Specifications, SER, Generic Letters, Pending LCR's);

1. The design bases review was to gain a comprehensive and consistent understanding of the system functional requirements for readiness.

• Evaluation of open action items against the Restart Screening Criteria:

1. Existing open corrective actions were screened to the restart criteria. The screening enables the return of the Salem units in a safe and reliable manner. Open corrective actions were also placed in the System Index Database (SIDs) for tracking.

2. New open corrective action items identified during the system readiness reviews were incorporated within the respective station process. Also, as necessary, an operability statement was included, per NC.NA-AP.ZZ-0006 and an open corrective work item requiring a Design Change Package per NC.NA-AP.ZZ-0008 had the appropriate process documentation initiated.

• Comprehensive system walkdowns were performed by multi-disciplined team and the System Manager;

1. The walkdowns created new corrective actions.

2. The walkdown open corrective action findings were reviewed, entered into appropriate NAP process, screened to the restart criteria, and recorded in SIDs.

3. A walkdown report was assembled by the System Manager. The report discussed the following: walkdown strategy; walkdown findings; corrective actions identified, and; a summary of system concerns and conditions.

• The System Readiness Review Report was developed by the System Readiness Review Team and attained the following:

1. Documented any identified Design/License Bases concerns;

2. Determination of Restart Workscope, Power Ascension Workscope, Post-Restart Workscope .and Canceled Action Items;

3. System Walkdown Report;

4. Aggregate evaluation of deferred work;

95. Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996

5. Identification of generic programs, process, and procedural concerns; and

6. Determination regarding new actions.

• Presentation of System Readiness Review (SRR) Reports to the System Readiness Review Board (SRRB):

1. The System Managers initiated Action Requests for new corrective action items that were identified during the system walkdowns. The items were evaluated and scoped within the SIDs database.

2. The System Manager presented the SRR Report to the SRRB. The SRRB approved, approved with comments, or disapproved the SRR Report.

3. If the SRR Report was approved, the System Manager presented the report to the MRC. If the SRR Report was approved with comments, the comments were resolved by the System Manager and the Engineering Supervisors reviewed and authorized the comment resolution. Once resolved, the System Manager presented the report to the MRC. If the SRRB disapproves the report, the System Manager re-presents the report to the SRRB until approval is obtained.

• The System Managers presented the System Readiness Review (SRR)

• Reports to the Management Review Committee (MRC). The MRC was established as an oversight committee to review the Restart Action Plans, System Readiness Reviews, the work scope of the outage, and other issues having the potential to impact restart. The MRC is chaired by the General Manager -- Salem Operations. MRC functions are described in its Charter:

1. The MRC approved or disapproved the SRR Reports.

2. If approved, the scope was included in the restart effort. If disapproved, the System Manager took appropriate actions and re-presented the SRR Report to the MRC.

• Phase II. the Restart Activities Monitoring (RAM.)

Phase II performed team follow-up on identified restart activities, scoping new open actions, and performing root cause analyses. These activities also include addressing SSC configuration and performance issues. This action provides reasonable assurance that SGS remains consistent with its design bases. The System Managers developed performance indicators based on SIDs to monitor approved system work scope and identify any new action items. Work activities were frequently field verified and monitored to insure corrective actions were being taken and no adverse plant conditions developed. The performance indicators used during Phase 11 reflected the percentage of completed restart 96 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 items within SIDs. Other actions taken as a part of Phase II include the following:

• Reviewed emergent action items for restart;

1. New work activities were screened similar to the evaluations of open action items and presented to the SRRB with relevant issues presented to the MRC for approval.

• Restart work within the appropriate System Work Window (to include Test Plans) was scheduled by the System Managers using the Desktop Guide Department Directive SC.TE-DD.ZZ-0004(z):

1. The System Managers using status reports and SIDs ensured that the system teams were aware of Restart Task Requirements to support the System Work Window Schedule.

• System walkdowns;

1. System walkdowns continued to occur during this phase. Members of the SRR team performed walkdowns on their systems.

• Established system priorities and performance indicators;

1. Performance indicators were developed based on SIDs to monitor approved system work scope and identify new action items; and

• Updated the SIDs database.

• Reviewed Phase II results for Restart Work Scope conducted by the Outage Scope Management Team, the System Readiness Review Board, and/or the Management Review Committee.

In review, Phase II included the formation of system teams, the review of open corrective action items, the evaluation of emergent action -items for unit restart, scheduling restart work within the appropriate System Work Window, examining field work, developing Startup System Test Plans, preparation for System Window closure and the FSRRs, the completion of regularly scheduled walkdowns by the System Manager, and updating of the SIDs database.

Phase Ill. the Final Svstem Readiness Review A FSRR was performed primarily to assess the aggregate impact of remaining work, and. to assess the remaining work that must be completed prior to restart.

A Final System Readiness Review Report was produced detailing results and identifying the remaining open items that must be completed prior to system affirmation.

97. Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 It is important to note that, within the scope of Phase Ill, PSE&G completed a Final System Readiness Review UFSAR Macro-review. The Macro-review assures the accuracy of the UFSAR to support the Salem Unit restart. The macro-views were conducted in accordance with S2.SE-DD.ZZ-0008(z), "System Engineering Final System Readiness Review UFSAR Macro-review Desk Guide". The effort entailed review of the 47 UFSAR systems identified for Maintenance Rule implementation and covered systems in the readiness reviews. For each system, the UFSAR chapter was reviewed for primary (macro) system parameters (such as, pressure, temperature, flow) and attributes (such as, single failure proof, seismic classification). These parameters or attributes were then compared with input to Chapter 15, Technical Specifications, other UFSAR chapters, and configuration drawings or documents. On average, approximately 40 to 45 attributes or parameters were reviewed per system for a total of roughly 2000. A separate report was provided for the systems that were reviewed, along with a listing of parameters or attributes established, and any deficiencies or discrepancies identified. Where deficiencies or discrepancies were uncovered, an Action Request (AR) was written.

The majority of the AR's involved minor inconsistencies or older information not thoroughly updated over the years of operation. Only about 3% of roughly 2000 system parameters or attributes reviewed during this program resulted in ARs.

Only 3 of which resulted in Level 2 AR's. This indicates that a very large percentage of information in the UFSAR is accurate. The high percentage of accurate UFSAR information provides reasonable assurance that SGS, Unit 2, will be operated in conformance with its design bases.

Other actions taken as a part of Phase Ill included the following:

• Completion of System Readiness Walkdowns by the System Manager and a representative of the Operations Department;

• Presentation of the Final System Readiness Review (FSRR) Report to the System Readiness Review Board and the Management Review Committee; and

1. The UFSAR Macro-review was included in the FSRR Report.

2. The FSRR Report includes the following but not limited to: Design Bases documentation that was reviewed to support the FSRR, such as, FSAR Sections, applicable Technical Specifications; SER, applicable NRC Generic Letters, Commitment Documentation, NRC Regulatory Guides and 98 Question C

SALEM .GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 CBDs; SIDs Reports; FSRR System Walkdown Reports; Restart Work Deferrals.

• Review of System Startup Test Plans.

In review, Phase Ill includes the System Readiness Walkdown that were completed by the System Manager and an assigned Operations representative, the presentation of the FSRR, prior to restart, to the SRRB and MRC, the review of work that was not completed prior to restart (i.e., review of the aggregate impact of deferred work), the initiation of the final review of the Startup System Test Plans, and the Macro-review of the appropriate sections of the UFSAR.

Phase IV. Startup and Power Ascension Upon final system affirmation, the SRRP returns the completed systems back to Operations. This action initiates the normal work control processes, and startLIP and power ascension testing. Final system affirmation includes the review of :,

Technical Specification action logs with Operations, identification of any open*

items required for restart, and completion of a comprehensive system team walkdown. Specific actions taken as a part of Phase IV include the following:

Reviewing designated test results identified by the Test Review Board (TRB);

Completing regularly scheduled system walkdowns; Completing the final review of System Startup Test Plans; Conducting oversight of Phase IV results by the SRRB and the MRC; and .

I I

• Completing System Readiness Affirmations. The System Manager assembles a System Readiness Affirmation Report, as documented by SC.SA-AP.ZZ-0035(Q), that reviewed the following: *

1. Current system License Change Requests.

2. Current operator workarounds.

3. Current corrective actions required of control room equipment (CR/Cls)

4. Currently installed T-Mods ..

5. Active operability determinations and JC O's, with the 10 CFR 50.59 for each operability determinations or JCO.

6. The remaining open "Exceptions" items and any "Exceptions" identified during the affirmation process.

7. Operations "Burden List" for the system.

8. Reference the FSRR Report.

99. Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996

9. Reference the system tests per the System Test Plan.

In review, Phase IV activities includes the review of test results identified by the TRB, system walkdowns by the System Managers, monitoring of any new corrective maintenance activities or condition reports by the System Manager, the completion of the final review of System Test Plans and completion of the System Readiness Affirmation.

