ML20134M158

Forwards Response to Request for Info Per 10CFR50.54(f) Re Adequacy & Availability of Design Bases Info

ML20134M158
Person / Time
Site:	Vermont Yankee
Issue date:	02/14/1997
From:	Reid D VERMONT YANKEE NUCLEAR POWER CORP.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
BVY-97-23, NUDOCS 9702200115
Download: ML20134M158 (78)

Text

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VERMONT YANKEE-

NUCLEAR POWER CORPORATION l .

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. Ferry Road, Brattleboro, VT 05301-7002 ENGINEERING OFFICE 580 MAIN STREET DOLTON. MA 01740 (508)779-6711 l

February 14,1997 BVY 97-23 L

United States Nuclear Regulatory Commission ATTN: Document Control Desk l Washington, DC 20555 l

References:

(a) License No. DPR-28 (Docket No. 50-271)

(b) Letter, USNRC to VYNPC, NVY 96-158, dated October 9,1996

Subject:

Vermont Yankee Response to Request for Information Pursuant to 10 CFR 50.54(f) Regarding Adequacy and Availability of Design Bases information The enclosure to this letter constitutes Vermont Yankee's response to the U.S. Nuclear Regulatory Commission's (NRC) request for information, pursuant to Section 182(a) of the Atomic Energy Act of 1954, as amended and 10 CFR 50.54(f), regarding the adequacy and availability of design bases information [ Reference (b)). The NRC request for information, dated October 9,1996, was received by Vermont Yankee on October 18,1996. This response regarding the adequacy and availability of the Vermont Yankee design bases information is current as of the date of this letter.

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Our reviews and assessments have confirmed to us that Vermont Yankee's overall performance in the areas of design and configuration controlis fundamentally sound. We  !

recognize that past activities have identified opportunities for improvement in these areas.

Findings identified in these areas, however, have not typically resulted in significant design bases information problems.

To improve our performance in the areas of problem identification and corrective action, we implemented the Event Report process in 1995. We have also taken measures to instill an aggressive, questioning attitude throughout alllevels of the Vermont Yankee organization.

These measures include encouraging an environment where personnel are supported by l management to identify, for formal evaluation and resolution, any issues which may concern them. In addition, we are presently taking further actions to enhance our Corrective Action Program based on insight received from the recent NRC Corrective Action Program inspection (NRC Inspection Report 50-271/96-200).

Notwithstanding our conclusion that we have reasonable assurance that the Vermont Yankee

! design bases have been adequately translated to the plant design and procedures, and that j the plant configuration is maintained in an appropriate manner, Vermont Yankee has committed to undertaking a series of actions designed to provide improved configuration management. These commitments, which are delineated in Section 10 of Enclosure 1, include i completion of the Design Basis Documentation Program, the improved Technical

(- Specifications Program, and an FSAR Verification Program. 1 I

9702200115 970214 L

/l PDR ADOCK 05000271 i P -PDR j

VERMONT YANKEE NUCLEAR POWER CORPORATION

! United States Nuclear Regulatory Commission l

February 14,1997 Page 2 of 2 l As a licensee, Vermont Yankee acknowledges its responsibility to operate and maintain the l Vermont Yankee Nuclear Power Station with the highest regard for safety. We recognize and t I understand the importance of maintaining sound design bases and implementing appropriate controls to operate within these bases through high standards of configuration control. Our l ongoing efforts to reinforce and continuously improve our safety culture, professionalism, self- l assessment and responsiveness are focused on meeting this responsibility.

We trust that this submittal provides the requested information, However, should you have questions or require additional information, please contact this office.  !

Sincerely, VERMONT YANKEE NUCLEAR POWER CORPORATION Donald A. Reid 4 l Vice President, Operations  !

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l c: Director, Office of Nuclear Reactor Regulation l USNRC Region i Administrator USNRC Project Manager- VYNPS l USNRC Resident inspector- VYNPS

m. .._

l k EQ o(>

STATE OF VERMONT ## @

WINDHAM COUNTY ) N01ARY n' Il k l Then personally appeared before me, Donald A. Reid, who being duly sworn, did state ghe ' ViMQ4@ent s ,,

l Operations, of Vermont Yankee Nuclear Power Corporation, that he is duly authorized gcu nd file the 39$regdin document in the name and on the behalf of Vermont Yankee Nuclear Power Corporatio that ientp[

l therein are true to the best of his knowledge and belief. q g {0UNY*q f L1/2 ,d if 5' ally A. Sa6ds'trum, Notary Public My Commission Expires February 10,1999

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ENCLOSURE 1 i.

VERMONT YANKEE NUCLEAR POWER STATION j.

RESPONSE TO

! NUCLEAR REGULATORY COMMISSION 3 50.54(f) LETTER DATED OCTOBER 9,1996 i

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TABLE OF CONTENTS Page 1.0 - EXECUTIVE

SUMMARY

. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2.0 I NTR O D U CTI ON . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 3.0 G ENE RAL IN FORMATION . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 3.1 Vermont Yankee Nuclear Power Station . . . . . . . . . . . . . . . . . . . . . . . . 5 3.2 Development of 50.54(f) Letter Response . . . . . . . . . . . . . . . . . . . . . . . 5 l 4.0 RESPONSE TO USNRC TOPIC (a) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 l

l 4.1 I nt rod u ction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 4.2 Design Change Process . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 4.3 10 CFR 50.59 Safety Screening and Safety Evaluation . . . . . . . . . . . . 10 1

l 4.3.1 Safety Screening and Evaluation . . . . . . . . . . . . . . . . . . . . . . . . 11 l 4.3.2 Screening and Safety Evaluation Review and Approval . . . . . . 11 j 4.4 Procedure Development and Revision . . . . . . . . . . . . . . . . . . . . . . . . . 12

l. 4.5 Minor Modifications . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 l 4.6 Setpoint Changes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 ,

! 4.7 Equivaiency Evaluations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 l 4.8 . Lineup Deviations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15  !

4.9 Corrective Update to Plant Drawings . . . . . . . . . . . . . . . . . . . . . . . . . . 16 l 4.10 FSAR Revisions and Periodic Operating Reports . . . . . . . . . . . . . . . . 16 l 5.0 RESPONSE TO USNRC TOPIC (b) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 l 5.1 Design Control and Procedure Change Processes . . . . . . . . . . . . . . . 19 l 5.2 Recent Program initiatives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 5.3 Audits, Assessments, and Inspections . . . . . . . . . . . . . . . . . . . . . . . . . 26 i 5.3.1 Vertical Slice Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26 5.3.2 Evaluation of USNRC Inspection Reports, Intemal VY Assessments, and ERs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 28  !

5.4 S u m m a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 9 6.0 RESPONSE TO USNRC TOPIC (c) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31

,. 6.1 Configuration ControI Processes . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 32 l l 6.2 Recent Program Improvements and Other Initiatives . . . . . . . . . . . . . . 33 l l 6.3 Walkdown and Modification Projects . . . . . . . . . . . . . . . . . . . . . . . . . . 42 j 6.4 ' Audits, Assessments, and Inspections . . . . . . . . . . . . . . . . . . . . . . . . . 44 6.4.1 Vertical Slice Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . 44 I

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TABLE OF CONTENTS (Continued)

P Page 6.4.2 50.54(f) Response-Specific Evaluations . . . . . . . . . . . . . . . . . 44 '

l 6.4.3 Evaluation of USNRC Inspection Reports and Internal VY l Asse ssments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 5 l

l 6.5 S u m m a ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 5 l 7.0 RESPONSE TO USNRC TOPIC (d) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47 ,

7.1 I nt rod uction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 47  !

! 7.2 Process for Identification of Problems . . . . . . . . . . . . . . . . . . . . . . . . . 47 l

7.2.1 Event identification and Processing . . . . . . . . . . . . . . . . . . . . . . 49 7.2.2 Events Screening Meeting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 49 7.2.3 Events Classification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 50 l l 7.3 Actions to Determine the Extent of Problems . . . . . . . . . . . . . . . . . . . . 52 I

7.3.1 Event investigation . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 7.3.2 ER Review and Approval . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 52 1 1

1 7.4 Implementation of Corrective Action . . . . . . . . . . . . . . . . . . . . . . . . . . . 53 i

7.5 Trending and Functional Area Assessment . . . . . . . . . . . . . . . . . . . . . 55 7.6 Reporting to USN RC . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 55 8.0 RESPONSE TO USNRC TOPIC (e) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 8.1 I nt rod uction . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57  :

I 8.2 Procedures and Programs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 l l

8.3 Recent Program initiatives . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 57 I 8.4 Review of Inspections, Audits, and Event Reports . . . . . . . . . . . . . . . . 57 i 8.5 Response Team Assessments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 58 l 8.6 Summary and Conclusion . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 59 9.0 RESPONSE TO USNRC REQUEST FOR OTHER INFORMATION . . . . . . . 61 9.1 Design Bases Documentation Program . . . . . . . . . . . . . . . . . . . . . . . . 61 9.2 ITS P roj e ct . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 63 10.0 C O M M ITM E NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 5 10.1 Design Bases Documentation Program . . . . . . . . . . . . . . . . . . . . . . . . 65 l 10.2 ITS P roj ect . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 5 l

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Page 10.3 FSAR Verification . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 65 10.4 Follow-Up Activities to this Response . . . . . . . . . . . . . . . . . . . . . . . . . 65 l

A P P E N D I X A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 66 Introduction to Appendix A . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 67 A.1 Review of Engineering Processes and Products . . . . . . . . . . . . . . . . . 68 A.2 Review of the Design Change Procedure . . . . . . . . . . . . . . . . . . . . . . 68 A.3 EDCR Closeout Documentation Review . . . . . . . . . . . . . . . . . . . . . . . 69 A.4 LE R Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 A.5 Corrective Action Review . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 70 AC R O N Y M S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 72 L

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iv Enclosure 1

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1.0 EXECUTIVE

SUMMARY

This document constitutes Vermont Yankee's response, pursuant to Section 182(a) of the Atomic Energy Act of 1954, as amended, and 10 CFR 50.54(f), to the U. S. Nuclear l Regulatory Commission ("USNRC" or "NRC") request for information regarding the adequacy and availability of design bases information at the Vermont Yankee Nuclear Power Station (VY).

The USNRC request identified five specific topics and included a request for certain other i information related to design review programs. In developing this response, VY has reviewed I

and evaluated existing documentation and has performed additional evaluations and investigations. As a result, it has been concluded that the programs and processes currently in l place provide a reasonable level of assurance that the plant configuration and performance are consistent with the design bases.

Sections 2.0 and 3.0 of the response provide background information on the USNRC request and VY, respectively. Section 4.0 describes the processes used to implement 10 CFR 50.59, 10 CFR 50.71(e), and 10 CFR 50, Appendix B. This discussion addresses the multiple, inter-related engineering design and configuration control processes which are used to control the translation of the plant design bases into the plant design and procedures. Foremost i among these processes are the Vermont Yankee Operational Quality Assurance Program l

(YOOAP-1-A), Engineering instruction WE-100 [ Engineering Design Change Requests], and Administrative Procedure AP-6002 [ Preparing 50.59 Evaluations and Final Safety Analysis Report (FSAR) Changes).

YOOAP-1-A, which is based on ANSI N45.2.11-1974, establishes the basic framework and guidelines within which design bases activities are performed at VY. VY procedures establish the mechanisms through which new design activities and/or proposea design changes are prepared, evaluated, reviewed and implemented. These procedures delineate the need to identify, review and consider existing design bases information prior to developing a change, how and by whom the proposed change is to be reviewed and approved, and how it is to be implemented, closed out and documented. The existence of these procedures reasonably assures that changes to the plant design will be handled in a formal and appropriate manner consistent with applicable regulatory and industry requirements.

Sections 5.0 and 6.0 detail VY's rationale for concluding that there is reasonable assurance that the processes in use effectively translate the design bases into the plant procedures and the configuration of systems, structures and components. This rationale is based on a number l of factors including:

• The original design bases were developed under the licensing requirements and l engineering practices in effect at the time that the plant was licensed.

• Procedures have been in place to provide the mechanisms to control the performance of design and FSAR change activities.

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The effectiveness of those procedures and mechanisms has been repeatedly assessed j l

by internal and external entities throughout the life of the plant. '

• In those instances where the need for revisions or improvements have been identified, VY has taken the actions deemed necessary.

Section 7.0 describes the processes for the identification of problems and implementation of corrective actions. This discussion addresses the role and importance of Administrative Procedure AP 0009 [ Event Reports) and the Event Report (ER) process in identifying problems and implementing solutions. The ER process is a recently upgraded and improved process which establishes the criteria for identifying and documenting potential problems and conditions adverse to quality or to the safe operation of the plant. The ER process lowered the threshold for problem reporting which, when combined with increased management encouragement of a questioning attitude on the part of VY personnel, is intended to increase the formal attention paid to potentially adverse conditions at the plant. In addition to establishing the lowered threshold, the ER process also includes detailed guidance regarding l the nature of reportable conditions, and how the conditions are to be reported, screened, evaluated, investigated, and resolved. The overall result is that a formal mechanism exists to control activities associated with design bases and plant procedural changes.

The most important aspects of the VY design change validation activities have been the ,

reviews and assessments performed by the VY staff, especially the on-going surveillances )

performed by Operations personnel, the audits performed by the Quality Assurance staff, and l the numerous issue-specific reviews and upgrade activities including programs designed to address such subject areas as Appendix R, Equipment Qualification, Generic Letter 89-13 ,

(Service Water), Motor Operated Valves (MOV), and Appendix J (Containment Leak Rate i l

Testing.) In addition, VY has been subjected to five (5) vertical slice inspections and other evaluations. The manner and timeliness in which identified shortcomings and concerns were addressed are also discussed in these sections. Even though many of these reviews were performed to address VY-specific or industry wide " problems," the result, once the reviews had been completed and necessary changes implemented, was a higher level of design confidence than had existed prior to the conduct of the specific reviews.

Section 8.0 summarizes and reiterates the information provided and positions taken in Sections 4.0 through 7.0. The information contained therein reinforces VY's conclusion that programs and processes are effective in providing reasonable assurance that the plant configuration and performance are consistent with the design bases.

Netwithstanding the conclusions that the existing processes are effective, VY has indicated its intent to undertake certain significant activities designed to provide increased confidence in the programs and processes in place. These programs include a Design Bases Documentation (DBD) Program to develop twenty-three (23) documents for safety-critical systems, an improved Technical Specifications (ITS) project, and an FSAR verification program which is intended to determine and verify the internal consistency of the FSAR and the manner in which FSAR information conforms with similar information found in other plant documents. The scope and schedule associated with these commitments are intended to be l consistent with the USNRC's recently announced discretionary moratorium in its Enforcement Policy.

2 Enclosure 1

In summation, the response establishes the bases for the conclusions that:

l The USNRC's concems have been addressed.

The manner in which the VY design bases have been implemented has been appropriate and consistent with applicable USNRC requirements and industry standards.

The actions that have been taken in the past, the programs and procedures in effect today, and the manner in which VY will operate in the future are supportive of the continued safe and effective operation of the plant in a manner which is consistent with the terms of the VY Operating License.

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2.0 INTRODUCTION

This document provides Vermont Yankee's response to the U.S. Nuclear Regulatory Commission (USNRC) letter dated October 9,1996, which requested information pursuant to 10 CFR 50.54(f) regarding adequacy and availability of design bases information for the Vermont Yankee Nuclear Power Station (VY). The USNRC letter was received by VY on October 18,1996.

The USNRC letter identified five (5) specific topics and included a request for certain other information. These include:

. Topic (a) - Description of engineering design and configuration control processes, including those that implement 10 CFR 50.59,10 CFR 50.71(e), and Appendix B to 10 CFR Part 50. This topic is discussed in Section 4.0.

• Topic (b) - Rationale for concluding that design bases requirements are translated into operating, maintenance, and testing procedures. This topic is discussed in Section 5.0.

- Topic (c) - Rationale for concluding that system, structure, and component configuration and performance are consistent with the design bases. This topic is discussed in Section 6.0.

• Topic (d) - Processes for identification of problems and implementation of corrective actions, including actions to determine the extent of problems, actions to prevent recurrence, and reporting to USNRC. This topic is discussed in Section 7.0.

. Topic (e) - The overall effectiveness of current processes and programs in concluding that the configuration of the plant is consistent with the design bases. This topic is discussed in Section 8.0.

- Request for other information - Indicate whether you have undertaken any design review or reconstitution programs, and if not, a rationale for not implementing such a program. If design review or reconstitution programs have been completed or are being conducted, provide a description of the review programs, including identification of Systems, Structures, and Components (SSCs), and plant-level design attributes (e.g.,

seismic, high-energy line break, moderate-energy line break). The description should include how the program ensures the correctness and accessibility of the design bases information for your plant and that the design bases remain current. If the program is being conducted but has not been completed, provide an implementation schedule for SSCs and plant-level design attribute reviews, the expected completion date, and method of SSC prioritization used for the review. This topic is discussed in Section 9.0.

The information described in Sections 1.0 through 9.0 of this document is not intended to create new commitments, unless explicitly stated otherwise in Section 10.0, and is not intended to preclude future procedural or process changes that would be made following VY's ,

procedures and practices. All commitments made in response to the referenced USNRC letter l are explicitly stated in Section 10.0.

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3.0 GENERAL INFORMATION 3.1 Vermont Yankee Nuclear Power Station Vermont Yankee (VY) is a 550 MWe boiling water reactor located in Vemon, Vermont. The VY license application and Plant Design and Analysis Report were submitted November 30,1966.

The construction permit was issued December 11,1967. The Final Safety Analysis Report (FSAR) was submitted December 31,1969.

The VY operating license was issued March 21,1972, authorizing operation at 1% of rated thermal power. Full power operation was authorized February 28,1973. The operating license was amended December 17,1990, to extend the expiration date of the license to March 21, 2012.

The FSAR format and content predate Regulatory Guide 1.70. The FSAR was revised in 1982 pursuant to 10 CFR 50.71 to update the information contained in the FSAR. The FSAR subsequently was revised on an annual basis until 1992 and on a refueling cycle basis since then. The next revision is scheduled for May,1997.

Control of the VY design typically has not been contracted to outside organizations. The organization would be classified as a Category 1 design organization by NUREG-1397 in that there is an extensive engineering and technical discipline oriented in-house staff with appropriate breadth and depth of engineering skills and knowledge. The Quality Assurance Program is applied under the Yankee Operational Quality Assurance Program (YOOAP-1-A).

YOOAP-1-A is based on ANSI N45.2.11 - 1974 (Quality Assurance Requirements for the Design of Nuclear Power Plants].

3.2 Development of 50.54(f) Letter Response VY established a multi-disciplined 50.54(f) Response Team to coordinate development of this response. A senior level Steering Committee provided overview of the Response Team's activities. This included overview of the Response Team's development of the work plan, review of information, and development of this response. The Steering Committee evaluated and recommended specific topics, based on its industry perspective, for further consideration by the Response Team.

This information focuses on the term " design bases" as defined in 10 CFR 50.2 and explained in footnote 4 of the referenced USNRC letter as:

"As described in 10 CFR 50.2, design bases is defined as, ' Design bases mean 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...' The design bases of a facility, as so defined, is a subset of the licensing basis and is contained in the FSAR. Information developed to implement the design bases is contained in other documents, some of which are docketed and some of which are retained by the licensee."

5 Enclosure 1

I Historical information was evaluated in preparing this response and is discussed in this document primarily to provide objective evidence of the design control, configuration control i and corrective action processes. To the extent practicable, the conclusions of this document j are drawn from the results of recent assessments. I 4

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l l 4.0 RESPONSE TO USNRC TOPIC (a)

Topic (a) - Description of engineering design and configuration control processes, including those that implement 10 CFR 50.59,10 CFR 50.71(e), and Appendix B to 10 CFR 50.

4.1 Introduction There are several programs that control the translation of the VY design bases into the engineering design and configuration control processes, foremost of which is the Quality Assurance Program. Quality Assurance activities for VY are conducted in accordance with the Yankee Operational Quality Assurance Program (YOOAP-1-A). YOQAP-1-A is based on ANSI N45.2.11 - 1974 [ Quality Assurance Requirements for the Design of Nuclear Power Plants).

The VY design control processes require that changes to the plant be evaluated against the design bases. Several of these processes are summarized below. Other processes, such as procedures for development of supporting calculations, are referenced within the primary process documents. The primary administrative and engineering processes include. .

l Desian Enaineerina Instruction WE-100 IEnaineerina Desian Chance Reauestl

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I controlling new designs or design changes prepared by Design Engineering.

VY Administrative Procedure AP 6004 [ Engineering Design Change Reauests]

controlling VY's review and approval of engineering design changes prepared by Design Engineering per WE-100.

VY Administrative Procedure AP 6002 [Preoaring 50.59 Evaluations and FSAR Changes] controlling performance of 10 CFR 50.59 screening evaluations and safety evaluations and processing of changes to the FSAR.

VY Administrative Procedure AP 0020 IControl of Temocrarv and Minor Modifications] i controlling davelopment and implementation of temporary and minor modifications.  !

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VY Administrative Procedure AP 0842 IEauivalencv Evaluation] controlling component equivalency evaluations for altemate replacement items.

= VY Administrative Procedure AP 0022 ISetooint Change Reauestsl controlling evaluation, review, approval, and implementation of changes to established setpoints.

• VY Administrative Procedure AP 0037 [ Plant Procedures] controlling development, review, approval, and revision of procedures.

VY Administrative Procedure AP 6001 Ilnstallation. Test and Soecial Test Procedures 1 controlling development and use of procedures for installation and testing of engineering design changes.

7 Enclosure 1

l VY Project Procedure VY 20 IFSAR Revision Process] controlling processing of

! changes to the FSAR.

e i VY Policy VYP 132 ITechnical Soecification Prooosed Changes and Interoretations]

controlling preparation, review, and implementation of proposed changes to the Technical Specifications.

i 4.2 Design Change Process i

l Engineering activities related to new designs and changes to existing designs of  ;

safety-related Structures, Systems, and Components (SSCs) requiring Quality Assurance are

! controlled by the above procedures.

Desian Change Initiation The normal process for initiating new additions and design changes starts with the identification of a need to add systems, structures or components to the xisting plant or the need to revise the existing design. During each cycle, such projects are delineated on the

, Major Project Work List from which VY selects those projects which are to be implemented.

Typically, the projects which are accorded the highest priority are those that are needed to comply with regulatory requirements or which will enhance plant operations.

Design Preoaration When proposed projects are selected and budgeted, Design Engineering develops the necessary design or design changes under Engineering Instruction WE-100 (Engineering Design Change Request). The developed design products or " packages" are referred to as

" Engineering Design Change Requests" or "EDCRs." WE-100 establishes the format and content of the EDCR. The procedure includes detailed lists of the parameters and design requirements that must be addressed during the development of the design " package" as well as the requirements and criteria that are to be addressed during the EDCR design review and verification processes. WE-100 requires the design change originator to evaluate the potential impact of the proposed change on the plant design bases and to document the principal design features, design assumptions, performance requirements, potential impact on plant operations, maintenance implications, testing requirements, etc. WE-100 further requires that the inputs and assumptions that are used in the development of the proposed desigr. change be described in sufficient detail to permit thorough review and subsequent correlation with the existing design bases. Where appropriate, the design change originator is also directed to consider the impact of previous changes to the system of interest and to address the cumulative effects of previous minor changes. The potential overall effect of the proposed change must be evaluated and documented in a safety evaluation.

l The design input documents are required to be prepared, reviewed, approved and controlled under a series of Engineerhg Instructions including WE-103 (Engineering Calculations and l Analyses), WE-104 (Qualification Tests), WE-105 [ Drawings), WE-106 [ Procedures and Instructions), and WE-107 (Specifications). As a result, each step of the process is required to 1

be performed in a controlled and documented manner. 1 l

8 Enclosure 1

The completed EDCR package is required to describe the proposed change and technical bases supporting the change. Typically, the package may refer to or include a number of reference documents such as:

10 CFR 50.59 safety evaluation.

Mark-ups to identify changes to affected licensing documents and other controlled documents such as the FSAR and Technical Specifications.

• Calculations.

i

. Procedures for special processes.

• Procurement information.

• Drawings, references, and review forms.

• Supporting vendor documentation.

• Project-specific enclosures.

• Engineering change notices.

Design Review The prepared EDCR must undergo independent review. The reviewer may follow the same

procedure used by the originator or may select an attemative means (i.e., alternative calculations, qualification testing) to verify the initial product. Considerations to be addressed

by the independent reviewer are also delineated in WE-100. The independent reviewer is l required to verify that the conclusions in the original safety evaluation are fully supported. In addition, the reviewer is required to verify that the pre-operational and subsequent test l requirements are appropriately specified and that necessary procedural changes are identified.

l i

In addition to the independent reviewer, a Lead Engineer is required to review the EDCR and may involve other Engineering discipline groups, as appropriate, for further review. Design Engineering management is required to perform the final Engineering review and approve the EDCR. Under AP-6004, the EDCR is required to be reviewed by the plant staff and the Plant

. Operations Review Committee (PORC) with final approval by the Plant Manager. Among other things, AP-6004 stipulates that the following actions must occur during the plant review process:

• Affected plant procedures must be identified.

• The need for revised training requirements must be verified.

• Copies of affected control room drawings must be identified and collected.

• An informational copy of the EDCR must be provided to Training. ,

• An Installation and Test Procedure must be prepared in accordance with AP-6001. J i

9 Enclosure 1 l l

l EDCR Closecut i

! Administrative Procedure AP 6022 [ Job Order Files] requires creation of a Job Order File for an approved EDCR. The Job Order File is used to compile documentation generated by a work activity performed on a system, component, or structure and to provide a tracking

! mechanism for verification of the revision of affected plant documents. In addition to compiling documents generated by the work activity, the procedure stipulates that the implementing Department Manager must verify that relevant supporting activities are properly '

performed and documented. Examples of such supporting activities would include identification of required program revisions or changes (e.g., to procedures, drawings, FSAR, Maintenance Rule Program, inservice Inspection (ISI) Program, inservice Test (IST) i Program, Appendix R Program), control of vendor technicalinformation, recommended Preventative Maintenance Program changes, etc. Tracking of such support activities is performed using the Commitment Tracking System. AP 6022 contains a detailed checklist for identification of documents relating to the design bases and stipulates that a review be conducted following implementation of the EDCR to check that changes to applicable documents have been initiated or are complete.

The Installation and Test Procedure requires verification that the installation is complete and accepted prior to declaring the system operable. After the requirements in the Installation and Test Procedure have been completed, Engineering is required to check that records are complete, test criteria have been met, and the installation and test requirements have been completed. Additionally, WE-100 requires that, as part of the close-out process, Engineering must check that the EDCR package contains the required documentation after the design change has been installed.

4.3 10 CFR 50.59 Safety Screening and Safety Evaluation l 1

VY and Design Engineering procedures stipulate safety screening and safety evaluation j activities and incorporate guidance from NSAC 125 [ Guidelines for 10 CFR 50.59 Safety )

Evaluations). Activities such as the following require safety screening or safety evaluation as I described below:

l Engineering Design Changes [AP 6004 and WE-100].

• Temporary Modifications and Minor Modifications [AP 0020].

• Lineup Deviations [AP 0155].

• Setpoint Changes [AP 0022].

• Equivalency Evaluations [AP 0842].

• FSAR Changes [AP 6002].

• Technical Specification Changes [VYP 132].

Plant Procedures [AP 0037].

Potential changes are required by procedure to be evaluated to determine if they change the facility as described in the FSAR (either by test, a change to a drawing, or a change to other information directly supporting the FSAR), or if they change a procedure described in the FSAR, or make other hardware or documentation changes that may have the potential to l i

involve an unreviewed safety question or impact the Technical Specifications. These l

l 10 Enclosure 1 I

procedures are intended to maintain accurate and up-to-date design bases information in the FSAR.

Changes which introduce an unreviewed safety question or involve a change to the Technical Specifications are submitted to the USNRC for prior approval.

4.3.1 Safety Screening and Evaluation Administrative Procedure AP 6002 [ Preparing 50.59 Evaluations and FSAR Changes]

controls the performance of a safety screening review to determine if a 10 CFR 50.59(a)(2) safety evaluation is required. AP 6002 provides a set of safety screening questions in the form of a checklist. These screening questions are evaluated to determine if a planned change requires a subsequent written 10 CFR 50.59 (a)(2) safety evaluation. The safety screening questions incorporate guidance from NSAC 125 [ Guidelines for 10 CFR 50.59 Safety Evaluations). The VY engineering and administrative processes used to control a planned change require that the safety screening questions be evaluated in accordance with AP 6002 to determine if a subsequent 10 CFR 50.59(a)(2) safety evaluation is required. AP 6002 also requires that potentially applicable or affected sections of the FSAR and Technical Specifications be identified, recorded, and reviewed during the safety screening and safety evaluation processes and that the associated changes be initiated. AP 6002 requires that the safety screening and evaluation be performed by an individual who has successfully completed training on the procedure.

Design Engineering prepares an EDCR under WE-100 [EDCR] and performs the 10 CFR 50.59(a)(2) safety evaluation. Each safety-related design change, as well as other design  !

changes that are specified to require Quality Assurance controls, is implemented by an Installation and Test Procedure. During development of the Installation and Test Procedure under AP 6001, the Cognizant Engineer determines if a supporting 10 CFR 50.59(a)(2) safety j evaluation is required to address installation-specific considerations. j 4.3.2 Screening and Safety Evaluation Review and Acoroval Requirements for review and approval of safety screenings and safety evaluations performed in accordance with AP 6002 include:

• Safety screenings require review and approval by the responsible Department Head for consistency with the design bases and additional review as required by the respective change document.

• 10 CFR 50.59(a)(2) safety evaluations require review and approval by the responsible Department Head, review by the PORC, and review and approval by the Plant Manager.

• A copy of the safety evaluation is provided to the Nuclear Safety Audit and Review Committee for review.

11 Enclosure 1

The safety screening and safety evaluation are QA records and are retained with the originating document. The completed safety evaluation is summarized for inclusion in the Periodic Operating Report.

The 10 CFR 50.59(a)(2) safety evaluation for a design change is performed under ,

Engineering Instruction WE-100, Attachment B. Design Engineering and VY review and I l approval of the safety evaluation are required in conjunction with the review of the EDCR l package. Acceptable review criteria are given in WE-100 and AP 6004. The review includes:

l

• Desian Enaineerina (oer Enainesrina Instruction WE-100)

~

i Independent review. ~  !

Lead Engineer review.  !

Supplemental review, as appropriate. I Design Engineering management review and acceptance.

• VY (oer Administrative Precedure AP 6004)

Cognizant Engineer review.

Engineering Manager review.

PORC review.

Plant Manager review and approval.

4.4 Procedure Development and Revision Administrative Procedure AP 0037 [ Plant Procedures] controls development, review and approval, and revision of plant procedures. New procedures and procedure revisions are required to be screened in accordance with AP 6002 (Preparing 50.59 Evaluations and FSAR Changes] to identify potential FSAR and Technical Specification changes and to determine if a 10 CFR 50.59(a)(2) safety evaluation is required.

Administrative Procedure AP 0037 govems the development, revision, and review of plant procedures including: ,

a Review for technical adequacy and compliance with the FSAR, Technical Specifications and Technical Specification Bases, and YOOAP-1-A requirements.

• Safety screening to determine if new or revised portions of the procedure require a written safety evaluation in accordance with 10 CFR 50.59(a)(2).

• Documentation for each new, revised, or canceled procedure to serve as the basis for the PORC's evaluation in accordance with AP 0030 [The Plant Operations Review Committee] of the results of the safety screening or the safety evaluation. The documentation is to summarize:

Non-editorial changes to the procedure and the assumptions and bases related i to the procedure revision, including differences from those used in the original i design or system / component operation.

t 12 Enclosure 1

The review of the FSAR, Technical Specifications and Technical Specifications Bases, and YOOAP-1-A to identify pertinent sections or the fact that the  !

equipment is not discussed in the documents. i The safety evaluation review. I Procedures are reviewed by additional reviewers (as assigned by the Department Head), the  ;

Department Head, and PORC, and are approved by the Plant Manager. Procedures that i implement the Quality Assurance Program also are reviewed by Quality Assurance.