• Discussion As of this writing Phases I through Ill of the System Readiness Review Process are substantially complete. Restart workscope was identified and the impact of open work items is being compiled and dispositioned. We believe this self-assessment demonstrates aggressive identification of a past programmatic engineering weakness. The aggressive identification of programmatic engineering weakness, and their subsequent resolution, provides PSE&G with reasonable assurance that SGS configuration and performance are consistent with the design bases.

The SRRP performed critical tasks that established current baseline information on the selected systems. The critical systems were reviewed against their design and also include a UFSAR Macro-review. The system readiness team through the System Manager tracked and maintained system status using SIDs, to include DCPs, current work activity, emergent work, operator workarounds, Operability Determinations, T-Mods, etc. In order to maintain system status, numerous system walkdowns tracked status as well as the SIDs database.

To date, work activities of the Salem Unit 2 Generating Station and the SRRP encompass the NAP processes that also maintained current system status as well as its design bases. While project activities focused on Unit 2, the vast majority of items are common to both units and will be addressed within the Salem Unit 1 SRRP (the UFSAR is common as are station procedures). Hence, general results and conclusions are applicable to both Salem units. From the culmination of these activities, PSE&G has reasonable assurance that equipment operability and reliability problems have been identified in a disciplined, systematic, and comprehensive manner that included design bases reviews, and that effective corrective actions have been taken to assure safe and dependable start-up and operation of the plant. As a result of the SRRP, PSE&G believes with reasonable assurance that Salem's performance programs are in place to maintain configuration control at a level sufficient to demonstrate that plant physical and functional characteristics are consistent with and are being maintained in accordance with their design bases.

100 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 D. Design Bases Review I FSAR Project The results of self assessments and an inspection of the UFSAR conducted by the NRC in the summer of 1996 (Report No. 96-80) indicated that there have been problems at the SGS with maintaining design and licensing bases documentation consistent with current plant configuration and operation. On June 11, 1996, PSE&G presented the FSAR Project Plan to NRC Region I.

Following a series of questions, formally responded to by PSE&G in letter LR-N96243 dated August 23, 1996, and a public meeting held at Salem station on July 2, 1996, the Region I staff indicated that the FSAR Project was structured to meet the reasonable assurance objective. The FSAR Project reasonable assurance objective, and the Salem improvement programs and processes, provides PSE&G with reasonable assurance that SGS operates within its design bases.

The FSAR Project, initiated in June 1996, combined a series of new initiatives with several recently completed or Salem improvement programs, that are in progress, related to the plant licensing and design bases. The FSAR Project was initiated to establish reasonable assurance that, upon restart, the licensing and design bases would be accurate and the station would be maintained and operated within those bases when taken together with the new initiatives and the process initiatives currently in progress. The project was completed in August 1996, and was closed by the Management Review Committee (MRC) in September 1996. While project activities focused on Unit 2, the vast majority of items are common to both units (the UFSAR is common as are station procedures). Hence, general results and conclusions are applicable to both Salem units. The FSAR Project involved integration of several new initiatives with various programs already in progress. The combination of new initiatives with the additional programs addressed the broad spectrum of design and licensing bases issues.

The new initiatives (discussed in more detail below) were conducted over a two month period, with a staff of approximately 50 engineers. These programs received substantial oversight and monitoring throughout the period including independent technical experts who reviewed the initial plan, other independent technical experts who monitored in-progress reviews and technical activities, Quality Assurance/Nuclear Safety Review (QA/NOR) surveillances, and numerous NRC inspections. At the end of the FSAR Project, an NRC team of three inspectors (from Region I and NRR) spent one week reviewing program results and interviewing engineering and operating staff personnel. Several weeks later, the NRC (Region I) initiated a full-scale safety system functional inspection (SSFI) of the component cooling water system. This inspection was conducted over a four week period by a team of six NRC engineers and contractors.

10:1 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 In the course of conducting the new initiatives, approximately 200 action requests were initiated. Approximately 30 items were marked for completion prior to restart of Unit 2, with the remainder being minor in nature and scheduled for post-restart correction. Three Action Requests eventually became Level 1 condition reports. These items are described below. These items have either been subsequently corrected or are presently work-in-progress:

1. Response time of containment fan cooler units (CFCUs) not in accordance with Technical Specifications;

2. Testing of auxiliary building and fuel handling building charcoal filters not in accordance with standards; and

3. Non-seismic piping not properly isolated from the refuelingwater storage*

tank. *

• The FSAR Project was approved for closure by the Salem Management Review Committee (MRC) on September 23, 1996. The MRC concurred that reasonable assurance of operation within the licensing and design bases had been provided since:

1. Chapter 15 accident analysis review had identified no major deficiencies;

2. Vertical-slice reviews provided favorable results in non-ventilation related systems; and

3. The UFSAR macro reviews had identified primarily minor errors or inconsistencies.

The additional programs were closed by MRC as separate activities. In addition, the NRC's Safety System Functional Inspection (SSFI) of the Component Cooling Water (CCW) system identified additional items for correction prior to restart. Upon completion of this inspection, the NRC inspection team noted that substantial improvements had been made to the CCW system over the past several years from a variety of programs, projects, or activities. Results of this independent assessment tend to corroborate PSE&G's conclusion from the FSAR Project, (i.e., PSE&G has reasonable assurances that SSC configuration and performance at SGS are consistent with the design bases) had been .

provided, with the possible exception of a CCW pump runout issue that was being technically resolved at the time of writing this response. Preliminary results indicate CCW pump runout is not a technical concern.

In summary the FSAR Project combined a series of new initiatives with Salem improvement programs that were directed at providing reasonable assurance that the Salem units would be operated and maintained within the design bases.

102 Question C

SALEM GENERATING S:TATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 These initiatives, when combined with process improvements provide reasonable assurance the plant will be operated within its licensing and design bases. Having established current "baseline" stat~s of the plant design bases through evaluations and upgrades and, PSE&Gs improved plant processes are now in place to maintain the design bases.

The initiatives of the FSAR Project have reviewed commitments against the UFSAR, Technical Specifications, Chapter 15 safety analyses, and plant engineering and design documents. The actual plant configuration and performance processes are also substantiated by walkdowns, EOP upgrades, Maintenance Rule implementation, and the TSSIP program that supports PSE&Gs rational for concluding with reasonable assurance that the configuration is consistent with the design bases. PSE&G has concluded with reasonable assurance that the plant configuration is known and adequate, based on the various activities described above, as well as the system readiness reviews described separately. PSE&G also concludes, with reasonable assurance, that performance processes maintain the design bases.

New Initiatives The new initiatives that reviewed the design and licensing bases against configuration documentation included:

Validation of accident analyses (UFSAR Chapter 15) input and assumptions.

• This review encompassed the relevant system and component inputs and assumptions (or parameters) that underlie the accident analyses found in .:.;_

Chapter 15 of the UFSAR. The parameters were reviewed against the appropriate calculation, surveillance test, or other supporting design bases documents. Where the parameters could be reviewed, they were documented in a matrix with their associated reference. Where they could not be reviewed, an Action Request was initiated.

• The Chapter 15 safety analysis review demonstrated a high degree of consistency and accuracy in the input and assumptions associated with systems and components. During the Chapter 15 Safety Analysis Review alone, slightly more than 400 inputs and assumptions were reviewed, approximately 87% were determined to be without deficiencies. Of the remaining 13%, calculations were in the process of being finalized (from a recent control room ventilation modification) or had to be obtained from the NSSS Vendor. As of writing this letter, only four NSSS Vendor calculations remain outstanding.

103 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Vertical Slice Reviews of safety systems.

• As part of the UFSAR Project, vertical slice reviews were conducted for seven safety-significant systems. Vertical slices were originally conducted for four additional systems as a pilot project. These pilot reviews were smaller in scope and were performed in less depth than the seven reviews conducted under this project. Hence, the FSAR Project does not take credit for the first four reviews except to the extent that they add some additional review and therefore additional credit to the overall effort, and add assurance that SSC configuration and performance conforms to the design bases. The seven systems for vertical slice review are shown in the table below and can be separated into two basic categories:

Non-Ventilation Systems Ventilation Systems

1. Fuel Pool Cooling 1. Auxiliary Building Ventilation

2. Safety Injection 2. Fuel Handlino Buildino Ventilation

3. Reactor Protection System 3. Containment Building Ventilation

4. Miscellaneous Ventilation Systems (Intake Structures, Emergency Diesel Generators, Switchgear &

Penetration)

• The vertical slice approach utilizes a multi-disciplinary team that systematically starts with a review of licensing bases information (e.g.,

UFSAR, licensing correspondence) and then evaluates the fidelity of that information to associated design bases (e.g., calculations, drawings, design packages) and the implementation or operating bases (e.g., operating and emergency procedures, testing, safety evaluations) documents. Where potential problems were identified, the team went further into review of support systems or .other interfacing areas (e.g., Motor Operated Valves (MOVs), fire protection). This technique has been used successfully by the NRC in major team inspections such as Safety System Functional Inspections.

• In view of the above technique, vertical slice reviews constitute discrete evaluations of specific programs, such as design change packages (DCPs) and design calculations. For example, in the course of the various vertical slice reviews, a total of 59 DCPs were reviewed and evaluated, including evaluations of supporting documents, resulting FSAR changes, and 10 CFR 50.59 safety evaluations. The primary focus of the DCP reviews was to determine if the DCPs adequately preserved the design and licensing basis of the system. These reviews indicates that PSE&G has reasonable 104 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 assurance that SGS configuration and performance processes are consistent with the design bases.

Macro-reviews of 46 UFSAR system descriptions.

• UFSAR Macro-reviews were previously covered in "Response to Question C" under SGS System Readiness Review Process, Phase Ill.

Reviews of completed Deficiency Reports (DEFs).

• The DEFs were previously covered in "Response to Question B" under FSAR Project.

Reviews of Engineering Evaluations (EE) that supported past Justifications for Continued Operation (JCO).

• The EEs and JCOs were previously covered in "Response to Question B" under FSAR Project.

Review of NRC Safety Evaluation Reports (SERs) Associated with Technical Specifications

• A review of SERs associated with NRC license amendments issued through May 1996 were reviewed to assure that they have been correctly incorporated into the UFSAR. The review consisted of 181 SGS Unit 1 and 162 SGS Unit 2 license amendments. SERs associated with the 16 safety analysis systems of UFSAR Chapter 15 were completed prior to restart.

• As a result of these reviews, a total of 28 ARs were written. The level 3 and 4 ARs primarily involved minor errors in the UFSAR updates, often times failure to update a table or update all information that is repeated in the UFSAR.

• While the need for enhancement in the UFSAR update process was indicated, no Technical Specification (TS) compliance issues were noted.

The lack of deficiencies, in the review, indicate that the TS changes were correctly implemented. The lack of TS issues and the minor nature of the failures to fully update the remainder of the FSAR are sufficient to provide PSE&G with reasonable assurance that SGS configuration and performance are consistent with the design bases.

Additional Initiatives Several additional initiatives which were occurring at the time of the FSAR Project contribute to the reasonable assurance that SSC configuration and performance are consistent with the design bases. While it is likely that some 105 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 discrepancies will remain to be identified as these programs continue, the scope of completed efforts and the results achieved provide the assurance that, SSC configuration and performance are consistent with the design bases, and that safe operation at SGS will not be impacted. The initiatives that provided the link between design documents and the as-built plant configuration are listed below.

Configuration Walk-downs Configuration walkdowns are conducted by System Managers as part of System Readiness Reviews. In addition, the Nuclear Engineering Design organization recognized the need to address Configuration Control Related (CCR) Incident Reports. As a result, that organization initiated configuration control walkdowns (previously discussed above) of systems considered important to safety and to reliable operation. From the outset, the configuration control walkdown project was aligned with the System Readiness Program. At the end of the system walkdown, a conformance meeting was held. During the meeting, the walkdown team reviewed their results with the System Manager and his Design Engineering counterpart. These conformance meetings assured alignment on CCR issue resolution. Although the primary focus of the walkdown effort was to verify the as-built piping, instrumentation and electrical configuration against plant Piping and Instrument Drawings (P&IDs) and Electrical Schematics and/or One-Line diagrams, supplemental efforts addressed specific configuration issues previously noted in Incident Reports (e.g., sampling pipe flange and bolting material). Final closure and acceptance of CCR issues will be made prior to the restart in accordance with the SRP for the Salem units. This provides PSE&G with reasonable assurance that SSC configuration and performance are consistent with the design bases.

Maintenance Rule Implementation This initiative involves the review of design and licensing bases information at Salem Generating Station. The scope of the Maintenance Rule, 10 CFR 50.65, is based upon system function(s) with regard to safe shutdown and 10 CFR Part 100 considerations. Risk significance is then determined on the bases of applicable Core Damage Frequency. The system functions, and associated risk significance, are verified by the respective System Managers, as well as reviewed and approved by an Expert Panel comprised of qualified Operations, Engineering, Maintenance, and PSA members.

Experience throughout the nuclear industry demonstrates that there is a clear link between effective maintenance and safety as it relates to such factors as the number of transients and challenges to safety system and associated need for operability, availability and reliability of safety equipment. Maintenance is also important to ensure that design assumptions and margins in the original design 106 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 bases are maintained and are not unacceptably degraded. Therefore, nuclear power plant maintenance is clearly important in protecting the public health and safety. Thus, the Maintenance Rule requires the monitoring of the overall continuing effectiveness of our maintenance programs. This ensures that:

1. Safety related and certain non-safety related SSCs are capable of performing their intended functions; and

2. For Non-safety Related equipment,

- Failures will not occur that prevent the fulfillment of safety related functions, and .

- Failures resulting in scrams and unnecessary actuations of safety related systems are minimized.

Both ris~ significant criteria and performance criteria were established to provide a standard to measure the performance of SS Cs. Risk significant SS Cs were *..

established using an Expert Panel and insights developed by plant specific PSA analysis. The PSA techniques included risk reduction worth, risk achievement worth, and contribution to core damage frequency. PSA concepts have been used to develop a risk matrix. The risk matrix clearly shows those combinations of equipment that, if removed from service simultaneously, will affect core damage frequency.

The Maintenance Rule System Function and Risk Significant Guide SE.MR.SA.01 provides an additional dimension affording us a greater margin of safety while operating within our normal design bases limits. lh addition to PSA use as a part of the Maintenance Rule application, the PSA Methodology Applications Report, dated December 12, 1996 functions to implement PSA into integrated plant operation. The purpose of Salem's Probabilistic Safety Assessments (PSA) application is to develop an assessment methodology for planning on-line maintenance not necessarily related to just the Maintenance Rule. This is accomplished by creating a plant specific risk matrix applicable to operating MODES 1 and 2. This provides an index of plant safety when one or more safety significant components are to be removed from service. The assessment relies on calculating the increase in core damage frequency (CDF) when one or more plant components become unavailable and unable to perform its intended function. By comparing the change in CDF when key components are available versus when they are unavailable aids in analyzing the safety significance of a particular maintenance evolution.

In effect, we analyze the unique, integral plant configuration, as determined by the maintenance activity, to requantify our model of the plant. This new dynamic 107 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 assessment tool aids in our process methods and provides PSE&G with reasonable assurance that Salem's performance programs maintain configuration control at a level sufficient to demonstrate that plant physical and functional characteristics are consistent with and are being maintained in accordance with their design bases.

The Technical Specification Surveillance Improvement Program (TSSIP)

As discussed in "Response to Question B", Unit 2 Licensee Event Report (LER) 311/95-008, dated January 15, 1996, reported a condition in which a Technical Specification (TS) surveillance requirement was not fully proceduralized and several prior instances of ineffective TS surveillance implementation, including one case whereby a requirement was not adequately proceduralized. The LER included a commitment to implement the TSSIP as an integral part of the Salem Operations Restart Action Plan. PSE&G initiated the Salem TSSIP to evaluate the quality of the TS surveillance program. TSSIP is designed to review the surveillance testing program to enhance administrative controls, assure proper scheduling and tracking of surveillances, and review the implementing procedures, with certain exceptions, for TS surveillance testing requirements.

For these reasons, the TSSIP project provides PSE&G reasonable assurance that SGS configuration and performance are consistent with the design bases.

• Emergency Operating Procedure (EOP) Upgrade The project, as discussed in "Response to Question B" of this response, was specifically organized in January 1996 to direct, and manage a review I upgrade based on recognized weaknesses in the SGS EOPs. The review specifically incorporated lessons learned from NRC EOP inspections (e.g., NUREG-1358) and industry experience (e.g., INP0-83-004, INP0-83-006). The EOP verification and validation has been completed, and EOPs have been upgraded and successfully utilized during simulator scenarios with crews during restart training. The EOP Upgrade Project, combined with the EOP maintenance function contributes to reasonable assurance that Salem operating procedures are consistent with the plant design bases (i.e., assumptions related to the UFSAR Chapter 15 accident analysis such as setpoints and operator response times), as discussed in "Response to Question B".

System Readiness Review Program Provides guidance in formatting, evaluating and determining system readiness for Salem restart. The SRRP was previously discussed as a part of "Response to Question C" in this section.

108 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 The System Restart Test Program.

The program was implemented through Phases Ill and IV of the SRRP. The Salem Restart Plan requires the establishment of a Startup and Power Ascension Program. Procedure SC.TE-Tl.ZZ-0001 (Q) defines the Startup and Power Ascension Program as consisting of component testing, system testing, integrated functional testing and power ascension testing to support the restart of the SGS. Procedure SC.TE-Tl.ZZ-0001 (Q) also provides guidance for the startup testing of critical systems defined in SC.SE-DD.ZZ-0001 (Z). Procedure SC.TE-Tl.ZZ-0001 (Q) is also applicable to the closure of system windows and the testing involved to demonstrate adequate functionality of the equipment and system as defined in SC.TE-DD.ZZ-0004(Q). The System Restart Test Program actions preserve SGS configuration, through its processes and provide PSE&G with reasonable assurance that SGS configuration and performance are consistent with the design bases.