Administrative Procedure AP 0030 [The Plant Operations Review Committee] stipulates that PORC review procedures (new procedures, revisions, and cancellations), setpoint changes, EDCRs, Installation and Test Procedures, temporary and minor modifications, Special Test Procedures, etc. PORC responsibilities include review of- 1 1

Proposed operations and maintenance procedures and changes, and other ,

procedures or changes as determined by the Plant Manager to affect nuclear safety.

• Proposed tests and experiments. l Proposed changes or modifications to plant systems or equipment that would require a )

change to operating, maintenance, or testing procedures. I Level 1 Event Reports (ERs) and resultant proposed corrective actions.

A periodic complete review of plant procedures is also required to be performed, including verification that:

• Each procedure step is correct.

• References in the procedure to the FSAR, Technical Specifications (including Bases),

YOOAP-1-A and other program requirements (IST, Fire Protection, Equipment Qualification, Emergency Plan, Security, Maintenance Rule, etc.) are appropriate and are addressed in the procedure.

4.5 Minor Modifications Administrative Procedure AP 0020 [ Control of Temporary and Minor Modifications] controls development, review, approval, and implementation of minor modifications to conform with design intent, plant configuration, operability requirements, and safety requirements. Minor modifications must be installed in accordance with AP 0021 [ Work Orders).

A minor modification is required to be screened in accordance with AP 6002 [ Preparing 50.59 Evaluations and FSAR Changes] to determine potential FSAR and Technical Specifications impact and if a 10 CFR 50.59(a)(2) safety evaluation is required. The minor modification process is limited to changes that do not require a 10 CFR 50.59 (a)(2) safety evaluation.

Minor modifications, by definition, do not require extensive analysis nor the additional controls provided by the EDCR process.

13 Enclosure 1

Engineering is required to review the minor modification against a check Ust of criteria to determine if additional design or programmatic reviews are necessary. The minor modification also is reviewed by the Operations Department, PORC, and Plant Manager.

Administrative Procedure AP 0020 requires creation of a Job Order File for the minor modification. AP 6022 [ Job Order Files) contains a detailed checklist for identification of documents relating to the design bases and stipulates review following implementation of the minor modification to check that changes to applicable documents have been initiated or completed.

AP 0020 further requires:

. The originator review other outstanding temporary modifications, minor modifications, and the Design Change List to determine whether the cumulative effect of other outstanding temporary modifications, minor modifications, or designs are factored into the documentation and evaluations of the temporary modification or minor modification.

f

. The originator determine if training is required by Departments whose areas of responsibility are affected by the change. A copy of the approved change is provided to Training for evaluation of the need for changes to training manuals and identification of the appropriate personnel to receive the new training.

1

• The originator specify installation details. l I

• Engineering review the modification and determine the need for additional detailed reviews.

l

• Quality Assurance determine the need to monitor the change.

4.6 Setpoint Changes

Administrative Procedure AP 0022 [Setpoint Change Requests) controls evaluation, review, approval, and implementation of setpoint changes.

4 AP 0022 stipulates that the setpoint change request include the technical basis for the proposed change and an assessment of the potential impact on the Technical Specifications, 4 FSAR, and plant procedures. The affected plant procedures are to be changed in conjunction with the approval of the setpoint change request.

A setpoint change request is required to be screened in accordance with AP 6002 [ Preparing 50.59 Evaluations and FSAR Changes] to identify potential FSAR and Technical Specification changes and to determine if a 10 CFR 50.59(a)(2) safety evaluation is required.

! The setpoint change request must be reviewed by the implementing Department for accuracy. It must also be reviewed by Operations to identify the need for procedure changes and approved by an Engineering Manager. Engineering may request additivnal reviews, as necessary, prior to approval of the change request.

14 Enclosure 1

! l l Additionally, re,iew by the PORC and approval by the Plant Manager are required if:

A 10 CFR 50.59(a)(2) safety evaluation is required.

l The setpoint is in the Technical Specifications.

The setpoint change requires a change to be made to a procedure which requires PORC review.

l 4.7 Equivalency Evaluations i

j Administrative Procedure AP 0842 [ Equivalency Evaluation) controls the component l equivalency evaluation for an alternate replacement item. AP 0842 controls the process for evaluating the attemate replacement item to determine equivalency (in form, fit, and function) l to the component that is to be replaced.

l AP 0842 requires:

l Identification of component Critical Design Characteristics (CDCs) and comparison of CDCs between the existing and proposed replacement components.

l Documentation of the basis for a replacement component's suitability in situations l where its CDCs are not obviously equal to or better than those of the component to be l

replaced.

I Assessment of the need for revision to the FSAR, or to plant documents, programs, i and operating procedures. l l

Determination of the need for additional review of the equivalency evaluation such as seismic adequacy, environmental qualification or system interaction.

Final review and approval.

The equivalency evaluation must be screened in accordance with AP 6002 [ Preparing 50.59 Evaluations and FSAR Changes] to identify potential FSAR and Technical Specification changes and to determine if a safety evaluation is required. If a safety evaluation is required, the equivalency process may not be used and the EDCR process is required.

l 4.8 Lineup Deviations  !

l l

Administrative Procedure AP 0155 [ Current System Valve and Breaker Lineup and Identification) requires that a lineup deviation that alters the design, function, or method of l

performing the function of a structure, system, or component be subjected to a safety

screening to determine if the lineup deviation willimpact the safety of plant operations. .

l Lineup deviations are screened in accordance with AP 6002 to identify potential FSAR and Technical Specification changes and determine if a 10 CFR 50.59(a)(2) safety evaluation is required.

15 Enclosure 1 t

_ _ _ _ . . _ __ _ ~ . _ _ . _ _ _ _ _ _ _ _ . _ . _ _

l 4.9 Corrective Update to Plant Drawings  ;

Administrative Procedure AP 6802 [ Drawings and Aperture Cards] controls corrective I

updates to plant drawings. Engineering is required to review and approve corrective updates to confirm the change is only a correction or clarification to the drawing and is not an inadvertent change to the facility or FSAR that would require review and approval through the EDCR or Minor Modification processes.

4.10 FSAR Revisions and Periodic Operating Reports I The FSAR is required to be periodically revised to include the effects of:

• Changes made in the facility or procedures as described in the FSAR.  ;

Safety evaluations performed either in support of requested license amendments or in - l support of conclusions that changes did not involve an unreviewed safety question.

l . Analyses of new safety issues performed by or on behalf of the licensee at .i Commission request.

Chances to information in the FSAR

, VY Administrative Procedure AP 6002 [ Preparing 50.59 Evaluations and FSAR Changes) and l VY Project Procedure VYP 20 [FSAR Revision Process] control the FSAR update process.

These procedures are intended to maintain accurate and up-to-date design bases information in the FSAR. AP 6002 and VYP 20 provide instructions for initiation and review of proposed changes to the FSAR. The procedures assess the need for a 10 CFR 50.59(a)(2)

safety evaluation and the need for supplemental technical reviews.

t l

Changes to the facility or procedures as described in the FSAR must be reviewed in a safety screening and, if applicable, a 10 CFR 50.59 (a) (2) safety evaluation. Changes that do not involve a change to the Technical Specifications or an unreviewed safety question are processed under VYP 20 or AP 6002 and may be implemented without prior USNRC

, approval. The effects of these changes and the supporting safety evaluations are required to j be documented in the operating cycle update to the FSAR as are the effects of safety l evaluations that were performed to support requested license amendments. After completion )

l of plant reviews and approvals under VYP 20 or AP 6002, the FSAR change request must be l l processed by Licensing under VYP 20 which controls the development and submittal of  !

operating cycle updates to the FSAR.

Obvious typographical errors and miscellaneous corrections to the FSAR are subjected to a 10 CFR 50.9 evaluation in lieu of a safety evaluation and must be processed by Licensing i under VYP 20.  !

New safety issues necessitating a change to the facility or procedures as described in the FSAR are required to be evaluated in accordance with 10 CFR 50.59 under AP 6002 or VYP {

l 20, as appropriate. The effects of these changes must be incorporated into the operating j cycle update.

16 Enclosure 1

Analyses documented in the FSAR are subject to re-analysis either to evaluate new safety l

issues at the request of the USNRC or to evaluate potential changes to the facility, procedures, tests or experiments. The results are required to be evaluated in accordance with 10 CFR 50.59, and the effects of changes to analyses incorporated into the operating j cycle update through AP 6002 or VYP 20.

Specific steps required by AP 6002 and VYP 20 are identified below:

l Administrative Procedure AP 6002 [ Preparing 50.59 Evaluations and FSAR Changes]

l stipulates requirements for:

• Development of the description of the proposed FSAR change and the basis for the ,

change. j l

Safety screening to determine if a 10 CFR50.59(a)(2) safety evaluation is required. I t

Review by an Engineering Manager and approval of the FSAR change and the supporting safety screen or safety evaluation.

l

.. Review by the PORC in accordance with AP 0030 [The Plant Operations Review

( Committee] if a 10 CFR50.59(a)(2) safety evaluation was required for the FSAR change, and forwarding of the evaluation to the Plant Manager for approval.

• Entry of the FSAR change into the FSAR Change Log and maintenance of a copy of the changes until the operating cycle FSAR update is completed.

• Forwarding of the FSAR change to Licensing. I VYP 20 [FSAR Revision Process] governs FSAR changes generated outside the AP 6002 process. It stipulates:

• The Cognizant Engineer identify and review the applicable FSAR sections and supporting information for operational, procedural, or design changes that affect plant equipment or procedures referenced in the FSAR.

• The FSAR change request routing form (and attached change request documentation) contain a safety evaluation or a statement as to why a 10 CFR 50.59(a)(2) safety evaluation is unnecessary.

• Licensing review each change request for accuracy and completeness, and specify supplemental reviews, as appropriate. The FSAR change request routing form i stipulates reviewers evaluate accuracy and completeness of the change and ensure assessment of the potential for a change in the design bases of the plant, as well as the potential effect on other design bases documentation (such as reports and calculations).

!

• Licensing administer and distribute the FSAR Proposed Change Matrix. (A copy of the l most current FSAR text may be requested for review. A revision bar in the margin l

17 Enclosure 1 1

e 3 annotates previously proposed FSAR changes. The change request is maintained in a retrievable fashion until the FSAR update is processed.)

The schedule be established for compilation, review and submittal of the FSAR update <

j in accordance with 10 CFR 50.71.

Periodic Ooeratina Reoort

. Completed 10 CFR 50.59 (a) (2) safety evaluations are required to be summarized for l inclusion in the Periodic Operating Report to the USNRC in accordance with AP 6002. The i Periodic Operating Report must include a description of the changes to the facility or

procedures as described in the FSAR and a summary of the associated safety evaluation.

k l

i 1-i i

i i

i i

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i 18 Enclosure 1

4 5.0 RESPONSE TO USNRC TOPIC (b)

Topic (b) - Rationale for concluding that design bases requirements are translated into operating, maintenance, and testing procedures.

VY's rationale for concluding there is reasonable assurance that design bases requirements are translated into operating, maintenance, and testing procedures is based on our design control and procedure change processes that are discussed in detail in Section 4.0, our recent program initiatives, and other audits, assessments, and inspections. A discussion of these areas is provided below.

5.1 Design Control and Procedure Change Processes VY's design change process includes measures and reviews requiring identification of the impact of design changes on plant procedures, and tracking and closeout provisions to determine that corresponding procedural revisions are made. The procedure change process also includes measures and reviews to determine compliance with the FSAR and Technical Specifications. These processes are designed to provide reasonable assurance that operations, maintenance, and testing are consistent with the design bases.

4 1

The initial VY procedures and controls were designed and implemented to test, operate, and maintain the plant in accordance with the design requirements. The initial construction, pre-l operational, and startup test programs formally tested and evaluated component, system, ,

and integrated plant performance. The results of these programs were documented,  !

reviewed, and approved under established procedures.

Operations and surveillance programs and procedures subsequently have been implemented to demonstrate performance of design features. Maintenance programs and procedures have been implemented, as have design change and modification programs, to correct hardware problems or to improve the level of component, system, and plant performance.

Over the lifetime of the plant, administrative and technical reviews and controls have been implemented to evaluate and control potential changes to the physical plant, as well as changes in operation and maintenance of equipment, and to update the corresponding l procedures. Intemal and external audits and other evaluations of work activities, plant changes, tests, operations, etc. have evaluated the effectiveness of design control,  ;

modification, testing, operations, and maintenance activities.  !

Individual tests have been performed to demonstrate design requirements on component, system, and system interface levels. Integrated tests, such as testing after the refueling outages, integrated Emergency Core Cooling System (ECCS) testing, testing after the recirculation system piping replacement, etc., have provided a wider demonstration of implementation and maintenance of design requirements.

The results of these tests have been the subject throughout the lifetime of the plant of internal and extemal review and audit. As concems have been identified in the physical plant equipment, operations, test conduct or test results, etc., these concems have been evaluated 19 Enclosure 1

l and resolved with involvement as appropriate by Engineering, Operations, the PORC, and the Nuclear Safety and Audit Review Committee (NSARC). This has resulted in physical plant i hardware and design configuration changes as well as testing, operational, and maintenance improvements over the years.

EDCRs [WE-100. AP 6004. Engineering Design Change Reauestl The need for procedure changes is required to be identified by cognizant engineering personnel during the development of the EDCR package. Marked-up procedures must be developed in conjunction with the responsible department. Operational and surveillance l testing procedures must be approved prior to declaring the system operable unless specific l exception is obtained. Tracking of required procedure changes is implemented through the >

Job Order closecut processes. A final review to verify that necessary procedure changes were made is required during project closeout.

Minor Modifications [AP 0020. Temocrarv and Minor Modifications 1 I

Procedures that are affected oy a minor modification are required to be identified by the preparer of the minor modification. Prior to approval of the minor modification, a review by Operations is required to verify that operational procedures are identified as well as other procedures requiring revision prior to implementation of the modification. For those procedures identified, controls are established to check that the revised procedure is issued prior to making the system operable. A final review of required procedure change;is made during project close-out.

Setooint Changes [AP 0022. Setooint Change Reauestl Procedures affected by setpoint changes are required to be identified during the development of the setpoint change document. Necessary changes to the procedures must be prepared and included in the document package for approval and implementation along with the setpoint change document.

Eauivalency Evaluations IAP 0842. Eauivalencv Evaluation 1

- Changes to controlled documents that are affected, including plant procedures, are required to be identified during the development of component equivalency replacement evaluations.

These changes either are made in conjunction with development of the evaluation or, if the replacement part is to be used at a later date, are deferred for submittal at that time.

Corrective Uodates to Plant Drawings IAP 6802. Drawings and Acerture Cards 1 Corrective updates to plant drawings are required to be reviewed by Engineering to determine if procedures are affected by the change. Necessary procedure changes would be identified to the responsible department for processing.

l 20 Enclosure 1

h

't 5.2 Recent Program initiatives

)

VY has recently performed, or is continuing to perform, a number of projects which are designed to result in improved design bases documentation and configuration control. The following is a synopsis of recent initiatives to provide objective evidence supporting the conclusion of reasonable assurance that design bases requirements are translated into )

operating, maintenance, and testing procedures. Although those and other programs were

! implemented to address changing regulatory requirements and potential concerns that had been raised during the course of plant operation, the fact that thorough programs were I undertaken and discrepancies resolved provides VY with an increased level of confidence l that the applicable processes are functioning properly in translating the design bases into I procedures.

Accendix R Project Design review of the Appendix R Program was undertaken as a result of self assessment findings (VY Licensee Event Report (LER) 95-14 and USNRC inspection Report No. 95-26) in July and August,1995. A Project Team, consisting of representatives from operations, fire protection, and various engineering disciplines, was formed in September 1995 to address the self assessment findings and perform a design review of compliance with Appendix R.

j This effort was completed prior to startup from the 1996 refueling outage and is documer.ted in the form of an updated Safe Shutdown Capability Analysis (SSCA), Revision 5.