Configuration Baseline Document (CBD) Project Principle Objectives of Initial CBD Project PSE&G developed a design bases review program, referred to as the Configuration Baseline Documentation (CBD) project. The principle objective of the CBD program was to develop, consolidate, and document the specific design bases of the company's nuclear power stations. The focus of this effort identified and documented "why" the plant and supporting systems are designed and constructed to specific technical standards, safety guidelines and/or specifications.

Program Description/Scope The CBDs were to be developed for each system or structure that was considered safety related, Technical Specification-related, or important to safety at the SGS. There were 44 CBDs developed for Salem. CBDs include the following elements: functional description of the system, applicable codes and standards, regulatory documents, system operation, system and component design, accident analysis as stated in the USFAR and references. The .

development of a CBD required the following actions: (1) data retrieval (includes Technical Specification, UFSAR, licensing correspondence and commitments),

(2) indexing of data, (3) comparison to as-built (identify design inputs and licensing commitments, identify design outputs, comparison of design requirements versus as-built condition, determine if design input criteria have been met, identify discrepancies), (4) draft document preparation, and (5) consultant peer review.

109* Question c

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Where discrepancies were identified between controlled documents, Discrepancy Evaluation Forms (DEFs) were generated. The DEF process included both operability and reportability reviews. Draft CBDs and associated DEFs were peer reviewed by the contract organization. The CBDs, DEFs, peer review comments, and comment resolutions were then submitted to PSE&G for review by the system specific project team that included representatives from Design and System Engineering. Once the comments were incorporated to the satisfaction of the project team, the final CBD was design verified by PSE&G and issued for use. A prioritization program, based on safety significance using Probabilistic Safety Assessment (PSA) techniques, (where PSA is the same as Probabilistic Risk Assessment), was used to categorize the DEF's for resolution schedules.

Initial CBD Verification and Validation Activities PSE&G instituted a process whereby a "vertical slice" review known as a Safety System Functional Review (SSFR) was conducted on selected systems to validate CBDs. As the CBDs were developed, checks were made to assure that the engineering data was consistent with the actual plant configuration. The configuration management program provided reasonable assurance that NRC requirements were met.

A procedure was developed to control the maintenance of the CBDs. Any change to the CBD that resulted from a Design Change Package (DCP) was included as part of the DCP. If the DCP required design verification, the CBD change would also receive design verification as part of the DCP.

However, CBDs may also be impacted by processes outside of the design change process (i.e., revisions to calculations, engineering evaluations, and review of SOERs). To include such changes, the CBDs were revised and submitted to the Sponsor Engineer (system representative from Design Engineering) for review and approved by a Functional Supervisor in the Design Engineering organization. This type of text change to the CBDs was incorporated without design verification. Therefore the CBDs could no longer be used as design input documents. This problem was identified in June of 1996.

In July of 1996 a letter was distributed to all engineering and licensing personnel discussing the proper application of the CBD's. The conclusion was that the CBDs should not be used as a design input source, but as a "road map" to the source documents. This approach was reinforced in a series of roll-out meetings conducted by the director of Design Engineering and Projects, as well as in the Engineering Support Program (ESP) training.

110 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Letter from E. C. Simpson to All Engineering Personnel Dated Julv 5. 1996 of Initial CBD Project In August 1992, the NRC issued a Commission policy statement "Availability and Adequacy of Design Bases Information at Nuclear Power Plants." This policy statement emphasized the importance of maintaining current and accessible design documentation. PSE&G believes, with reasonable assurance, that Salem's CBD performance process provides the availability and adequacy of design bases information at a level that provides reasonable assurance that plant physical and functional characteristics are consistent with and are being maintained in accordance with their design bases.

In addition to PSE&Gs performance process, Senior Management directives have provided expectations and direct guidance as demonstrated by the subject letter, quoting:

".... while the Configuration Baseline Documents (CBDs) for both Salem and Hope Creek are not design bases documents (i.e., single source of design bases information), they provide. us a road map to find and understand the design bases. Within the Engineering Assurance organization, these documents are undergoing a review to ensure that the design bases are correctly reflected in our plant operating documents."

IV. Assessment of Effectiveness A. INDEPENDENT ASSESSMENTS This part of the response focuses on independent assessments, and how they contribute to, or reinforce reasonable assurance that SSC configuration and performance at SGS are consistent with the design bases.

• NRG Special Team Inspection (ST!) Report 96-80 Between May and October of 1996, NRC Region I conducted Special Team Inspection (STI) No. 96-80 at SGS. It found examples in the control air system, fuel handling ventilation system, and containment cooling system of Salem Unit 2 that indicated the design and licensing bases were not well understood and connected. To a large degree, these findings, when considered with other self-identified examples of design and licensing bases concerns, led directly to initiation of the FSAR Project. The FSAR Project plan, that was presented to Region I on June 11, 1996, contained features specifically intended to address concerns identified during inspection 96-80, such as the UFSAR Macro-reviews, the DEF/DES closure reviews, and the vertical slice reviews. For example, two of the vertical slices during the FSAR Project reviewed the fuel handling 111 . Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 ventilation system and containment cooling system specifically because of problems identified by the NRC during inspection 96-80.

The FSAR Project was initiated several weeks after the NRC completed the STI.

Various levels of oversight activities were imposed on the FSAR reviews, including a team of independent industry experts who monitored the FSAR Project on two separate occasions and a number of NRC visits, both in-process visits and inspections following completion. As a result of these oversight activities, the NRC provided the following comments in the executive summary of the 96-80 inspection report:

• An FSAR Project conducted in July-August 1996 was well managed and represented an effective effort to review the Salem Updated Final Safety Analysis Report (UFSAR). The project had strong elements of oversight and independent confirmation.

• In the FSAR Project, over 400 input assumptions for UFSAR Chapter 15 safety analyses were reviewed by reference to calculations, test procedures and other design documents. With exception of 9 instances involving incomplete source documents and some relatively minor discrepancies, the Chapter 15 reviews demonstrated a high degree of consistency and accuracy, providing assurance of operation within the bounds of the safety analyses.

In summary, the STI findings tend to substantiate the FSAR Project. PSE&G concluded from the FSAR Project that the corrective action processes were adequately addressing deficiencies based on:

• The relatively large project sample and few AR problems found;

• The importance of the problems, which were primarily related to documentation; and

• The fact that these issues related to process concerns more reflective of the 1988-93 time frame, rather than post-1995.

The FSAR Project combined a series of new initiatives with Salem improvement programs that were directed at reviewing the accuracy of design bases. These initiatives combine with process improvements provide reasonable assurance the plant will be operated within its design bases. Having established current "baseline" status of the plant design bases through evaluations and upgrades and, PSE&Gs performance processes such as 10 CFR 50.59 Safety Evaluations, Corrective Action Program, UFSAR Maintenance, Self Assessments provides reasonable assurance that SSC configuration and performance are consistent with and maintain the design bases.

112 Question c

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter

• Dated October 9, 1996 Special Inspection Team (SIT) Report Nos. 50-272195-80 and 50-311I95-80.

dated June 9. 1995 On January 12, 1995, the NRC presented PSE&G's Systematic Assessment of Licensee Performance (SALP) for both Salem units. As a result of the SALP assessment, PSE&G developed performance improvement plans and in March 1995, chartered an organizational assessment team to critically review Salem activities and the improvement plans.

• In April 1995, NRC Region I determined that an SIT inspection should be conducted to assess how effectively PSE&G was performing from a safety perspective. The SIT assessment was a comprehensive evaluation of SGS performance. The SIT assessed how effectively PSE&G performed from a safety perspective in the conduct of day-to-day activities.

The NCR's concern is that the SIT team found areas of poor performance such as configuration control. In view of the previously discussed FSAR Project, combined with a series of new initiatives and Salem improvement programs, now provides reasonable assurance the plant will be operated within its design bases. Having established a current "baseline" status for SGS configuration and performance, PSE&Gs performance processes such as 10 CFR 50.59 Safety Evaluations, Corrective Action Program, UFSAR Maintenance, Self Assessments, provides reasonable assurance that the design bases is maintained.

CCW System Safety System Functional Inspection (SSFI). 1996 The NRC completed a SSFI at SGS Unit 2, examining the Component Cooling*

Water (CCW) system. The SSFI was conducted to independently assess the scope, depth and quality of the FSAR Project. The NRC's CCW system SSFI was the most recent, and most comprehensive design and design bases review performed by the NRC at SGS. It was initiated specifically to confirm or challenge the SGS FSAR Project reasonable-assurance conclusions. Several technical issues were identified, and are being addressed through ARs, prior to plant restart (such as issues associated with pump runout, room cooler single failures, and technical specification interpretations).