The scope of work completed by the Project Team included: l 1

• Review of safe shutdown systems and their major components.

• Review of the selection of safe shutdown cables and associated components.

• Cable routing and field walkdowns.

• Documentation of safe shutdown system availability by fire area / zone.

. Licensing documentation review.

• Fire area / zone boundary review.

• Issue resolution in the form of strategy changes, design changes, procedure changes, and licensing changes.

Operating procedures (OP-3126 [ Shutdown Using Alternate Shutdown Methods) and I OP-3020 [ Fire Brigade and Fire Fighting Procedure)) were updated to reflect the safe i I

shutdown strategy changes documented in the updated SSCA. The Fire Protection Plan was revised (Revision 13, October 21,1996) to provide clear definition of the Fire Protection Program and Appendix R Program requirements. The Fire Protection Plan integrates the SCCA into the Fire Protection Program.

21 Enclosure 1

l l Eauioment Qualification (EO) Self Assessment An EQ Program audit (Audit Report No. VY-96-25) was conducted in April and May,1996, at l the request of senior management to determine if adequate design bases were established l and maintained for the EQ Program. The audit utilized a vertical slice methodology; ,

interviews with responsible engineering, management, and operational personnel; review of i l EQ bases and documentation; and review of supporting technical analyses.

The audit concluded that the EQ Program is supported by adequate design bases, documentation, procedures, and supporting technical analyses. The audit reviewed how I changes to Operational Emergency Procedures affect EO, how upgrades to the EQ Program affect the Operational Emergency Procedures, and the review and control of those changes.  !

The audit team considered the EQ Program to be directed, controlled, and supported by l l technically knowledgeable personnel. Previous self-assessments of the EO Program i activities were determined to be adequate with well-established and meaningful corrective I actions, lmolementation of Technical Soecification Surveillance Reauirements in April,1994, VY performed a self assessment to evaluate the adequacy of procedures and I' controls necessary to implement Technical Specifications surveillance requirements. During this initiative, over 180 procedures and 650 individual tests were evaluated. The evaluations included assessments of:

• Compliance with Technical Specifications. l

= Ownership. I

• Procedure clarity. l

• Acceptance criteria.

I A substantial number of comments and improvement suggestions resulted from this effort. No immediate operability concems were identified. Many of the comments and suggestions have been addressed during procedure updates, in 1995, our internal audit process (Audit Report No. 95-01) found that several issues had not been dispositioned. Corrective actions were taken and the issues identified in the audit have since been dispositioned. The improved Technical Specification Project will assess and disposition items relating to the Technical Specifications in conjunction with the conversion to the improved Technical Specificatior.s (ITS). Disposition of remaining items will be addressed in a follow up self-assessment is discussed in Section 10.0.

The overall assessment indicated that in-place procedures and controls were adequate to meet the intent of Technical Specification surveillance requirements.

1 1

22 Enclosure 1

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I  :

I t  !

l  ;

l Service Water Generic Letter 89-13 Proaram Self-Assessment l i

VY performed a Service Water Operational Performance Inspection (SWOPI) l self-assessment in 1994. This self-assessment included the Plant Service Water, Residual  :

Heat Removal Service Water (RHRSW), and Altemate Cooling Systems (ACS) and included j plant walkdowns of those systems. l The assessment identified deficiencies related to the design bases. Specific findings related  !

l to inadequate definition and documentation of the design bases including incompleteness of l ACS design bases. Findings related to the translation of design bases into operating, t maintenance, and testing procedures included: i i

Service Water System and ACS operation inconsistent with the design bases. i Inadequate thermal performance testing of heat exchangers. l 1

Use of potentially non-conservative fouling factors. l

• Maintenance and inspection procedures lacking sufficient detail to ensure consistent results. ,

1 VY formed a Service Water System Design Bases Project Team to correct deficiencies found during the SWOPl. Corrective actions included:

• Operability assessments.

• Design bases calculations and evaluations.

• ACS and heat exchanger testing enhancements.

• Maintenance and operating procedure enhancements.

In the course of generating and validating the original design bases for the Alternate Cooling Water System, testing of the system for flow and heat transfer capacity was performed and procedural modifications were implemented. The testing verified the functional capacity of the original Altemate Cooling Water System equipment as well as the procedure enhancements.

A specific procedural modification was the addition of requirements for verification that average silt level in the Cooling Tower Deep Basin is maintained at less than 5.5 inches to provide adequate system inventory. Procedure requirements stipulate that the basin be  ;

inspected each cycle to maintain this design basis. Significant work was performed to drain the Deep Basin and remove the accumulated silt. An additional project involved cleaning the supply piping from the Deep Basin to the suction of the RHRSW pumps.

Thermal performance calculations for Service Water heat exchangers have been revised to reflect either more conservative fouling factors or increased maintenance frequencies for the i Emergency Diesel Generators (EDGs), and testing improvements have been made in accordance with Generic Letter 89-13 commitments.

r 23 Enclosure 1

- - . - - - - _ _ . - .-- - . - - - - _ - . _ - - - . ~ - . . . -

i I 1  !

\

l l

Motor-Ooerated Valve (MOV) Program J The VY Generic Letter 89-10 MOV Program is designed to reasonably assure MOV  ;

oparability under design bases differential pressure and flow conditions. A total of 85 MOVs  !

3 are in the program.  !

i l The USNRC performed a MOV Program inspection in 1995 (Inspection Report 95-03). This inspection found the MOV design bases review documents to be comprehensive, j Weaknesses found included a potential for pressure locking of the Core Spray (CS) injection i valves. Another finding was that the number of dynamic tests planned was small compared to i

, the population of Generic Letter 89-10 MOVs. Although a number of weaknesses were found  ;

in the program, VY resolved the matter in a timely manner. Evaluations for susceptibility to )

j pressure locking and thermal binding were completed, and modifications and procedure ,

changes were implemented. Dynamic (i.e., differential pressure) testing has been completed j for thirty two (32) MOVs in the GL 89-10 Program. Seven (7) of the thirty two (32) have been i retested dynamically after major valve rework. Periodic dynamic testing will also be -

I considered in the MOV Periodic Verification Program which is being developed. These

! actions served to strengthen the program and enhance the level of confidence that exists in l l the program. This position was reinforced by a USNRC inspection (Inspection Report 96-05) j j in May,1996, which concluded that VY has implemented an acceptable Generic Letter 89-10  !

l program to verify the design bases capability of safety-related MOVs. Requirements l necessary for operation'under design basis conditions (differential pressure effect, degraded l

voltage, unwedging load, weak link calculations, lubrication issues) were translated into MOV l

. switch setting and testing requirements. Programmatic changes were made to MOV I maintenance and test procedures to implement these requirements. The MOV Program, l design modifications, and special tests have translated these design bases requirements into i the operating, maintenance and testing procedures.

Anoendix J Program Assessment i

VY performed an assessment during 1993 to 1996 of the Appendix J Program. This j assessment included

• Field walkdowns of penetrations.

!

• Penetration-by-penetration review of the program document.

• Assessment of plant configuration to the General Design Criteria (10 CFR 50, ,

l Appendix A), as applicable to VY. .

l

• Review of the testing methods and test procedures.  !

l

• Review of Safety Evaluation Reports (SERs) and other licensing bases documents.

• Review of other program correspondence.

This assessment resulted in a revision to the VY Primary Containment Leak Rate Test Program, and revisions to the supporting test procedures, i 24 Enclosure 1

1

, The assessment identified a number of improvements to the program as well as certain conditions that were reported as LERs (94-10,95-01,96-04 Supplement 1,96-24, and 96-30). An assessment contained within these LERs concluded that there was no significant impact on plant safety.

During December,1996, VY performed an independent assessment (Audit Report )

No. VY 96-26) of the program. The assessment report is presently being reviewed to identify ,

potential areas for program improvement. These areas will be reviewed further within our l internal audit process. i VY believes the above improvements and assessments provide reasonable assurance that the containment leakage design requirements are translated into the applicable station procedures.

IST Proaram Assessment During 1996, VY completed an assessment of the IST Program. This included: l Review of each system within the scope of the program.

Assessment of components (pumps, valves, and test instrumentation) within these systems and the safety functions they perform.

• Review of the applicable Technical Specifications, FSAR, system operating procedures, and other information to identify design functions.

• Assessment of whether the testing procedures are adequate to meet the guidance l provided in NUREG 1482.

• Development of an IST Component Basis Document which provides the basis for )

component testing requirements and an upgraded IST Program Document.

i This assessment identified a number of improvements to the program as well as certain l conditions deemed reportable as LERs (96-04 Supplement 1,96-24, and 96-30). An 1 assessment contained within these LERs concluded that there was no significant impact on

plant safety.

During November and December,1996, an independent assessment (VY Audit Report VY-96-27) was performed of the IST Program. The assessment report is presently being reviewed. The report identified that testing during the 1996 refueling outage was effective and complete. The report also identified potential areas for program improvement. These areas will be reviewed further within our internal audit process.

Overall, the report determined that the IST Program meets regulatory requirements and is consMered effective in specifying the appropriate equipment and associated testing necessary to comply with code requirements.

l 25 Enclosure 1

5.3 Audits, Assessments, and Inspections VY's configuration control processes are routinely evaluated for effectiveness through our Quality Assurance audit and corrective action programs, in addition, several special assessments using a vertical slice approach have been performed since 1988. USNRC inspections and assessments also routinely evaluate this area. The following discussion summarizes results from various audits, assessments, and inspections regarding the translation of design bases into operations, maintenance, and testing procedures.

5.3.1 Vertical Slice Assessments In addition to the SWOPI described above, five (5) plant systems have been the subject of vertical slice inspections. These include: High Pressure Coolant injection (HPCI) and EDG (1988); CS and RHRSW (1990); and electrical distribution (1992).

VY Safety Svstem Functional Insoection (SSFh of HPCI and EDG Svstems - 1988 SSFis of the HPCI and EDG systems were performed in 1988. The scope of the inspection included systems which support the EDG and HPCI systems, such as Heating, Ventilation, and Air Conditioning (HVAC), service water, Direct Current (DC) power, instrument air, and diesel fuel oil.

The inspection included review of design, operations, and management related documents, walkdowns of the system and interfacing equipment, and interviews with engineering, licensing, operations, maintenance and management personnel.

Strengths observed by the team included maintenance that was well done, post-maintenance testing, and very few examples of repeat maintenance. However, some weaknesses were observed relative to translating design bases into operations, maintenance, and testing procedures such as:

• Lack of root cause corrective action to address anomalies in surveillance test results for flow and temperature of the service water to the diesel generator. Upon further review, this subsequently was corrected by a design change which added a new discharge line from the EDGs.

• EDG load, as described in the FSAR, was not supported by a design calculation and was not reflected in the loss of normal power procedure. An EDG loading calculation was performed to support the FSAR. Operating Procedure OT 3122 [ Loss of Normal Power) was revised to be consistent with the design calculation and the FSAR. l This inspection was VY's first vertical slice inspection. Overall, the inspection team found the EDG and HPCI systems to be functional in accordance with the criteria detailed in the USNRC Inspection and Enforcement (IE) Manual, Chapter 2515.

USNRC SSFl of CS and RHRSW Systems - 1990 The USNRC performed a SSFI of the CS and RHRSW Systems in 1990.

i I

26 Enclosure 1

-- - _ - - - -- . .- _. . ~ - __ - -

The SSFl assessed the operational capability of the CS System and the service water portion of the RHR system to perform their design bases safety functions. The SSFl Team evaluated the adequacy of operating procedures, test practices, and maintenance policies as they contribute to system reliability. The Team also addressed the quality of design control and j other management programs as applied to the systems. '

l l The inspection was divided into the areas of mechanical, electrical, instruments and controls, operations, surveillance and testing, and maintenance. The following provides a brief summary of each area inspected:

l l Mechanical

• Structural and seismic design was adequate.

l

• RHRSW system was capable of removing heat at the design rate during analyzed accident conditions; however, it was not evident that the ACS was capable of meeting  :

all of its design functions. (Testing and analysis performed following the 1994 SWOPl  !

demonstrated the capability of the ACS to meet its design functions.) l

• In the RHRSW system, a potential existed for the unavailability of all pumps due to disabling of a common drain header line and/or drain line valve in the RHRSW motor cooling line. Also, the SSF1 team considered the 10 CFR 50.59 safety evaluation )

associated with the addition of the drain line valve to be inadequate. (A subsequent modification eliminated the concern for a potential common mode failure.)

• The CS system was found to be mechanically adequate, and capable of performing its intended design functions.

Electrical

• Questions were raised concerning EDG procedures and periodic surveillances, and j the loading calculations for 480 volt Motor Control Center (MCC) 8A. Prompt corrective  ;

action was taken to resolve the issues. l Instrumentation and Control

• Instrumentation and control components and subsystems were capable of performing )

their design function.

Ooerations. Maintenance. and Surveillance and Testing

• RHRSW was capable of providing adequate cooling to remove decay heat from the plant.

• Operation, surveillance, and testing of the CS system indicated that the system will fulfill its design function, however, there was a concem for potential CS pump damage due to overheating at minimum flow conditions during a small break Loss-of-Coolant l Accident (LOCA). VY completed an assessment of this concern in June,1991, and 27 Enclosure 1

--- - - . - . - - .- - - - . - ~ . . . . - , - - . - . - . - - . . -

4 I

e i

concluded that, for the spectrum of small break LOCAs which would require CS I i injection, the pump would operate in the minimum flow condition well below the time  ;

limit for intermittent operation.

• A 10 CFR 50.59 safety evaluation which addressed the closure of CS Valve 118 had ,

4 not recognized the increase in the probability of subsystem malfunction from adding j

one extra active component in the system. This issue was reevaluated with the i i resulting conclusion that the increase in probability was not significant.

The material condition of the RHR and CS system components were being effectively maintained, and effective preventative and corrective maintenance and ISI Programs l were being implemented. ,

i The team concluded overall that the systems were capable of performing their design safety  ;

functions. i l

USNRC Electrical Distribution Svstem Functional Insoection (EDSFI) - 1992 I I

The USNRC performed an EDSFl in 1992. The USNRC's overall conclusion from the EDSFl ,

was that the design implementation in the areas inspected in the electrical distribution system i was acceptable, however certain weaknesses were identified in relation to the translation of J design bases into operating, maintenance, and testing procedures, including:

l

• Unavailability of design calculations.

• Two check valves in the EDG starting air systems were not tested in a manner that proves that the disk travels to the seat promptly on cessation or reversal of flow.

Prior to the EDSFI, VY had initiated a program to upgrade or develop electrical design calculations. For example, station battery sizing calculations, Attemating Current (AC) circuit breaker coordination calculations, and diesel generator loading calculations were completed before the EDSFl. Subsequent to the EDSFI, additional calculations were completed including cable 'ampacity reviews, control circuit evaluations, and voltage regulation on low voltage systems. Additional analyses also were conducted which resulted in elimination of reverse flow test requirements for the subject EDG starting air check valves.

As indicated above, the overall conclusion was that design implementation was acceptable in the areas inspected.  !

l 1

5.3.2 Evaluation of USNRC inspection Reports, Internal VY Assessments, and ERs j l

Thirty nine (39) USNRC inspection Reports issued in 1995 and 1996 were reviewed to assess the degree to which the results of these reports would or would not support a conclusion of reasonable assurance that the VY design bases had been adequately translated into procedures and processes in relation to operations, maintenance and testing.

Additionally 18 VY intemal Audit Reports and 67 design-related VY ERs for 1995 and 1996 also were reviewed in similar assessments.

28 Enclosure 1

(-.-. . - _ - - . . - .-.--, .-------.. .--- .

I i

Noted strengths included continued improvement in management's proactive culture; j operations performance pedaining to plant safety; maintenance and I&C activities such as j Technical Specification surveillances, Limiting Conditions for Operations (LCO) corrective  ;

maintenance, and calibration methods; and Generic Letter 89-10 assessments and l evaluations, i Noted weaknesses included thresholds for event reporting, timely corrective actions,  !

Emergency Operating Procedure (EOP) calculations, preventive maintenance scheduling, )

maintenance procedures, and the maintenance scheduling database.

l As a result of the reviews conducted of USNRC Inspection Reports, intemal audit reports,  !

and design-related ERs, VY has concluded is that there is reasonable assurance that the '

design bases have been translated into operating, maintenance and test procedures, although some discrepancies have been noted. The procedures, programs ano processes '

being implemented by VY for the conduct of operations, maintenance and testing are l

, structured and are periodically evaluated to determine consistency with design bases. While l l 1 certain deficiencies were identified in both USNRC and intemal VY reviews, no safety i

! significant problems or programmatic breakdowns were identified as a result of not l maintaining or incorporating design bases information. The majority of the identified concems l have been dispositioned or are being tracked for further action and closure under AP 0028

[ Operating Experience Review and Assessment / Commitment Tracking).

L l 5.4 Summary l Over the lifetime of the plant, the reviews that procedures and programs have undergone l have resulted in an increased level of confidence that design bases requirements have been incorporated into operating, maintenance and testing procedures. As discussed in Sections 5.1,5.2 and 5.3, there have been a number of instances where reevaluations of systems or programs have been required. In certain instances, these reevaluations have resulted from issues or concems that were identified either intemally by VY or by the USNRC. In either case, however, the fact that such reviews were conducted increases our confidence in the ability of both the physical plant hardware and VY's processes to function as anticipated. This increased confidence results from the scrutiny that VY's systems and process have  !

undergone and the resolution programs that have been completed to resolve specific issues.  !

In other words, what at first may have seemed to be negative issues when initially identified l

has in the long term resulted in an added level of confidence in the assessment of how the design bases are translated into procedures and programs.  !

VY's procedures for controlling plant configuration, operations, surveillance, maintenance, and testing are designed and intended to maintain the plant's design bases. Review of recent l l program initiatives, and other audits, assessments, and inspections provides objective evidence of the generally effective implementation of these procedures and programs. The i review of design control and procedure change processes confirms that these processes contain the requirement for incorporating new or changed design bases requirements into the l appropriate operating, maintenance and testing procedures. In addition, recent program i' i initiatives and assessments (i.e., Appendix R, Equipment Qualification, Implementation of l Technical Specification Surveillance Requirements, MOV Program, Appendix J Program, IST

! Program, etc.) have resulted in program or procedure enhancements and changes that 29 Enclosure 1

provide reasonable assurance that design bases requirements are incorporated into the ,

appropriate operating, maintenance, and testing procedures. Therefore, VY concludes that i there is reasonable assurance that design bases requirements are translated into operating, maintenance, and testing procedures.

I

}

l 1

i l

l 30 Enclosure 1

6.0 RESPONSE TO USNRC TOPIC (c)

Topic (c) - Rationale for concluding that system, structure, and component configuration and performance are consistent with the design bases.

VY's rationale for concluding there is reasonable assurance that system, structure, and component configuration and performance are consistent with the design bases is based on the results from our configuration control processes, recent program improvements and other initiatives, walkdown and modification projects, and other audits, assessments, and inspections. A discussion of these areas is provided below.

For modifications since the original design, the design bases requirements are translated into the physical design through the configuration control processes. Physical installation is required to be accomplished with approved procedures with in-process quality inspections and post-modification acceptance testing as necessary. These inspections and test procedures include acceptance criteria established from the design inputs.

Maintenance and operational changes to the physical plant and system lineups are required to be evaluated against the approved physical plant configuration so that changes to I

( controlled documents are identified and reviewed. They also are controlled procedurally during the specific activity and during restoration to an approved configuration.

Normal operations, routine operator rounds, surveillance tests (including integrated ECCS J

tests, logic tests, system and component functional tests, calibration tests, etc.), routine and l corrective maintenance, post maintenance tests, and post-event reviews provide a continuing  !

assessment of the physical configuration and performance compared to design bases acceptance criteria. Deviations from established procedural criteria must be recorded and reported for further evaluation, review, and corrective action.

The Basis for Maintaining Operation (BMO) Guideline provides management's expectations  !

regarding appropriate actions to be taken when systems or components are or potentially may be:

. Degraded such that performance or operability is in question. i

• Non-conforming because qualification is in question.

• Potentially in an unanalyzed condition or outside the design bases.

The deficiency is screened to determine if a 10 CFR50.59 (a) (2) safety evaluation is required. A BMO evaluation is performed to provide the basis for maintaining continued operation with a known or potential deficiency in the analysis, design, or qualification of safety related, Equipment Qualification Program, or other systems, structures, or components identified in the Technical Specifications. The purpose is to demonstrate and document that there is no unacceptable reduction in the protection vf public health arid safety, and/or there are appropriate compensatory factors that can be applied in the interim until the deficiency is corrected. Engineering is responsible for performing an independent technical review of the BMO's operability determination to verify the adequacy of the supporting information. PORC review and Plant Manager approva! of the BMO are required.

31 Enclosure 1

6.1 Configuration Control Processes l VY's design control processes, described in Section 4.0, call for the identification and update of affected plant configuration control documents (e.g., drawings, FSAR, etc.) and completion of post-modification testing to verify adequate installation and testing of plant design activities. Controlled documents exist to maintain plant configuration consistent with design bases requirements and to maintain surveillance procedures so that they test systems to design requirements. A summary of these processes is provided below.

ED_CR IWE-100. AP 6001. AP 6004. Engineering Design Chanae Reauests]

Changes to configuration control documents (e.g., drawings, FSAR, etc.) are required to be identified by cognizant Engineering personnel during development of the EDCR package.

Marked up copies must be provided as an enclosure to the EDCR during review and approval. During the review, approval, and implementation of the EDCR, changes are made as necessary to support implementation. These changes are required to be reviewed by the cognizant design engineer for consistency with the intent of the design. Final as-built drawings are generated to reflect the final configuration. A computerized listing of plant drawings is available to engineering personnel to aid in the identification of potentially j affected drawings. Changes to other configuration control documents must be initiated during  ;

project closecut and must be tracked to completion.

Testing requirements to verify the design adequacy are established during development of 2 the EDCR and the Instaliation and Test Procedure. Modifications to periodic surveillance I testing procedures also are identified and resulting changes to procedures are made. A final ,

review during Job Order File closeout is performed to determine that drawing and procedure changes are completed or are tracked to completion. The Response Team evaluations of the l design change closecuts discussed in Appendix A to this enclosure provide confidence that i the process is effective in maintaining configuration consistent with design bases.

Minor Modification IAP 0020. Temocrarv and Minor Modifications]

Changes to configuration control documents are required to be identified by the cognizant Engineering personnel during development of the minor modification. Changes to these configuration control documents must be submitted during project closeout.

Installation verification and testing requirements are required to be identified during development of the minor modification. An appendix is provided in AP 0020 as guidance in developing the required tests. Necessary changes to surveillance tests also must be identified to provide for periodic testing as necessary.

Setooint Changes (AP 0022. Setooint Change Reauest]

Changes to configuration control documents must be initiated during development of setpoint changes by the cognizant Engineering personnel. Affected surveillance testing procedures also are to be identified and changed along with the setpoint change to provide for periodic testing of the revised setpoint.

32 Enclosure 1

Eauivalency Evaluations IAP 0842. Eauivalency Evaluations]

Changes to configuration control documents must be identified during development of the equivalency evaluation by the cognizant Engineering personnel. Changes to plant procedures ,

are to be identified during development of the equivalency evaluation. These changes either l are made in conjunction with development of the evaluation or, if the replacement part is to )

be used at a later date, are deferred for submittal at that time, if the replacement part is to be '

used at a later date, changes to configuration control documents will be made as part of the i work order closecut process.

Corrective Uodates to Plant Drawinas IAP 6802. Drawing and Acerture Cards)  !

Corrective updates to plant drawings are required to be reviewed by Engineering to determine if changes to configuration control documents are needed due to the draw.ag correction. Identified changes are required to be developed and initiated in conjunction with the dramng update. i 6.2 Recent Program improvements and Other Initiatives VY has completed a number of efforts to assess and improve the adequacy of plant SSCs configuration as well as to evaluate the adequacy of existing surveillance procedures in verifying design requirements. A summary of these efforts is provided below.

In the past, the engineering programs that provide programmatic approaches and controls, as discussed in Section 5.0, typically have been independently reviewed, and certain aspects of those programs have been the subject of additional program development and confirmation activities. Additionally, the Response Team performed additional assessments of VY programs and their products in developing this response. These additional assessments are discussed furtherin Appendix A.

VY is currently assessing, validating, and improving a number of engineering programs.

These programs are living, ongoing processes and provide additional assurance that the design bases are maintained. In addition to program improvements, other initiatives have also been undertaken to improve configuration control. The following programs and initiatives ,

are described further in this section- i l

• Appendix R Project / Fire Protection Improvement Program. i

• Equipment Qualification Program.  !

• Technical Specification Surveillances. j

• IST Program.

• Appendix J Leak Rate Testing.

• Maintenance Rule Program.

• Post Accident Monitoring Program.

Technical Specifications improvement Project.

• Improved Setpoint Program.

• Input Assumptions Source Document (IASD).

• LOCA Analyses. '

l

• ECCS Single Failure Analysis Re-evaluation.

1 33 Enclosure 1

J l

• FSAR Update Efforts.

• Systems Engineering Initiative.

• Safety Classification Manual. l l

These programs and initiatives are designed to result in improved design bases l documentation and configuration control. The following is a brief synopsis of these programs and initiatives from the standpoint of conformance of system, structure, and component configuration and performance to the design bases.

Aooendix R Prolect/ Fire Protection Imorovement Program The Appendix R Project evaluated the SSCA design bases documentation in conjunction with development of Revision 5 of the SSCA. This included plant walkdowns to document the conformance of the SSCA cable routing through fire areas to the assumptions in the SSCA.

Physical fire protection features (dampers, seals, barriers, doors, suppression capability) that are relied on in the Appendix R analysis were walked down or evsluated to document how their configuration and integrity support the SSCA assumptions.

The Fire Protection Plan was revised to provide clear definition of the Fire Protection Program and Appendix R Program requirements, and to integrate the SSCA into the Fire Protection Program. Ongoing procedural surveillances of Fire Protection equipment and features are conducted to establish that configuration and performance are maintained in accordance with established criteria.

Self-assessment of the Appendix R and Fire Protection Programs identified deficiencies requiring deployment of fire watch personnel until additional design modifications are implemented. These deficiencies are described in ERs (95-03,94-11,95-14) and USNRC Inspection Report No. 95-26. The reports identified problems in implementing Section Ill. G and Ill J requirements. These deficiencies were not a degradation of original design bases, but rather were design deficiencies resulting from an improper or incomplete understanding .i of new requirements promulgated in the late 1970s and early 1980s.

l Eauioment Qualification Program The Equipment Qualification Program is implemented through detailed procedural processes ^i to evaluate, document, and maintain component qualification for accident environments.

Equipment qualification design criteria are implemented through procedural processes controlling the procurement, storage, dedication, substitution, maintenance, and replacement activities necessary to maintain qualification status.

A series of engineering walkdowns were formed during and subsequent to initiation of the Environmental Qualification Program. Initial walkdowns documented equipment and component location and installation. The purpose of these walkdowns was to verify that equipment and components met qualification requirements for potential post accident harsh environmental conditions. Subsequent activities verified or refined the building structural configurations to support the models used to calculate environmental conditions following various hypothesized accident scenarios such as high energy line breaks and post accident heatup in the Reactor Building.

34 Enclosure 1

Concerns relative to the effectiveness of VY's implementation of new design and licensing requirements imposed on the plant subsequent to receipt of the operating license (e.g. the 1995 Appendix R self assessment findings) led VY management to require a detailed assessment of the Environmental Qualification Program. The assessment was performed in 1996 with the audit conclusion that the program is supported by adequate design bases, documentation, and supporting technical analyses.

A more recent effort has been undertaken to enhance the Reactor Building model, using a new computer code, to calculate Reactor Building environments following various high energy

, line break scenarios. This effort is designed to analyze and provide operational flexibility by i modeling potential accident environment communication pathways between building areas (i.e, between rooms, floor elevations, etc.) This work currently is in progress and involves further walkdowns and analyses. The new code will provide an improved and more accurate treatment of steam condensation and will represent a program improvement when I

implemented.

j Technical Soecification Surveillances As discussed in Section 5.2, VY performed an assessment of the adequacy of Technical l Specification surveillance implementing procedures in 1994. Surveillance testing require' d by ]

l the Technical Specifications is intended to confirm the plant is operated and maintained '

within its design bases envelope. The assessment evaluated over 180 procedures and 650 i

individual tests and concluded overall that, while a number of procedure improvements were recommended, the in-place procedures and controls were adequate to meet the intent of I Technical Specification surveillance requirements. j IST Prooram An assessment of the IST Program was completed in 1996. This assessmant included a

! system-by-system and component-by-component evaluation of 20 systems and  !

l approximately 800 components. The review determined that design-related requirements are-tested by the IST surveillance testing program. The bases for including or not including certain components in the program were reviewed and documented in an IST Component Basis Document.

Components that were determined to be subject to IST but had not been included in the Program were added to the scope. Implementing test procedures were written for the 1996 refueling outage for these components, and the additional testing was completed during the outage. The supporting IST Component Basis Document was revised to describe the critical attributes to be tested.

An audit (Audit Report No. VY-96-27, November and December,1996) of the IST Program determined that the IST Program meets regulatory requirements and is considered effective in specifying the appropriate equipment and associated testing necessary to comply with Code requirements. The audit resulted in an administrative finding and recommendations for improvement that are being dispositioned.

35 Enclosure 1 1

i Accendix J Leak Rate Testina i

l The Appendix J Leak Rate Testing Program underwent a penetration-by-penetration review l in the 1993 to 1996 timeframe to determine that testing is in accordance with 10 CFR 50 Appendix J, the FSAR, and applicable SERs. This review included physical walkdowns of accessible penetrations.

l The review identified valves that had not been leak rate tested under the Appendix J i Program. Modification packages and associated test procedures were developed, and the l components were tested. These issues were evaluated to not be safety significant but were necessary to bring the Program into compliance with 10 CFR 50 Appendix J.

Audit (Audit Report No. VY-96-26, December,1996) of the Appendix J Program determined that the Primary Containment Leak Rate Testing Program meets regulatory requirements in i that the appropriate plant systems, components, and associated testing comply with the requirements of 10 CFR 50 Appendices A and J. The audit resulted in findings which were j entered into the ER process, and one recommended area for improvement that will be dispositioned.

Maintenance Rule l \

The VY Maintenance Rule Program is designed to identify maintenance and other plant l performance issues and to implement corrective actions to improve performance. The results l of the program are intended to provide reasonable assurance that design assumptions and margins are maintained and to assist in setting priorities and allocating resources on the basis of performance monitoring data and risk significance. Integration of the Maintenance Rule Program with the ER process and the Operating Experience Review process is designed to result in adjustments to the plant maintenance program such that there is an increased likelihood that SSCs in the scope of the Maintenance Rule will perform their intended function.

In January,1996, the Maintenance Rule Implementation Plan was assessed at the request of VY management. The overall conclusion was that VY had a deliberate and conservative

, approach to the Maintenance Rule and that strengths included prograrn documentation, PRA involvement, formation of a Systems Engineering function, plant personnel awareness of the Maintenance Rule, and Project Team knowiadge.

i The monthly Performance Monitoring and System Status Report is designed to provide performance data and status for systems within the scope of the Maintenance Rule. The report is design to summarize and highlight those systems that warrant special attention based on evaluation of failure data against prescribed performance criteria. The December, l 1996, report identified seven (7) systems in the (a) (1) Category of " Performance l Improvement Plan (PIP) and Goal Monitoring Required." The remaining sixty eight (6f systems were in the (a) (2) Category of " Monitoring is Demonstrating Acceptable Performance." Of these sixty eight (68), nine (9) are approaching the (a) (1) Category criteria and another five (5) are receiving additional performance trend monitoring or performance evaluation for disposition as (a) (1) or (a) (2) Category status. The Maintenance Rule Program is producing meaningful results and corrective actions. As an example, corrective 36 Enclosure 1

i l

actions within the Performance improvement Plans have ranged from specific improvements  !

in plant maintenance activities to the identification of design changes.