The SSFI team asked PSE&G over 200 questions, many with multiple parts, such that 400 to 500 questions were ultimately asked and resolved. The fact that several questions (1 Oto 15) remained to be resolved tends to substantiate our FSAR Project reasonable assurance conclusion, particularly when the two major issues which the NRC identified as potentially challenging the capability of the system have been preliminarily resolved.

113 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 In view of the above, the SSFI results tended to substantiate conclusions from the FSAR Project. Many deficiencies identified by the NRC during the SSFI were corrected as restart requirements, with the remainder scheduled for correction post-restart. Nevertheless, the improving trends observed by both the FSAR Project and the NRC SSFI indicate that Salem's current programs and procedures have been effective. Therefore, PSE&G has reasonable assurance that SSC configuration and performance at SGS.are consistent with the design bases.

Salem SERTISSFR Closure Review Project Report. NSRAS-96-10 The assessment was initiated to ensure that the corrective actions identified in the Salem Significant Event Response Team (SERT) and Safety System Functional Reviews (SSFRs) have been addressed or are being considered as part of the Salem Restart effort. The review included the Salem SERT reports SERT 90-1 through 95-02 (a total of 38 SERTs) and all of the Salem SSFRs (a total of 7 SSRFs.)

• Of the 655 SERT recommendations reviewed, 612, (93%) were categorized as CLOSED and 43, (7%) were categorized as OPEN.

• Of the 497 SSFR recommendations reviewed, 372, (75%) were categorized as CLOSED and 125 (25%) were categorized as OPEN.

The assessment revealed that the bases for closure documented in ATS for many SERT and SSFR recommendations were inconclusive and in some instances non-existent. Action Requests were generated to document and bring these concerns to closure. Issues from the SERT and SSFR recommendations that were categorized as OPEN were also documented to bring them to closure.

The review established reasonable assurance that those recommendations, categorized as CLOSED, meet one of the following criteria:

• The issue is no longer valid; *

• The issue has been addressed; or

• The issue is being addressed by the Restart Plan or some other tracked activity.

Therefore, as evident above, the independent assessments provide PSE&G reasonable assurance that SSC configuration and performance at SGS is maintained consistent with its design bases.. .

• 114 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter

• Dated October 9, 1996 OSR Review of the 10 CFR 50.59 Safety Evaluation Program. dated May 17.

1996 The 50.59 program is one of PSE&G's performance processes that maintains SGS configuration and performance consistent with the design bases. The program is initiated by procedure NAP-59 and it is the Licensing Department's responsibility. However, 10 CFR 50.59 Safety Evaluations (50.59 SEs) are initiated by several other programs I processes. These processes are controlled by individual procedures outside the responsibility of Licensing. In May of 1995, SGS Units 1 and 2 received a Level IV violation (NRG IR No. 272/95-06 and 311/95-07) for an inadequate Applicability Review and an inadequate 50.59 SE.

The program failed to provide adequate bases that a USQ was not involved. The NRC's Restart Action Plan, dated February 23, 1996, did not specifically identify 50.59 SEs however it is addressed within Programmatic Restart Issue Item Ill, No. 11, "Engineering Problem Resolution, Including Safety Evaluations."

In November of 1995, the Offsite Safety Review (OSR) conducted an independent assessment of the accuracy of 10 CFR 50.59 Applicability Reviews to support Salem restart. The review covered a sampling of 48 documents. The assessment identified the following:

• Total Sample Size of 48 Documents 46 Documents Description of Finding Were found to correctly conclude that 10 CFR 50.59 did not apply.

2 Documents Incorrectly concluded that 10 CFR 50.59 did not apply.

The above two documents that incorrectly concluded that 10 CFR 50.59 did not apply were related to T-Mods. The T-Mods changed the facility as described in the SAR so a Safety Evaluation was required. PSE&G understands the importance of the 10 CFR 50.59 programs as a key element for maintaining configuration and performance issues within the design bases. The assessment concluded that although weaknesses were identified, no important safety issues were found.

The OSR conducted an independent assessment, NSRAO 96-18 dated May 17, 1996, of the adequacy of the 50.59 SE program to support Salem restart. The assessment focused on procedures and controls, program implementation, training and qualification. OSR concluded that the 50.59 Program is adequate to support Salem restart. This conclusion was based upon previous OSR assessments and assessments performed by both Licensing and Engineering (prior to May 17, 1996). However, since the time assessment NSRAO 96-18 115 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 was performed, a potential USQ has been identified. In January of 1997, the potential USQ was identified by PSE&G's OSR group during the review of the 50.59 SE. The 50.59 SE was performed to support a request to revise Chapter 15 of the UFSAR with a new analysis. The 50.59 SE should have addressed an apparent deviation between the proposed change, and the Basis of the Technical Specification regarding the Pressurizer PORVs. The USQ disposition is currently being evaluated as a Level 2 Action Request. A Technical Specification Amendent to resolve this issue has been forwarded to the NRC.

Assessments and reviews, such as the above discussions, provides PSE&G additional recommendations that will continue to strengthen the program.

Although deficiencies continue to be found they will be documented and evaluated in accordance with the corrective action program. PSE&G believes with reasonable assurance that the 50.59 SE Program meets the requirements of 10 CFR 50.59, and that SSC configuration and performance at SGS are consistent with the design bases.

\I. Improvements Implemented or Planned The Nuclear Engineering Department recently implemented or reinforced several actions to provide additional assurance that configuration and performance of SSCs are maintained in accordance the design bases at Salem.

• Require a line by line detailed review of the DCP by the Peer Reviewer

• Schedule the appropriate time required by the Peer Reviewer to perform the required detailed review

• Document deviations from standard practices, vendor recommendations and document acceptability of specified equipment

• Require Peer Review by Installation and Test Group for testing sections of the DCP .

• Require conceptual and constructability walkdowns and document in design analysis

• Require Installation and Test Engineering to walkdown the modification prior to installation

• Include Station Maintenance, Operations, Station Planning and the DCP Installer to review the change package .

• Implement a training and qualification program for Modification Engineers

• Perform third party reviews Additionally a Design Change Process Review is occurring as part of overall process improvement. This review will again benchmark the process, and recommend procedure changes to improve the program. The process reviews 116 Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter

• in 1997.

VI. Conclusion Dated October 9, 1996 are underway, and recommendations to streamline are planned for incorporation In view of the foregoing information in this part of the response, PSE&G believes that reasonable assurance exists such that SSC configuration and performance are consistent with the SGS design bases as further summarized below:

• Plant configuration was originally licensed by the NRC consistent with the design bases;

• Processes are in place that ensures the consistency between the configuration I performance and the design bases;

• These processes are effective while the results of performance-based assessments and readiness programs indicate minor deficiencies remain to be corrected, important safety items have not been identified to date. The primary assessments or programs which support this conclusion, all completed in 1996, are:

1. The FSAR Project

2. The Salem Restart Test Program .

3. The Configuration Walkdown Process

4. The Salem System Readiness Review Process

• The results of the NRC's Component Cooling Water System SSFI, conducted in December 1996, tended to support the above conclusion. Many .

I I

deficiencies identified by the NRC during the SSFI were corrected as restart I requirements, with the remainder scheduled for correction post-restart.

Nevertheless, the improving trends observed by both the FSAR Project and the NRC SSFI indicate that Salem's current programs and procedures have been effective. Therefore, PSE&G has reasonable assurance that SSC configuration and performance at SGS are consistent with the design bases.

• The Salem Corrective Action Program, along with other configuration control processes, were demonstrated to be effective during the above-mentioned assessments or programs in maintaining configuration in accordance with the design bases..

• In addition, to provide added assurance in this area, beyond the reasonable assurance that has been provided in this response for Salem, activities will be conducted to further review the plant is configured and operated in accordance with the design bases. These commitments are discussed in Attachment 2 of this response.

117. Question C

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 RESPONSE TO QUESTION D Provide information on the processes for identification of problems and implementation of corrective actions, including actions to determine the extent of problems, action to prevent recurrence, and reporting to NRC.

I. Corrective Action Program A. Salem Action Request Process The Salem site uses the Action Request Process to collect, track and control all types of problems for resolution. This process encompasses nonconformances in plant design and problems in plant design documents. For conditions adverse to quality, the Action Request Process provides for evaluation of the problem.

Resolution of the problem and implementation of any corrective actions is contained in the Corrective Action Program part of the process. Problems can be identified from numerous sources, both internal and external. This process allows all employees and contractors the ability to enter any problems for evaluation and response.

This section provides a brief overview of the operation of the Action Request Process and the Corrective Action Program including conditions adverse to quality, operability determinations, reportability determinations and action closure. A summary of how the Operating Experience Program adds external information to the process is included. The important role of the Employee Concerns Program for identifying internal issues for entry into the Action Request Process is also described. In addition, in Section V, information is provided on the effectiveness of PSE&G's implementation of these programs.