Based on the results of data on system perforrrance as evaluated at the end of 1996, corrective actions and positive results from the Maintenance Rule Program are in evidence l supporting the conclusion of reasonable assurance that design basis configuration and i performance are being maintained.

Post Accident Monitorina Proaram Regulatory Guide 1.97, Revision 3, identifies the parameters to be monitored during and i following an accident and the range over which each parameter is to be monitored. The VY I Regulatory Guide 1.97 response was resubmitted in March,1996, to provide an updated i commitment status. A Regulatory Guide 1.97 Guidance Document (Revision 0) was issued in '

December,1996, on an interim basis to provide guidance on VY's Regulatory Guide 1.97 commitments and on VY's approach to maintaining these commitments. Personnel training, l procedural revisions, and upgrades to equipment (if required) are planned. i The Regulatory Guide 1.97 Guidance Document is designed to provide the means for individuals to have ready access to the Regulatory Guide 1.97 commitments and to maintain those commitments. The document addresses the Regulatory Guide criteria and establishes

" ground rules" and specifics as to how VY satisfies these criteria.

Imoroved Technical Soecifications (ITS) Project I The VY ITS Project is designed to convert the VY custom Technical Specifications into the format defined in NUREC 1433, Revision 1. The Project also is designed to confirm, identify, or develop and document design bases for the setpoints and specifications in the ITS. In situations where calculations or analyses to adequately document these design bases are not located, the bases are required to be reestablished through performance of new calculations, analyses, etc.

The ITS Project is conducted under a formal program plan incorporating guidance from NUREG 1433, Revision 1 (BWR/4), NUREG 1434, Revision 1 (BWR/6), and Nuclear Energy Institute (NEI) 96-06 [lmproved Technical Specifications Conversion Guidance). The Project is designed to improve operational safety and provide a clear understanding of the Technical Specifications requirements by confirmation, identification, or development and documentation of the bases for each setpoint or specification in the ITS. VY letter BVY 95-122, dated November 13,1995, advised the USNRC of VY's plan to implement the ITS.

The ITS development phase began in June,1996, and will continue until the submittal to the USNRC which is scheduled for September,1997. The implementation phase will continue until the specifications have been approved by the USNRC and are in effect at VY.

The basis for a numerical value in the iTS is required to have a reference that typically will be either a calculation package or a General Electric document. Other values (numbers) or statements in the Technical Specifications and the Technical Specification Bases are 37 Enclosure 1

l required to be evaluated to confirm that the values or statements are correct and the capability exists for adequate monitoring of the values.

Project activities to date have identified instances where adequate bases could not be located, including situations where calculations were not available to document the bases.

l Such deficiencies or questions are required to be evaluated by the ITS Team for potential l operability considerations. If it is determined to be a potential operability issue, the deficiency l or question is entered into the ER process and the need for an operability determination is l made under the ER process. These instances are to be tracked in the ITS ER tracking log.

The ITS Project involves several associated initiatives to assemble or verify design bases information. These include:

Instrument Setpoint Project - A setpoint program will be implemented that includes setpoint and drift calculation guidance documents. Existing setpoint calculations will be revised and new calculations developed in accordance with the setpoint program to provide the bases for the allowable values identified in the ITS. The setpoint program will be designed to meet the intent of Instrument Society of America (ISA) SP67.04

[Setpoints for 'aclear Safety-Related Instrumentation Used in Nuclear Power Plants).

l Improvements in the Technical Specifications Bases to clearly state the intent and assumptions associated with the Technical Specification requirements.

Verification of Technical Specifications Bases against the design bases contained in the FSAR; SERs; other design documents such as General Electric specifications, studies, calculations, etc.; and existing design bases documentation.

Coordination of appropriate corrections of inconsistencies or inaccuracies between the Technical Specifications Bases and other design bases documentation.

• Documentation of the bases for equipment allowable out of service times.

• Documentation of the bases for extended surveillance test intervals.

• Re-evaluation of limiting and non-limiting transients and associated setpoints.

• Recalculation of torus volume and temperature requirements.

Implementation of the ITS Program is designed to improve operational safety by allowing operators to focus on those requirements which are most important. In addition, the improved Bases are designed to provide a clearer understanding of the Limiting Conditions for Operation and Surveillance Requirements, as well as to provide references to additional or more in depth bases documents.

Imoroved Setooint Proaram l VY has developed uncertainty and setpoint calculations for the instrumentation associated with the Technical Specifications and Regulatory Guide 1.97. The calculations were 38 Enclosure 1

l developed to meet the intent of Regulatory Guide 1.105 [ Instrument Setpoints for Safety-Related Systems) and ISA SP67.04. The basic methodology was developed for calculation of setpoints, and the results using this methodology were incorporated into the analysis of

, record, as appropriate.

In 1995, it was recognized that additional controls would be beneficial to maintain the calculations and to continue to adequately reflect the installed condition of the plant. The 4

Improved Setpoint Program was begun, and selected plant procedures to inform the Setpoint

, Control Coordinator of changes were identified for revision. The Improved Setpoint Program has obtained plant-specific "as found"/"as left" data and is in the process of performing statistical drift analyses under an approved Drift Calculation Guide. A Setpoint Design Guide also is being developed. The design guides are intended to be major components of the improved Setpoint Program. It is expected that the existing setpoint and uncertainty calculations will be revised and new calculations will be generated.

The Improved Setpoint Program is an integral part of the ITS Program. The results will be coordinated with the Departments impacted by the results of the uncertainty and setpoint calculations.

Inout Assumotions Source Document The IASD was developed to contain the assumptions used in the plant safety analyses, j including physical plant conditions or bounding conditions. Information from the FSAR and the installed plant configuration were used in development of the IASD. The IASD is designed s

to provide a central location of information on the input assumptions used in the safety )

analyses to allow comparison of specific plant or equipment situations to the analyzed conditions. This input information is available for safety analyses or reviews of reload

analyses, design changes, LERs, vendor Service Information Letters (SILs), industry issues,

! etc. This information also is available to supplement the FSAR Chapter 14 information when specific safety evaluations are performed.

i Loss of Coolant Accident Analysis A self assessment of LOCA analysis activities was performed in March,1996, to evaluate

. performance in meetirig requirements of 10 CFR 50.46 and the RELAP5YA SER for VY LOCA Analysis and other applications of RELAP5YA. The self assessment concluded that the requirements of 10 CFR 50.46 and the RELAP5YA SERs were met and the results of each application were appropriate as documented in the respective calculations.

While performing the Appendix R Program design review as discussed earlier, the Appendix R Project Team's questioning attitude resulted in identification of an RHR system minimum flow valve single failure vulnerability (ER 96-0229) that potentially could have rendered each RHR train inoperable. The single failure vulnerability resulted from implementation of an RHR system engineering design change (EDCR 73-31) in 1974 in which it was not recognized that the minimum flow valves had been changed in 1971, prior to plant startup, from "normally open" to "normally closed." The LOCA analysis had assumed these valves were "normally open." In this instance, the physical plant configuration differed after 1974 from the LOCA analyses assumptions. When discovered in 1996, engineering 39 Enclosure 1

assessments and a 10 CFR 50.59 safety evaluation were performed to return the minimum flow valves to "normally open" status. Short and long term corrective actions were identified.

Plant programs were reviewed to check that the correct position of the minimum flow valves was documented in each program. The review identified one program (Appendix J) which required revision to incorporate the change in valve position. This has been completed. The ECCS also were re-examined for single failure vulnerabilities. Remaining corrective actions belude scheduled completion of training, submittal of Vermont Yankee's updated FSAR under 10 CFR 50.71(e) and completion of Design Basis Documents. The corrective actions are currently scheduled for completion in 1997.

ECCS Single Failure Analysis Re-evaluation The LOCA analysis documented in YAEC-1772 [VY LOCA Analysis - June 1993] relies on the single failure assessment previously performed by General Electric. A re-assessment was recently performed of the single active failure assumptions used in the LOCA analysis. The .

purpose of the re-assessment was to detemiine that the limiting single active component failures of the "as-built" and "as operated" ECCS are reflected in the current LOCA analysis.

The evaluation objectives included identifying the most limiting single active component l failures and confirming the ECCS / LOCA design bases analysis of record encompasses the most limiting ECCS configuration. The evaluation included a review based on VY design documents. These included LOCA analyses calculation files; design documents (piping and instrumentation diagrams, one-line diagrams, etc.); operating procedures for front-line ECCS systems including High Pressure Coolant Injection, Automatic Depressurization, Low Pressure Coolant injection, and Low Pressure Core Spray; one-line diagrams for electrical support systems; logics for ECCS system actuation and auxiliary trip logics; and Service Water System Single Failure Analysis. The evaluation considered the "short term" ECCS  ;

injection mode which, from a design basis LOCA analysis standpoint, occurs in approximately ,

the first ten (10) minutes of an event and refers to the post accident period including vessel '

blowdown, lower plenum refill, and core reflood to re-establish core cooling.

The evaluation summarized the short term total ECCS available for each limiting single active l failure and accident break location. The total ECCS available was compared to the limiting cases evaluated in the design basis LOCA analysis for the short term ECCS injection mode.

The results of the evaluation demonstrated that the total ECCS available during the short term for each limiting single active failure (a support system failure) and accident break location meets or exceeds the assumptions in the LOCA analysis of record. The current ECCS design and standby mode alignment were determined to meet or exceed design basis LOCA analysis assumptions. No additional (i.e., more limiting) single active failures for the short term ECCS injection mode were identified. This single failure analysis confirms the adequacy of the limiting single failure assumptions used in the VY LOCA analyses.

FSAR Uodate Efforts Between 1994 and 1996, certain FSAR Chapters were revised to verify, clarify, and update portions of the document. These chapters included:

4

• Chapter 6 Core Standby Cooling Systems 40 Enclosure 1

• Chapter 7 Control and instrumentation

• Chapter 8 Station Electric Power Systems Chapter 10 Station Auxiliary Systems (partial)

These FSAR chapter updates are separate initiatives from the periodic FSAR updates that are required on an annual or refueling cycle basis. These chapter updates have resulted from i

specific initiatives by Engineering to confirm the existing FSAR information in the respective chapters against engineering and design bases information and to make corrective updates to the FSAR information.10 CFR 50.9 /10 CFR 50.59 evaluations were performed for discrepancies identified in the FSAR information.

Systems Engineerina Initiative VY has established the Systems Engineering Department with objectives to develop and perform vital roles in design baser lmplementation and maintenance. The roles are expected to grow into:

Systems engineering responsibility for design bases documentation.

• Dedicated systems depth in the overall configuration management process.

• ' Emphasis in testing and maintaining system design bases functions through the VY operating, maintenance, and testing procedures.

Safetv Classification Manual An initiative is underway which is designed to enhance the existing Safety Classification Manual into a Safety Classification Program that will be implemented under a controlled Program Document, Program Manual, and implementing procedures. The Safety Classification Program is intended to identify and document the importance of items (i.e.,

structures, systems, components, parts) to the safety of the plant. This classification is intended to indicate the requirements of the Quality Assurance Program that are to be applied to activities (i.e., maintenance, testing, part or material replacement, etc.) affecting the item.

The initial activities are designed to include development of program procedures and review or revision of the current Manual. Safety classification identification activities involve review of the existing documentation to determine what safety classification, if any, has been established by previous technical reviews for an item. Each system's components and boundaries (including composite systems, as applicable, such as primary containment, reactor coolant, ECCS, etc.) will be confirmed or established. Subsequent safety classification determinations will be designed to document the safety classification of an item from technical review and evaluation of the item's impact on the safety function of the parent system or component.

i The Program Document will be designed to include the safety classification criteria and the l bases for significant elements of the Program. Safety classification information currently contained in the existing Safety Classification Manual will be evaluated for inclusion. The 41 Enclosure 1

classification of other structures, systems, components, and items will be determined through procedural /worksheet safety classification determinations.

1 As DBDs are developed, it is anticipated that the safety classific ation information may be i cross-referenced to the DBDs to bridge between the design bases and the safety classification for a system. For a system where a DBD is not planned, the Safety Classification Program Document may serve as a repository for the definition of the system's design bases safety function.

6.3 Walkdown and Modification Projects i

Walkdowns I Engineering evalumn, walkdown, and other assessment activities through the years of plant operation have involved field verification of various aspects of the design of systems, structures and components and have been designed to evaluate the consistency of the physical plant configuration with the respective design bases. Each time a specific walkdown effort was planned or a potential design change or improvement was evaluated, the specific activity provided another opportunity to review and evaluate how the design bases have been translated into the physical plant and how the design bases have been maintained. This I provides the opportunity to re-look at the design bases and to re-confirm that the bases have been translated and maintained adequately.

In response to IE Bulletin (IEB) 80-11 [ Masonry Wall Design], VY created a comprehensive data base of concrete block wc.lls in the plant. These walls were evaluated for potential failure during a design basis earthquake and determined to be acceptable based on either the then existing configuration or subsequent modification upgrades. The program is maintained ,

through periodic walkdowns and administrative controls (i.e., signs, drawings, procedures), l VY identified an error in a block wall calculatior, in 1996 for masonry walls in the main station battery room. This error was discovered while performing walkdowns and evaluations as part l of the program for Unresolved Safety issue (USI) A-46 (Seismic Qualification of Equipment in l Operating Nuclear Power Plants). A modification was made to the battery racks to correct the problem.

In addition to walkdowns that have been conducted in the Equipment Qualification and Appendix R Programs, other plant walkdowns have been conducted to respond to IEB 79-02 ,

[ Pipe Support Base Plate Design Usir.g Concrete Expansion Anchor Bolts] and IEB 79-14 )

[ Seismic Analysis tur As-Built Safety-Related Piping Systems). In response to these bulletins, I safety-related piping systems were reanalyzed and reconstituted. This reanalysis resulted in design changes, new or modified supports, and a new baseline for the safety related piping  ;

systems.

VY completed additional walkdowns in 1995 and 1996 to respond to USI A-46. Modifications were implemented to seismically upgrade equipment seismic supports, and a schedule was established to resolve additional outliers in accordance with USI A-46 criteria.

Plant walkdowns were performed in support of SSFls, an EDSFl, and a SWOPl for the Service Water, RHR Service Water and the Alternate Cooling Water Systems. Minor 42 Enclosure 1

l t

) deviations from design bases configurations were found. However, the deviations proved not to be safety significant, and there were no implications of programmatic problems.

! These walkdowns and resultant modifications serve to provide confidence in the ability of the  :

l VY processes to maintain the design bases of the plant. i Modifications j

, Design change activities are required to be evaluated, reviewed, approved, and performed to

procedural controls which specify that potential impacts on the design bases be evaluated and documented in the change package. Over the years, the design and engineering  :

programs, as well as other plant programs, have been strengthened in line' with industry .  :

practices. As actual or potential deficiencies were identified in these processes, specific -  !

. improvements or enhancements were implemented. Significant design changes or

. modifications over the years have included among others:  ;

i-

• Cross powering of RHR pumps - removal of loop selection logic. y

7 l

• Advanced off gas system. l

t. .. .  !

' Mark I containment (torus) modifications.

j p

• Containment inerting. l 1

l

• Regulatory Guide 1.97 instrumentation.

• Equipment Qualification Program implementation.

i

• Recirculation system piping replacement.

• Piping stress reanalyses for seismic Class I systems.

]

• Spent fuel pool cooling system upgrade (addition of a larger capacity, redundant, j seismically and environmentally qualified system).

j

• Containment vent capability for beyond design bases conditions.

l

• Station blackout modifications including underground relocation of the Vernon Tie line i to provide a backup AC power source in the event of loss of the normal onsite and

offsite AC power sources.

4'

• New low pressure turbines.

i

• Moisture separator upgrade.

}

• Motor operated valve modifications.

i

• New 3A/B,4A/B, SA/B feedwater heaters.

43 Enclosure 1

} >

4  !

l l

Seismic / structural modifications.

10 CFR 50, Appendix R upgrade.

• Feedwater check valves.

Scram solenoid pilot valve replacement.