The process has evolved substantially from the spring of 1995, after various program weaknesses were identified by both internal and external assessments.

These program weaknesses included: reporting threshold was too high, root cause analyses did not get to the root cause or failed to correct the root cause, there was insufficient follow through on corrective actions, root cause/cause determinations and corrective actions were not completed in a timely manner, and corrective actions were not always effective at correcting problems. To correct these weaknesses, a new process was designed and new governing procedures were implemented. A root cause manual was developed and training was provided for root cause investigators. Section C contains recent inspection results on the Corrective Action Program. These inspections generally discuss the improved process as providing the necessary capabilities.

118 Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 The Corrective Action Program uses evaluation managers to lead the evaluation of problems and to assign tasks for resolution as appropriate. The program is designed such that the responsible department manager retains ownership and accountability for resolution of the issues. Operability impact and reportability determinations are made within the evaluation process. The evaluation manager is responsible for assuring adequate and timely resolution of the problem.

Conditions adverse to quality require causal analysis to help ensure effective countermeasures are implemented. The type of causal analysis depends on the significance level of the condition adverse to quality ( Level 3 conditions require trending rather than causal analysis.). Recurrences of conditions adverse to quality result in higher action levels within the Corrective Action Program.

• Those problems requiring design or configuration changes for resolution are tracked by Salem site Engineering in the Design/Configuration Change TrackiQg System. This system tracks creation of the design change package, implementation and closure [described in "Response to Question A"]. All design changes require 10CFRS0.59 reviews and updating of the UFSAR as applicable.

8. Action Request Process Operation This Action Request (AR) Process description applies to the Nuclear Business Unit (NBU) which encompasses both Salem and Hope Creek Generating Stations. The AR Process provides the method for reporting and resolving conditions adverse to quality as defined in 10CFRSO, Appendix B, Criteria XVI, and business process type enhancements, thus providing a single point of entry for items requiring corrective action. Coding is provided to separate conditions adverse to quality from business process items in the following manner:

• Corrective Maintenance (CM) action requests are used for reporting equipment related conditions adverse to quality

• Condition Resolution (CR) action requests are used for reporting programmatic conditions adverse to quality other than hardware (i.e.,

administrative control, procedural or human error events)

• Business Process (BP) action requests are used to report business process enhancements, requests for information or support, ~tc.

• 119. Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 The AR Process is described in procedure NC.NA-AP.ZZ-0000(0), "Action Request Process". This procedure provides the directions necessary for conditions requiring immediate action, initiation, review, approval, management review, and closure. The three AR types are controlled by the following procedures:

• Corrective Maintenance (CM) type: NC.NA-AP.ZZ-0009(0), "Work Control Process"

• Condition Resolution (CR) type: NC.NA-AP.ZZ-0006(0), "Corrective Action Program"

• Business Process (BP) type: NC.NA-BP.ZZ-OOOO(Z), "Business Processing of Action Requests" During the initiation phase of the AR process, individuals document the condition by providing a description of the condition. In addition, the process calls for an impact statement, identification of regulatory reporting requirements that may apply, suspected cause or source of condition, corrective actions previously taken, recommended corrective action and recommended evaluation manager.

Significance levels are determined by the initiator and reassessed by reviewers and approvers to establish the importance of the action.

While the major source of actions identified come from employees and contractors during the normal course of their day-to-day work, other programs such as the Operating Experience Program, 10CFR Part 21 reports and the Employee Concerns Program provide actions to be processed in the AR Process.

C. Conditions Adverse to Quality The Corrective Action Program [NC.NA-AP.ZZ-0006(0)] provides the instructions and guidance to ensure conditions adverse to quality are dispositioned and corrected in compliance with 10CFRSO, Appendix B, Criterion XV, (Nonconforming Materials, Parts, or Components) and Criterion XVI, (Corrective Action). The program applies to Corrective Maintenance (CM) and Condition Resolution (CR) type Action Requests initiated in accordance with

. NC.NA-AP.ZZ-0000(0) Action Request Process. Evaluations are required by procedure and are conducted to determine the extent of the condition and to determine corrective actions as follows:

120 Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter

• Dated October 9, .1996 Significance level 1 - A root cause investigation and a root cause analysis report (RCAR) is completed in accordance with DTG-CAP-003, "Root Cause Manual."

The RCAR is expected to include a detailed root cause and causal factors analysis, recommended corrective actions, and follow-up actions to verify the effectiveness of the corrective actions.

Significance level 2 - An apparent cause evaluation is conducted in accordance with the Root Cause Manual. The evaluation is expected to include an apparent cause and causal factors analysis, and any corrective actions to correct the condition.

Significance level 3 - The Evaluation Manager may either specify and initiate actions to correct the condition, document the evaluation, or assign trend code(s) and close the action if no open tasks remain.

Significance level 4 - These tasks have been evaluated to not be conditions adverse to quality and are no longer a level in the Corrective Action Program.

• They are now in the BP or Business Process part of the Action Request Process.

Trending of problems and the implementation of the Maintenance Rule requirements have been incorporated into the Corrective Action Program:

D. Operability Determinations The reviewer of an AR is expected to immediately notify the SRO approver if an operability concern exists. The SRO approver is then responsible to immediately notify the on-duty SNSS/NSS if a condition adverse to quality exists requiring an operability determination. Operability determinations are performed in accordance with Operability Determination procedures and documented. If required, engineering support may be requested to perform an in-depth follow-up assessment of operability and provide documentation.

Operations Management presents Significance Level 1, 2 and 3 conditions requiring follow-up operability assessment to the station management team for consideration of the following:

• Designation of the Evaluation Manager or Responsible Department. *

• Concurrence with or change to the significance level.

• Designation of any short term actions and requirements.

• Identification of any special considerations.

• Authorization of a team investigation for significance level 1 conditions or other conditions as deemed appropriate.

121 Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Upon being informed of an operability issue, the on-duty SNSS/NSS determines and directs any immediate actions needed to place the plant in a safe condition.

The results of the initial operability screening for the affected Structure, System, or Component (SSC) are documented in accordance with Operations Department procedures.

The operability determination process is the element of the Action Request process whereby adverse conditions are evaluated for impact on the plant design bases relative to the ability of SSC's to perform their intended safety function. This provides a mechanism for the Operations Departments to maintain cognizance of adverse conditions with the potential to impact plant safety, and thereby implement any restrictions on plant operation or compensatory actions to maintain the plant in a safe condition.

E. Reportability 10CFR50.72 & 10CFR50.73 The Senior Reactor Operator (SRO) approver has the responsibility to determine the effect of conditions adverse to quality on structures, systems, and components (SSC). If the condition identified requires 10CFR50.72 notification, the SRO approver immediately contacts the on-duty Senior Nuclear Shift Supervisor/Nuclear Shift Supervisor (SNSS/NSS). Licensing is contacted for assistance in determining reportability.

The Nuclear Licensing and Regulation Department reviews the reportability of conditions adverse to quality according to state and federal regulations. When a 10CFR50.73 reportable event is identified, Licensing notifies the evaluation manager responsible for resolving the condition and obtains information for required report preparation. If it is determined that a notification was not made within the required time, the licensing representative would contact the SNSS/NSS to make the required notification.

F. Action Closure Corrective actions are tracked to completion/implementation and closure actions are documented by the assigned Evaluation Manager. The organization assigned the implementation of the associated corrective action(s) is required to maintain the activity in the "active status" until the corrective action is implemented and documentation of completion is provided.

122 Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter

• Dated October 9, 1996 When all evaluations, corrective actions, verifications, and all other activities are implemented and completed, the Evaluation Manager verifies that actions specified in the evaluation have been implemented and the condition has either been resolved or new activities are created to correct the remaining conditions, as required. The Evaluation Manager assures codes are assigned according to the Root Cause Manual to assist with the identification of adverse trends. When all activities are implemented and documented as "complete," the Evaluation Manager then places the action in the approved status. The Evaluation Manager also ensures hard copy documentation is forwarded for appropriate record keeping.

G. Training and Responsibility Identification General Employee Training for access to the Salem site includes a section on the Corrective Action Program. The training identifies the responsibility of every employee to identify conditions adverse to quality and to report these conditions through the Action Reqwest Process. The electronic entry and manual form method of entry are described. Also, the process of contacting the Senior Nuclear Shift Supervisor for operability issues is described. Employees are also told that additional information and training is available from the Corrective Action Group. Training is available from the Corrective Action Group upon request.

The NBU Work Standards Handbook also reinforces the Corrective Action Program by stating that every employee is responsible and has the authority to identify conditions adverse to quality and to report these conditions through the Corrective Action Program. Management expectations are constantly reinforced during indicator review meetings and during problem identification discussions.