Reactor vessel core shroud repair, When the translation of the design bases is revisited in a design or design change activity, potential inconsistencies that may be identified are required to be evaluated and resolved. As a result, that specific portion of the design bases is further solidified and thus provides an increased level of confidence in the implementation and maintenance of the bases. The results from the VY Design Bases Documentation Program described in Section 9.1 are intended to inomase this level of confidence even further.

6.4 Audits, Assessments, and inspections i VY's configuration and control processes are routinely evaluated for effectiveness through our Quality Assurance audit and corrective action programs. In addition, as described in r Section 5.3, several specific assessments using a vertical slice approach have been performed since 1988. USNRC inspections and assessments also routinely evaluate this area. The following discussion addresses various audits, assessments, and inspections regarding the conformance of SSC configuration and performance with tha design bases.

6.4.1 Vertical Slice Assessments As described in Section 5.0, six (6) vertical slice assessments have been performed. These assessments identified certain issues relative to discrepancy between SSC configuration and performance and their design bases. VY's resolution of these issues also is discussed.

6.4.2 50.54(f) Response-Specific Evaluations Specific additional verifications were conducted in preparing this response to provide an additional, up-to-date perspective of conformance to the design bases. Targeted validation i reviews of a sample of engineering products were conducted. These validation reviews were performed on engineering products produced since 1989 which would be representative of recent engineering efforts associated with VY. The validation reviews included:

• Design change control processes, including the supporting VY and Design Engineering procedures.

• Calculations, engineering analyses, and engineering evaluations.

• 10 CFR 50.59 safety screening and safety evaluation processes, including supporting VY and Design Engineering procedures.

44 Enclosure 1

l l The conclusions from these reviews support our conclusion that there is reasonable l assurance that system, structure, and component configuration and performance are consistent with the design bases. The summary of these reviews is provided in Appendix A.1.

In addition, the Response Team reviewed the design change procedure (Design Engineering Instruction WE-100 [EDCR]) against key elements based on the criteria for design change controlin ANSI N45.2.11. The review concluded that each of these key elements to meet the goal of providing design control was evident. The summary of this review is provided in ,

Appendix A.2.

6.4.3 Evaluation of USNRC Inspection Reports and Internal VY Assessmenta The Response Team reviewed thirty five (35) USNRC Inspection Reports issued in 1995 and 1996 to assess the degree to which the results of these reports would or would not support a conclusion of reasonable assurance that system, structure, and component configuration and performance are consistent with the design bases. Additional!y, forty nine (49) Internal Audit Reports also were reviewed in a cimilar assessment.

Issues identified in the above reports involved the MOV, IST, Appendix R, and Appendix J Programs. The approach in correcting these issues was appropriate and aggressive. This is illustrated by the implementation of the ER system, which in itself has resulted in an increase in the number of scif-identified configuration and performance-related issues that have been identified and investigated in 1995 and 1996. We believe the ER process in conjunction with the recent strengthening of the Appendix R, IST, Appendix J, MOV, and other programs currently provide confidence and are expected in the future to provide increased confidence in the adequacy of conformance to and maintenance of the design bases.

A recent inspection (NRC Inspection Report 96-200) found that many items are being corrected at VY and that the ER process is capable of identifying problems. The report found that the ER initiation threshold for engineering issues was appropriate to support operations and maintenance, and the quality of the BMO process (to determine if continued plant operation could be justified following the identification of a nonconformance) effectively addressed the operability concerns presented by an ER condition. Analyses also were found to be detailed and to have properly evaluated safety concerns. However, the report identified weaknesses involving the overall ER initiation threshold, ER trending, and success in meeting ER timeliness goals. The ER process is described in detail in Section 7.0.

Our conclusion from review of the USNRC inspection reports and internal audit reports is that there is reasonable assurance that SSC configuration and performance are consistent with the design basis.

6.5 Summary l

l VY's procedures and processes for controlling plant configuration, operations, surveillance, l maintenance, and testing are designed and intended to maintain the plant's design bases. In l addition, recent program improvements and initiatives for Appendix R/ Fire Protection,

Equipment Qualification, Technical Specification Surveillances, IST, Appendix J, Maintenance Rule, Regulatory Guide 1.97, Technical Specification Improvement Project, improved 45 Enclosurs 1

1 Setpoint Program, Loss-of-Coolant Analysis, ECCS Single Failure Analyses, and FSAR  !

l update add confidence and reasonable assurance that the plant configuration and design .

bases are being maintained and that a questioning attitude is displayed in surfacing design I bases-related inconsistencies.

Earlier walkdowns and design modifications also have provided additional opportunities to review and evaluate how the design bases have been translated into the physical plant and how they have been maintained. These instances have resulted in identification of certain i discrepancies; however, they have provided objective evidence that our procedures and processes are typically effective and properly implemented. During the life of the plant, there have been a number of instances where it has been necessary to conduct a re-evaluation of programs, systems or processes. These re-evaluations often have resulted from changes in USNRC requirements, higher self and industry expectations, and, in some instances, from  ;

concerns identified internally or by the USNRC. The fact that such reviews were conducted, and subsequent improvements implemented, serves to heighten our confidence in the ability of the system or process to function as designed.

Our review of audits, assessments, and inspections plus our 50.54(f) response - specific ,

evaluations of engineering products support the conclusion that, while some discrepancies and areas for improvement have been identified, there is reasonable assurance that system, ,

structure, and component configuration and performance are consistent with the design bases. The Design Bases Documentation Program described in Section 9.1 is being implemented to strengthen this assurance.  ;

I

(

l I

46 Enclosure 1

7.0 RESPONSE TO USNRC TOPIC (d)

Topic (d)- Process for identification of problems and implementation of corrective actions, including actions to determine the extent of problems, actions to prevent recurrence and reporting to USNRC.

I 7.1 Introduction VY performed a comprehensive self-assessment of the corrective action process in 1994. A multi-disciplined team evaluated the existing program and benchmarked other industry programs.

1 Recommendations contained in the self assessment included the following improvements:

• Consolidate multiple existing processes.

• Implement a " tool box approach" to root cause assessment. l

• Improve trending capabilities.

• Establish performance indicators to monitor the program.

• Provide comprehensive training.

• Dedicate resources to coordinate the program. )

l The result of this effort was the development of a new and improved VY Event Report process which is specified in Procedure AP 0009 [ Event Reports]. Along with the improvement recommendations listed above, AP 0009 implements the requirements of 10 CFR 50, Appendix B, Criterion XV [ Nonconforming Materials, Parts, or Components], and Criterion XVI [ Corrective Action], as described in the Quality Assurance Program Manual (YOQAP-1-A).

7.2 Process for Identification of Problems A primary objective associated with implementation of the ER process has been to lower the threshold for identifying and reporting conditions. VY management has worked to instill an environment where employees feel free to aggressively demonstrate a questioning attitude in raising issues (regardless of significance) to higher levels of attention through the ER process.

Each employee, including contract personnel, is responsible for identifying and reporting problems and conditions adverse to quality or safe operation of the plant. Upon discovery of an issue, employees and contractors are expected to take immediate corrective actions (where appropriate) within their capability to mitigate the event.

The ER process is designed to handle a variety of conditions. AP 0009 specifies that ERs be generated in relation to:

• Operation, maintenance, surveillance testing, calibration or other activities that identify

! conditions outside proceduralized administrative limits that are indicative of installed plant equipment problems where function is impacted.

47 Enclosure 1

Radiation protection conditions.

Design-related conditions including but not limited to potential conditions outside the design bases, unanalyzed conditions, errors in approved calculations or calculational methods, etc.

Chemistry-related conditions.

• Fire protection conditions.

Non conformance conditions involving safety related systems, structures and components.

Programmatic conditions requiring correction.

Conditions resulting in an unplanned reduction of station generating capacity or unplanned loss of safety system capability.

Personnel safety conditions.

Nuclear safety conditions.

External findings (audits, surveillance's, violations).

Operating experience conditions that are determined to be applicable to VY and pose an immediate operability concem.

Human performance related conditions that resulted in or almost resulted in (i.e.,

near-miss) an actual mistake being made.

• Adverse trends.

• Materials-related conditions.

Environmental-related conditions.

In addition to the conditions listed above, employees and contractors are encouraged to generate ERs for other situations where they feel further investigation, management attention, or corrective action may be required. The ER process is designed to provide early identification of a variety of non-consequential or near-miss conditions in order to prevent more consequential events from occurring.

The ER process incorporates the guidance in SOER 92-01 [ Reducing the Occurrences of Plant Events through improved Human Performance). The General Employee Training program provides training on the ER process to personnel approved for unescorted access.

The primary process requirements are outlined below:

48 Enclosure 1

l l

7.2.1 Event identification and Processina i

Once an ER is initiated, it is routed to the initiator's Department Head. The respective Department Head coordinates the subsequent processing and evaluation of the Report. The Department Head is directed by AP 0009 to immediately contact the Shift Supervisor if there ,

, is an immediate operability concem or reportability concern. Operability concerns resulting )

4 from installed non-conforming plant equipment are to be addressed using the BMO Guideline '

to determine the impact of continued operation with potentially degraded equipment.

AP 0009 requires the Department Head to: 4 i  !

Review the event for immediate reportability and operability concerns and, if necessary, notify the Shift Supervisor.

Determine if the system, structure, or component will be considered operable with the ,

identified deficiency, assess the need to document the basis for considering the I component operable and the need to generate a formal BMO in accordance with the BMO Guideline. The BMO Guideline provides guidance consistent with the intent of Generic Letter 91-18 for performing immediate operability determinations.

Determine if review by the Shift Supervisor is required. Shift Supervisor review is required for an ER that is potentially reportable in accordance with AP 0156

[ Notification of Significant Events); requires other notifications; requires action on the

part of the Shift Supervisor such as entry into a Technical Specification Limiting Condition for Operation or potentially impacts plant operability or involves degradation of plant equipment.

Recommend an Event Level from Level 4 (less significant) to Level 1 (potentially safety significant requiring a higher level of management attention).

4

• Identify most probable cause for a Level 4 ER.

• Assess the need for early notification of the industry (via the INPO Nuclear Network 4 program) or the USNRC Resident inspector.

4 Forvard the ER to either the Shift Supervisor or the Events Screening Meeting.

AP 0156 [ Notifications of Significant Events) stipulates the Shift Supervisor determine the

required initial notifications. (AP 0156 lists the notifications required by 10 CFR 50.72 and other license conditions.)

7.2.2 Events Screening Meeting Following the above actions, ERs are screened at an Events Screening Meeting. This meeting is led by a senior member of the plant staff (Superintendent level or as designated by the Plant Manager). The Operations and Engineering Departments and other departments directly involved with the ER participate in the Events Screening Meeting.

49 Enclosure 1

4 AP 0009 stipulates the Events Screening Meeting: I Review initial notifications and assessments to verify they were appropriate.

l l -

i Assess the need for a reportability assessment in accordance with AP 0010  !

-[ Notifications and Reports Due). AP 0010 assesses the requirement to generate a LER per 10 CFR 50.73 or other special reports required by the Technical Specifications or i other license conditions. A formal assessment is performed by Engineering per i l AP 0010. AP 0010 provides detailed checklists for the reportability assessment and  ;

! the required reports. Reportability assessments are reviewed by the Technical Support .

I Manager and approved by the Plant Manager.

Determine if the issue is a potential Maintenance Rule issue and, if so, ensure the appropriate Event Level is assigned.

, Determine additional items to be addressed as part of the investigation.

I i Confirm or re-assign the Event Level using the criteria described below and assign a responsible Department Head to investigate the event. i l 7.2.3 Events Classification i The ER process has four event levels:  !

Level 1 Events - events that are potentially safety significant and require a high level l l of management attention. These events typically involve repetitive failures of  !

safety-related equipment, significant items of non-compliance, or items where there is a potential impact on nuclear, personnel, or public safety. A Root Cause Analysis (RCA)is required for Level 1 events.  ;

Level 2 Events - events that require superintendent level (the level above Department Head) attention due to significance or complexity, or because they involve multiple i departments. A RCA is required for Level 2 events.  !

• Level 3 Events - events that typically involve one department, have minimal impact l outside the department, and warrant additional investigation. Either an RCA or an Apparent Cause Evaluation (ACE) is performed based on complexity as determined by j the responsible Department Head.

• Level 4 Events - events where the documented actions taken are considered appropriate to address the issue, there is minor potential safety impact, and prior i

occurrences have been minimal. These events are entered by the Technical Support l Department into the ER database for trending and use in assessing repeat events. No further investigation is performed beyond recording the "most probable cause."

The flowchart on the following page illustrates the ER process.

i 50 Enclosure 1

i 1

1 Entry / Initiating l Condition '

i

2 1r

,'poj'" Any Employee / Contractor l

i DH Review / pproval:

. - Review BMO Guideline )

- Initiate immediate Department Head i 4 Corrective Action

- Approve

,r Reportability Determination:

- Operability SS (If Required)

- Notifications .

l ir ,

l.

Moming Meeting Potentially Reportable De . ' I : (50.73)? Reportability Evaluation gn i  ; ,

. Event Level Yes (Per AP 0010) j (copy)

- Reportability (50.73) ir LER/Special No (File copy)

Report orieinal Required?

Yes ir 1r Event Report to Use LER/Special Responsible Department -------+ Report Process for Processing (Root Cause Thru ER) ir ir ir Disposition:

Independent Reviews  : - PORC (Level i Only)  : Ter.hnical Support

- Management Approval ir Database and Trending ir Effectiveness Review 51 Enclosure 1

7.3 Actions to Determine the Extent of Problems 7.3.1 Event Investigation The Events Screening Meeting assigns responsibility for the event investigation to an appropriate Department Head. The Department Head subsequently assigns the actual event investigation to an individual who has completed training in the ER process. The Root Cause Analysis Guideline and associated Root Causa Analysis Training provide the details and training to qualify personnel to perform Root Cause Analysis and apparent cause evaluations.

The event investigation is performed to:

• Develop the problem statement.

• Develop the event description.

Determine the root cause or apparent cause using the Root Cause Analysis Guideline.

The Guideline includes provisions for data collection, personnel interviews, Barrier Analysis, Change Analysis, and Event and Causal Factor Charting.

Determine if the event resulted in a Maintenance Rule Functional Failure (MRFF).

• Determine the potential for similar conditions at the plant.

Search for repetitive events by reviewing past W events and industry operating experience sources. l l

• Identify immediate and long term corrective actions.

• Determine the need to perform an assessment of the corrective actions following implementation to evaluate their effectiveness and to consider the need for additional action. l l

7.3.2 ER Review and Acoroval Review and approval requirements for the ER are based on Event Level as described below:

• Level 1 ERs are reviewed by Quality Assurance and the PORC, and approved by the Department Manager, Superintendent Level, Plant Manager, and Vice Presiderit -

Operations. Level 1 ERs also are subsequently reviewed by the Nuclear Safety Audit and Review Committee (NSARC).

• Level 2 ERs are reviewed and approved by the Department Manager and l Superintendent Level.

l

• Level 3 ERs are reviewed and approved by the Department Manager.

I l

52 Enclosure 1

Level 4 ERs are dispositioned by the Department Manager and are provided to the Technical Support Department for trending. No subsequent review is required.

The assigned Department Manager is required to review Level 1,2, or 3 Event Investigation Reports to check that the following elements are addressed in the report:

Root and contributing or apparent causes.

Maintenance Rule Fenctional Failure determination.

• Potentially similar conditions.

Search for repetitive events if the event is determined to be repetitive, the report is to j include an assessment of why previous corrective action failed to prevent reoccurrence.

Corrective action. I Follow up verification.  !

The Department Manager must check that appropriate input was received from Departments l that will be assigned corrective action commitments. The Manager also is required to specify i that coordinators of plant programs and additional reviewers, as appropriate based on the specific type of event, review the adequacy of the investigation and corrective actions. i The Quality Assurance Department is required to provide independent oversight by reviewing Level 1 ERs to determine if the elements of the Root Cause Analysis are addressed in the Event Investigation Report and are adequate to prevent the event from reoccurring. The  ;

Quality Assurance Department also is required to verify that appropriate corrective action has  !

been taken or is planned.

l The Department Manager is required to route a copy of an ER involving a design deficiency j or a "use-as-is" disposition to Design Engineering for review under Engineering Instruction l WE-201 [Non-Conformance Reports).