II. Operating Experience Program The Salem Operating Experience Program is used to analyze industry operating experience in order to implement timely actions to prevent or reduce the consequences of similar occurrences at this site. [Procedure NC.NA-AP.ZZ-0054 (Q), "Operating Experience (OE) Program"]

Documents containing operating experience information from outside the Salem site are screened for applicability to the Salem site by the OE Program. The following documents are included in the screening:

1. INPO Significant Operating Experience Reports(SOERs)

2. INPO Significant Event Reports (SERs)

3. INPO Significant Event Notifications (SENs) 123. Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996

4. INPO Significant by Others (SOs)

5. INPO Operations and Maintenance Reminders (O&MRs)

6. NRC Bulletins and NRC Generic Letters: These important documents within the OE Program have provided information on design problems.

Many of these documents have resulted in the implementation of design improvements at Salem and Hope Creek. Most of these documents have required responses back to the NRC describing the design improvement actions taken.

7. NRC Information Notices

8. Vendor Reports (including 10CFR Part 21 reports)

9. Industry Operating Experience Reports (OEs)

10. Industry Plant Status Reports (PSs)

11. Any other documents containing information appropriate to the OE Program

12. Nuclear Plant Reliability Data System (NPRDS)

Problems identified as applicable to the Salem site are documented and entered into the Action Request process for resolution. The corrective action database is then used to track the status of evaluations and corrective actions.

OE Program effectiveness is monitored by means of indicators. Periodic self-assessments are conducted to assure continued program effectiveness and improvements.

Ill. 10CFR PART 21 Process Conditions adverse to quality that are provided to PSE&G under 10CFR21 and conditions adverse to quality that may require reporting by PSE&G under 10CFR21 are entered into the Corrective Action Program. The Corrective Action Program provides the process to determine operability impact and reportability requirements. The Corrective Action Program tracks the issue until the necessary corrective actions are completed.

124 Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 The conditions adverse to quality provided to PSE&G under 10CFR21.21 (b) are received by our Materials Center and forwarded to the Operating Experience Program for action assignment in the Corrective Action Program. Vendor documents that could result in a 10CFR21 reporting requirement for PSE&G are also entered into the Corrective Action Program by the Operating Experience Program to assign evaluation responsibilities. Conditions adverse to qL1ality, identified at PSE&G, that have potential reporting responsibility under 10CFR21 are entered into the Corrective Action Program by the identifying department.

Those conditions adverse to quality that could have reporting requirements under 10CFR21 have to have Form 1 of procedure NC.NA-AP.ZZ-0035(0),

Nuclear Licensing and Reporting completed, to provide establishment of reporting requirements.

IV. Employee Concerns Program The Salem site employee concerns program can be a source for design bases issues and configuration control issues. When such issues arise, and evaluations have to be performed to determine the need for corrective actions, the actions are entered into the Action Request Process. The employee concerns program provides another way of assuring design information and configuration control is maintained.

V. Effectiveness A. Program Requirements The Corrective Action Program at Salem has programmatic requirements for ensuring that conditions adverse to quality are corrected and that those corrective measures preclude recurrence. The program requires that an effectiveness review be performed for any significance level 1 condition adverse to quality after all evaluations, activities, and corrective actions are complete.

Effectiveness review plans are required to be presented to the Corrective Action Review Board (CARS) during the root cause presentation for all significance level 1 ARs and are subsequently required to return to CARB for validation of effectiveness results prior to the issue being closed. An effectiveness review is expected to assure the following:

1. All corrective actions are complete

2. The condition and causes(s) were corrected and have remained corrected

3. That no additional actions are required

4. That the specified corrective actions did not create any new conditions 125 Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Corrective Action Program effectiveness is ensured and monitored for all Condition Resolutions (CRs) by the following:

1. Program Performance Indicators

2. Evaluation Manager's review of completed activities

3. Corrective Action Review Board for root cause evaluations (significance level 1 issues)

4. Corrective Action Review Committee for apparent cause evaluations (significance level 2 issues) *

5. Program Internal Self Assessments

6. Internal Audit Programs Corrective Action Program performance indicators include components in the areas of timeliness, schedule adherence, and quality. Timeliness indicators include average age, average time to complete evaluations, average time to complete tasks, and total actions open. Schedule adherence tracks task completion to the schedule. Quality assessments are done by the Corrective Action Review Board and Corrective Action Review Committee on the level 1 and level 2 causal evaluations.

B. Results The current Corrective Action Program and the Action Tracking System (previous corrective action program) contain numerous examples of design and design bases issues that were correctly addressed and the design documentation updated including the UFSAR. Both tracking systems require all actions to be completed before an issue is closed. The following table shows examples of issues which were evaluating design bases questions that resulted in the submittal of LERs and have either design bases corrections or design changes in progress or completed:

126 Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 DESIGN BASES REVIEWS RESULTING IN LERs LER DATE ISSUE DESCRIPTION NUMBER 272/96-017 7/18/95 Failure of Control Room Air System modifications Conditioning to meet GDC are in progress via DCP 19 criteria. # 1EC-3505 including design bases documentation updates.

272/95-006 4/5/95 Control Room Air System modifications Conditioning would not are in progress along have automatically with design bases actuated to the emergency documentation updates.

recirculation mode in response to a high radiation signal due to 2R 1B monitor not being verified unblocked prior to blocking 2R1A.

• These examples of previously identified problems and the implemented corrective actions provide confidence that our corrective action process is effective in meeting Appendix B criteria.

C. Inspection Results This section contains some of the more significant statements from recent Salem inspections and reports from the NRC and our QA Department regarding the Corrective Action Program and related areas. The Action Request Process is an NBU process that is central to both Hope Creek and Salem. Recent inspections of the process have occurred at Hope Creek and are discussed in the Hope Creek submittal, but the results apply equally to Salem. These inspections indicate that we do have the appropriate process installed. Some of these inspections did identify implementation weaknesses which are being addressed under the response effort to the individual inspections. However, the inspections do show that the overall process effectiveness is sound.

NRC Integrated Inspection Report 50-272/96-07, 50-311/96-07 07.2 Operating Experience Feedback (OEF) Program, NRC Restart Inspection Item 11.9 (Open); Dated July 19, 1996 127 Question D

r' SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 "c.

Conclusions:

The Salem staff improved the effectiveness of the OEF program through better performance monitoring, improved screening, enhanced communication with the training organization, and revisions to the OEF procedure. The performance monitoring resulted in greater accountability for implementing corrective actions. The inspector noted, however, that the OEF staff could not furnish evidence that plant staff met all of the responsibilities established in the OEF procedure."

Section 3.3.6 Operating Experience Review "c.

Conclusions:

The team concluded that the licensee was conducting effective reviews of industry experience and that applicable information was being forwarded to the responsible division for corrective action and training of department personnel, as appropriate."

Quality Assessment Department Audit Report 96-190-2, Corrective Action Audit - Performed October 30 through November 26, 1996 Executive Summary "Effectiveness Statement: The audit team determined that the Corrective Action Program is adequate and supportive of the continued operation of Hope Creek and Salem operating mode changes. Areas of improved performance were noted; however, the audit identified ten (10) areas of deficiencies that will require further refinement and or management attention to improve performance.

Increased emphasis for the implementation of the program is needed particularly for the performance of the evaluation manager's role and responsibilities to ensure appropriate corrective action items are identified and implemented satisfactorily."

"Improved Performance: The audit team continued to see improvement in problem identification as evidenced by the continuing number of action requests initiated. The QA corrective action cultural survey continues to demonstrate an increased understanding of when and how to initiate an action request and management support for the process."

"The use of dedicated corrective action groups within the station's departments continues to demonstrate improved focus for implementation of the process and timeliness."

128 Question D

SALEM GENERATING STATION

-Response to 10 CFR 50.54(f) letter Dated October 9, 1996 "The Corrective Action Program's control of documentation and transmittal for record retention has demonstrated improvement. The audit identified one instance where the Root Cause Analysis Report was not transmitted to the Corrective Action Group (CAG)."

"During the last QA Corrective Action Audit it was identified through a Maintenance Surveillance and the survey that people were reluctant to initiate Action Requests for issues related to human performance. This audit's survey attempted to clarify this issue through responses from the following:

• People are reluctant to initiate ARs.

• View themselves as being reluctant to initiate ARs.

• Aware of instances where people were reluctant to initiate ARs within the last three months.

• Knowledge of what actions were taken even though the individuals may have been reluctant to initiate the AR."

"The survey's responses demonstrated that the reluctance issue reported during the last QA audit was perceived rather than factual."

VI. Discussion Of Recent Findings While we believe reasonable assurance has been provided in "Response to Questions 8 and C", PSE&G recognizes that deficiencies in design bases consistency within Salem 1 and 2 plant configuration and procedures have recently been discovered. Some examples of recent findings are listed below.

Engineering personnel at Salem concluded that both of the units had the potential to operate the Containment Fan Coil Units (CFCU) outside the plant design bases (this addresses the industry issue of GL 96-06). Additionally, there was a potential for the Service Water System (SWS) strainer design to be outside of the design bases, and a failure of one of the two switchgear penetration exhaust fans may place Salem in an unanalyzed condition.

The issues described above are being aggressively addressed by the engineering staff and tracked through the corrective action process. The goal is to minimize these types of occurrences. However, if a problem does occur, it will be dealt with in accordance with Salem Technical Specifications and the Corrective Action Program.

129 Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 VII. Conclusion The Action Request Process is in place at Salem. The process is the focal point for many problem identification modes, and, based upon programmatic assessments and performance indicators, appears to be functioning effectively.

While the process is always evolving through continuous self-assessment and enhancements, the process provides confidence that when problems are identified, timely action is taken. The extent and cause of the problem is determined and evaluations for operability and reportability impacts are included. This is a key process that allows us to identify and resolve the thousands of actions required to assure the safe, reliable and cost effective operation of our nuclear power plants.

130 Question D

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 RESPONSE TO QUESTION E Provide an assessment of the overall effectiveness of the current processes and programs in concluding the plant configuration is consistent with the design bases.

The foregoing sections of this response provide descriptions of configuration control processes at Salem Generating Station, including the process for problem identification and corrective action implementation. These processes, along with the identified validation reviews and independent assessments which evaluated process effectiveness, provide PSE&G with reasonable assurance that both plant procedures and plant configuration are in conformance with Salem's design bases.

In evaluating the information provided in the "Responses to Question A through D", PSE&G believes sufficient information has been provided to conclude with reasonable assurance that, upon plant restart, the Salem Generating Station will be configured, operated, and maintained within its design bases. As additional issues are discovered through continuing review activities, they will be addressed and resolved in accordance with our corrective action program and plant Technical Specifications.

The above conclusion is based upon the following information:

1. The Salem Generating Station was designed in the early 1970s and was licensed by the NRC for operation in 1976 for Unit 1, and 1982 for Unit 2.

[Note: Industry standards for design bases and configuration control documentation have substantially changed since initial plant licensing.

Consequently, this letter begins with design basis improvement efforts of the early 1990s and does not address plant status prior to 1990.]

2. The Salem design was baselined and the design bases were compared to the as-built configuration during the 1991 to 1994 time frame, when 44 system and structure Configuration Baseline Documents (CBDs) were prepared and approved. This provides reasonable assurance that design bases information is adequate.

3. Deficiencies identified coincident with CBD preparation were resolved (with some minor items still being resolved) in accordance with the existing corrective action program. During the FSAR Project in 1996, some 500 closed Discrepancy Evaluation Forms (DEFs) were reviewed and, with few minor exceptions, closures were judged acceptable .

• 131 Question E

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996

4. The current processes used to maintain and control changes to the design have been upgraded and based upon recent assessments, have been shown to be adequate.

5. Substantial improvements were made during the current extended Salem outages in both plant procedures and installed plant configuration, such as upgrading emergency operating procedures, substantial reductions in engineering back-log items, upgrading of MOV and IST programs, and balancing building ventilation systems.

6. During the current extended outages, a large number of programs were initiated that provide us reasonable assurance that design bases information is adequate and has been adequately translated into plant procedures and configurations. These programs include the configuration walkdowns, system readiness review program, and the integrated restart test program.

7. In addition to the above, results of validation activities and independent assessments provide additional assurance that Salem Generating Station is configured and will be operated and maintained consistent with its design bases. The FSAR Project, which included a comprehensive review of safety *analysis input and assumptions, 7 system vertical-slice reviews, and 46 UFSAR system description reviews, was initiated specifically to document a basis for reasonable assurance of operation within the design bases.

8. The NRG safety system functional inspection (SSFI) of the component cooling water system in December 1996, tended to confirm PSE&G conclusions from the FSAR Project. Items with the potential to challenge FSAR Project conclusions were being resolved with the NRG as this letter was being prepared. Current analysis results indicate the FSAR Project's reasonable assurance conclusion Will not be impacted by the SSFI issues. Issues will be resolved with the NRG prior to Salem Unit 2 restart.

9. As additional issues are discovered through continuing review activities, they will be addressed and resolved in accordance with our corrective action program and plant Technical Specifications.

10. PSE&G has an effective corrective action program in place to identify and resolve design bases problems in a timely manner.

As part of the many upgrade activities undertaken in preparation for Salem restart, there have been many significant design related improvements that provide additional confidence in the plant design bases and design bases consistency. These include:

• Configuration walkdown of 46 systems 132 Question E

SALEM GENERATING STATION Response to 10 CFR 50.54(f) letter Dated October 9, 1996 Seven vertical slice assessments Validation of Chapter 15 accident analyses inputs and assumptions Ventilation systeni flow balances

• Cooling water system flow balances

• Correction of approximately 23,000 deficiencies or work items.

PSE&G has also implemented several initiatives to improve personnel performance relative to design bases issues and sensitivity to design bases issues. Also, as part of the Salem Restart Plan, a comprehensive start-up test

• program will be conducted that will further validate the adequacy of major design improvements (such as digital feedwater).

• In general, PSE&G believes the evidence indicates our corrective action program, document control program, and plant operating procedures are strong, I,*

but that UFSAR consistency can continue to be improved, CBDs should be updated to reflect recent plant changes, and engineering staff design b.ases knowledge can be improved. A program to continue design bases related improvements is outlined in Attachment 2 of this letter.

133 Question E

'1

• SALEM GENERATING STATION Response to 10 CFR 50.54(f} Letter Dated October 9, 1996 ATTACHMENT 2

n * . ""

SALEM GENERATING STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 DISCUSSION OF FUTURE ACTIONS I COMMITMENTS provides the rationale for our reasonable assurrance that design bases requirements are translated into the plant configuration and operating procedures for the Salem plants. This section provides a brief discussion of both ongoing and future actions that will be completed over the next two years to provide additional confirmation of compliance with design bases. The detailed plans for conducting these further assessments, including specific information on scope and schedule, will be provided to the NRC within 60 days of the date of this letter.

Ongoing Activities There are several ongoing initiatives discussed in this response that are in progress or will be completed (some of these may be rolled into the future reviews discussed below ). These include the following:

• Configuration Baseline Document (CBD) Validation Program.

• Technical Specification Surveillance Improvement Project, Phase II

• System Readiness Reviews and affirmations

• Startup Testing Program New Initiatives In addition to the completed and ongoing activities to provide reasonable assurance of compliance with design bases, PSE&G will embark on a comprehensive Design Basis Review project. This project will focus on areas of review not yet covered in previous assessments and weaknesses discussed in this response. These reviews will be prioritized based on the safety significance of the plant systems using the Maintenance Rule definitions. Systems will be categorized according to their risk significance (e.g., Safety Analysis, Risk Significant, Risk Important systems and Other systems). The initial focus will be on the risk significant and safety analysis systems. The scope of the project will be changed as appropriate based on findings (e.g., may shift the focus if appropriate). The desired end state of this project is expected to be a fully documented design/ licensing bases which have been validated through plant procedures and is easily accessible for use in day to day conduct of work. To the extent that they are not addressed by completed or ongoing activities, items 1 Attachment 2

SALEM GENERATING STATION Response to 10 CFR 50.54{f} Letter Dated October 9, 1996 that will be evaluated as part of this project to ensure consistency with the design bases include:

• Reviews of UFSAR versus plant configuration and operations

• Key SER reviews for consistency with UFSARS

• Regulatory Commitment review

• Setpoint limits

• Verify testing adequacy

• Reviews of plant procedures against design bases documentation (e.g.,

UFSAR).

• Configuration walkdowns

• Safety System Functional Review type Vertical Slice reviews for selected systems

• Confirmation of Salem FSAR project for Salem Unit 1

• Review of system lineups

• Drawing reviews The approach planned for these reviews is to maximize the use of PSE&G resources to ensure clear ownership and improve the understanding of design bases by PSE&G personnel. Multidisciplinary teams will be formed to perform a comprehensive review of design bases information on a system I structure basis.

Wherever possible, these teams will be structured to comprise the system manager, system design engineer, system senior reactor operator and other cross-disciplinary engineering support as may be required by the particular system (e.g., specialty engineering). The team will then (1) reconfirm or identify design bases requirements and related system attributes, (2) initiate changes to the CBD, as appropriate, including updating from the recent outage and (3) initiate UFSAR change requests (including the related safety evaluation) to clearly delineate both the design bases requirements and system attributes which implement the design bases.

In conclusion, PSE&G plans to undertake a comprehensive Design I Licensing Bases Review project which will complement the work already completed to

• • provide further assurance that the Salem plants are operated in accordance with their design bases. The specific details relative to the scope and schedule of 2 Attachment 2

\,

SALEM GENERATING"STATION Response to 10 CFR 50.54(f) Letter Dated October 9, 1996 this project will be provided in a follow-up letter within 60 days from the date of this response.

3 Attachment 2