After review comments are resolved, the originator is required to obtain management and '

PORC approvals, as required, based on the severity level of the ER. The approved ER is forwarded to the Technical Support Department where corrective action commitment tracking is initiated under the AP 0028 [ Operating Experience Review and Assessment / Commitment ,

Tracking) process. '

7.4 Implementation of Corrective Action Identification and Assianment of Corrective Action  !

The event investigation is required to identify and document: i 1

53 Enclosure 1 l

l

• Root and contributing causes for Level 1 and 2 ERs and for Level 3 ERs where a Root Cause Analysis was perfomied. (Apparent causes are identified for the other Level 3 ERs.)

• Immediate corrective actions that were taken in response to the event along with supporting evidence for corrective actions or related evaluations that have been completed.

• Additional corrective actions to prevent reoccurrence.

• Corrective action recommendations from the reportability evaluation.

• The Department assigned to implement each corrective action, as well as when the corrective action is schedu!ed to be completed. (in developing the corrective actions, concurrence by the assigned manager (s) is required to be obtained under AP 0028.)

• Corrective action that is required to be completed prior to retuming equipment to service.

Specific follow up verifications required to ensure the corrective action was effective in preventing reoccurrence, as well as the Department responsible for the follow up verification, and the follow up verification schedule. ,

l The Department Manager (under whose responsibility the event investigation was performed) l must check that the required elements have been included or considered in the event investigation and appropriate input has been received from those Departments that will receive action item commitments as part of the recommended corrective actions.

The Department Manager also is required to designate additional reviewers (such as coordinators of affected programs or other Departments based on the type of event) to review the adequacy of the event investigation and corrective actions. Quality Assurance Department review must be obtained for Level 1 ERs to determine that appropriate corrective action has been taken or is planned and will be adequate to prevent reoccurrence.

Trackina of Corrective Action Corrective actions must be tracked in the Commitment Tracking System, which is implemented under AP 0028 [ Operating Experience Review and Assessment / Commitment Tracking], to confirm the corrective actions are taken by the appropriate personnel and are tracked to completion. When commitment due dates were assigned in the event investigation, they were to have been confirmed with the assigned Department Head such that the scope and due dates were well understood.

l The corrective action is required to be entered into the Commitment Tracking System as a l

commitment with an assigned responsible Department Head and due date. Administrative t

approval levels are specified to control changes to an assigned commitment or commitment due date.

54 Enclosure 1

Level 1 ER corrective action commitments that require additional actions in accordance with a work order, design change, etc., are required to remain open until the work order is closed or the design change is installed.

1 If a Level 2 or 3 ER corrective action commitment is based on proposed actions that are not tracked by another commitment, the proposed action must be tracked in another process (e.g. Design Change List, Maintenance Planning and Control (MPAC) database, AP 0037 (Plant Procedures) change form) with specific direction that the proposed actions are not to be canceled without approval of the Department Manager who closed the original commitment.

1 Administrative approvallevels also are specified to control review and approval of the closure of assigned corrective action commitments. The closure review will determine either that:

The item is complete and no further tracking is required, or Additiond tracking is required and is to be performed in accordance with additional l AP 002G commitments and due dates, or Additional tracking is required by a process other than AP 0028 (e.g., Design Change l List, MPAC, AP 0037 change forms). i 7.5 Trending and Functional Area Assessment Level 1 through Level 4 ERs are required to be entered into the ER database by the Technical Support Department. ER trending is performed by the Technical Support Department and by key functional departments during Functional Area Assessments (FAAs).

The Technical Support Department distributes Trend Reports and may generate subsequent ERs as a result of trends that have been identified.

i FAAs are detailed self assessment of key functional areas and are required by the Quality Assurance Program (YOOAP-1-A). The FAAs are developed based on the guidance provided in AP 6005 (Functional Area Assessment Development].

7.6 Reporting to USNRC The initiation of an ER in accordance with AP 0009 sets in motion the process of initial and follow up confirmatory reviews, as applicable, by the originator, the originator's Department Manager, Shift Supervisor, Events Screening Committee, Engineering, PORC, and plant management for operability and reportability determinations. Plant procedures for event investigation and reportability determinations are designed to promptly and consarvatively screen events and determine reportability as described below in relation to 10 CFR 50.72

[Immediate Notification Requirements for Operating Nuclear Reactors] as well as other reporting requirements.

Administrative Procedure AP 0156 [ Notification of Significant Events) requires that conditions be evaluated for reportability to the USNRC pursuant to the provisions of 10 CFR 50.72. An ER is initiated in accordance with AP 0009 to notify the originator's Department Head and the 55 Enclosure 1

=

Shift Supervisor of the event. Procedure AP 0156 and the Event Notification Worksheet Form stipulate notifications of declaration of any of the Emergency Classes specified in the Emergency Plan; 1-hour nonemergency notifications; 4-hour nonemergency notifications; etc.

AP-0156 Appendix A specifies 1-hour notifications for seriously degraded conditions, unanalyzed conditions, and situations outside the design bases. Appendix B specifies situations subject to 4-hour notifications. Appendix C provides additional clarification of the Appendix A and B categories and specifies notification of events where a safety system or structure could have failed to perform its intended function because of personnel errors, equipment failures, or deficiencies (design, analysis, fabrication, construction, or procedural).

Administrative Procedure AP 0010 [ Notifications and Reports Due] requires review of ERs to  :

determine the reporting requirements for:

Reportable occurrences per 10 CFR 50.73 [LER System].

Defect or non-compliance or other reportable information per 10 CFR Part 21

[ Reporting of Defects and Noncompliance].

Miscellaneous reportable events or other reportable information per 10 CFR 50.9

[ Completeness and Accuracy of Information].

• Other notifications to meet license conditions.

The responsible Company officer is required to be notified if an event is identified as potentially reportable per 10 CFR Part 21 and to have a formal assessment completed in accordance with Technical Administrative Guide (TAG) No. 6 [10 CFR Part 21 Reporting) which provides the process for evaluating events or conditions which may represent a substantial safety hazard in regard to 10 CFR Part 21.

Administrative Procedure AP 0009 requires that a copy of the ER be routed to Engineering for reportability assessment under AP 0010. Administrative Procedure AP 0010, Appendices G, H, and I stipulate notification and reporting requirements for radiological events per 10 CFR 20 [ Standards for Protection Against Radiation Subpart M - Reports). Appendix J [LER Review Checklist] stipulates the assessment include a summary of the design bases and safety functions for the system, structure, or component, and a statement of how the conclusion was reached that no safety consequences resulted from the event.

AP 0010, Appendix K [Reportability Checklist] provides a checklist-type summary of the reporting requirements per 10 CFR Parts 21,50.9,50.36(c)1 regarding Safety Limits and Limiting Safety System Settings,50.36(c)2 regarding Limiting Conditions for Operation, and 50.73 regarding the Licensee Event Report (LER) reporting system.

56 Enclosure 1

l 4

8,0 RESPONSE TO USNRC TOPIC (e)

Topic (e) - The overall effectiveness of the current processes and programs in concluding that the configuration of the plant is consistent with the design bases.

8.1 Introduction We believe our current processes and programs are sufficiently effective to provide reasonable assurance that plant configuration is consistent with the design bases. This conclusion is based both on the structure of our programs and processes, which include l configuration control elements derived from USNRC requirements and industry practice, and on the overall results of audits, inspections, and assessments.  ;

l 8.2 Procedures and Programs Our configuration control procedures and programs are described in Section 4.0. These procedures and programs are conducted under the Vermont Yankee Operational Quality Assurance Program (YOOAP-1-A) and are employed to perform design activities in a planned, controlled, orderly, correct, and documented manner. As discussed in Sections 5.0 and 6.0, a number of audits, inspections, and assessments have provided us confidence and reasonable assurance that design bases requirements are translated into operating, maintenance, and testing procedures and that system, structure, and component configuration and performance are consistent with the design bases. Section 7.0 also describes the Event Reporting and corrective action processes. Significant improvements were achieved by the introduction of a lower reporting threshold, an easier to use ER process, and an environment where employees feel free to aggressively demonstrate a questioning attitude in raising issues to higher levels of attention through the ER process.

8.3 Recent Program Initiatives Our recent or ongoing enhancement of the overall effectiveness of processes and programs is indicated by the recent success of the initiatives described in Sections 5.2 and 6.2. For example, the Appendix R/ Fire Protection, MOV, IST, and Appendix J Programs have all undergone significant upgrades resulting in improved documentation and control of the design bases. The ITS Project discussed in Section 6.2 is another example where improved i documentation of design bases information currently is being produced. VY management is I committed to improvement in the documentation of our design bases and considers this a key )

ingredient to our configuration control processes. As an example, we have a Design Basis Documentation Program underway in addition to the improved Technical Specifications Project as described in Section 9.2.

8.4 Review of Inspections, Audits, and Event Reports Sections 5.3.2 and 6.4.3 summarize the results from recent inspections, audits, and ERs.

These summaries identify the strengths and certain weaknesses related to the overall 57 Enclosure 1 l

l

effectiveness of our configuration control process. Overall these summaries indicate reasonable assurance that configuration control is being effectively maintained.

A recent NRC inspection (NRC Inspection Report 96-200) determined that the ER process successfully has resulted in the identification and correction of problems at VY In addition, the ER initiation threshold demonstrated by Engineering personnel appeared appropriate to support operations and maintenance; the quality of the BMO process was found to effectively address the operability concems presented by a condition; and root cause analyses were found to be generally detailed and to address the stated concerns. However, certain weaknesses of the program, including the overall initiation threshold and the timeliness with which corrective actions are taken, were identified. VY is taking the necessary actions to evaluate and address the negative findings and to implement appropriate revisions to the process and/or to employee training in a timely manner.

The nature of the findings from recent intemal and external audits and inspections has re-emphasized the need for VY management to stress the importance of the shared responsibility between VY and our personnel for continuous improvement, safety, professionalism and leadership. VY will continue its evaluation of the findings resulting from the recent internal and external audits including those documented by the Response Team as discussed in Section 10.0.

I 8.5 Response Team Assessments Appendix A provides summaries of targeted validation assessments performed by the Response Team. Engineering processes and products were sampled for evaluation. This evaluation found that the calculation process was a strength; the EDCR and safety evaluation processes were adequate with recommendations of areas ter improvement; and the minor and temporary modification process was considered to be weam An ER was prepared by VY based on the Response Team's assessment of the need to clarify independent review requirements for minor and temporary modifications.

The EDCR procedure (WE-100) was reviewed by the Response Team to key elements based on the criteria in ANSI N45.2.11. This review determined that the procedure contains each of the elements necessary to control design changes.

EDCR closecut documentation was reviewed. The assessment of closeout documentation ,

from 1995 and 1996 indicates that improvements to the Job Order closeout requirements have resulted in effective closeout documentation.

LERs reviewed indicate that a questioning attitude exists during the investigation of events as well as during normal day-to-day activities. A separate corrective action review was performed by reviewing recent ERs, USNRC inspection reports, and intemal audits and assessments. Concerns identified by the Response Team included timeliness of resolution, tracking of commitments, and adequacy of previous corrective actions. Although these concerns are supported in part by the results of the USNRC Inspection Report 96-200, the Response Team concluded that the Event Report and corrective action processes are adequate when considered in conjunction with Vermont Yankee's continued emphasis to aggressively identify, evaluate, and resolve issues.

58 Enclosure 1

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The Response Team's targeted assessments indicate that:

VY should continue to improve configuration control processes and programs, which we are doing.

The current questioning attitude and Event Reporting threshold are helping to identify areas for improvements but the efforts should be increased, which we are doing throughout the organization.

There is reasonable assurance that configuration of the plant is being adequately maintained consistent with the design bases.

8.6 Summary and Conclusion The information summarized in this section and discussed elsewhere in this response supports VY's conclusion of reasonable assurance that the design and configuration control processes currently in use provide effective mechanisms for translation of the design bases into the plant design, programs, and procedures.

As discussed in Sections 5.0 and 6.0, the major upgrades to engineering programs plus the other initiatives, assessments, walkdowns, design and operational evaluations, and design changes throughout the lifetime of the plant have continued to benchmark the configuration, performance and maintenance of SSCs against the design bases. The incremental and cumulative outcomes of these activities have been to increase the accuracy and documentation of the design bases requirements versus the physical plant.

As discussed in Section 7.0, VY procedures are designed to efficiently identify problems and implement corrective actions. When potential conformance to design bases issues are identified, these issues are required to be processed to the appropriate levels of evaluation, review, approval, and resolution. As future questions arise in relation to the physical plant conformance with design bases requirements, the existing procedural controls for engineering design, configuration control, problem identification, and corrective action are in place to assess and maintain the physical plant configuration and performance in accordance with design bases requirements.

During the lifetime of the plant, there have been a number of instances where improvements have resulted from vertical slice assessments, engineering program reviews, self assessments, inspections, and other activities. Given the large number of SSCs in a nuclear ,

power facility, instances of weaknesses in configuration control aie expected to occasionally I be identified. VY believes that sound design basis documentation and strong configuration l control implementing procedures are key to maintaining the plant consistent with design l bases. Section 10.0 identifies VY's current plans related to improvements in design basis l documentation and configuration control processes.

VY's overall conclusion is that there is reasonable assurance of the effectiveness of the current processes and programs which maintain the configuration of the plant consistent with .

the design bases. Areas for improvement have been identified and will be addressed. As l 4

59 Enclosure 1

future quections arise in regard to the design bases requirements, effective procedural  ;

controls are in place to evaluate and disposition discrepancies.

1 i

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l 60 Enclosure 1

9.0 RESPONSE TO USNRC REQUEST FOR OTHER i INFORMATION Indicate whether you have undertaken any design review or reconstitution programs, and if not, a rationale for not implementing such a program. If design review or reconstitution programs have been completed or are being conducted, provide a description of the review programs, including identification of SSCs, and plant-level design attributes (e.g., seismic, high energy line break, moderate-energy line break). The description should include how the program ensures the correctness and accessibility of the design bases information for your plant and that the design bases remain current. If the program is being conducted but has not been completed, provide an implementation schedule for SSCs and plant-level design attribute reviews, the expected completion date, and method of SSC prioritization used for the review.

9.1 Design Bases Documentation Program VY currently is implementing a DBD Program as described below.

Proaram Scoce and Schedule The twenty-three (23) systems indicated below are presently in the scope of the DBD Program:

Automatic Depressurization System Reactor Protection System

• CS System Control Rod Drive System Service Water /RHRSW/ACSs

• 4 kV/480 V AC System

+ 125 V DC System

• RHR System Primary Containment Pressure Suppression System Primary Containment isolation System a Nuclear Boiler System

• Main Steam System a EDG and Auxiliaries Standby Gas Treatment System / Secondary Containment

= 24 V DC System

• 120 V Vital AC System

• Standby Liquid Control System

• HPCI System

• Reactor Core Isolation Cooling System

. Reactor Building Closed Cooling Water System

• Feedwater/ Condensate System

• Nitrogen inerting/ CAD /PCAC

• Turbine Building Closed Cooling Water System 61 Enciosure 1

l l

l These systems represent the high and medium risk systems identified using the VY Individual Plant Examination (IPE) models. The system DBDs will define the system functional and l performance requirements as well as associated component level requirements. The I Program is designed to capture and organize information to provide effective and efficient use of the current VY design, operational, and licensing bases. The major benefits of this effort are expected to result in documented and easily retrievable design bases information, l

improved operational safety through clearer understanding of the current and as-licensed requirements, and long term assurance of control and maintenance of the design and licensing bases.

l The process for developing and implementing the system DBDs is planned in two phases, the l Development Phase and the Validation Phase. The Development Phase includes: I

=

Preparation of procedures for development, control, and use of the DBDs.

Preparation of the DBDs including identification and management of open items such as missing or discrepant information.

1 Independent review / verification of the DBDs including verification that appropriate source documents are referenced and that the scope of the design bases requirements identified is complete.

In addition to independent review, supplemental reviews by other engineering disciplines and VY departments, as applicable, will be performed to enhance the level of confidence in the l' accuracy and completeness of the DBDs. After completion of independent review / verification, the DBD will be issued for use.

After the DBD is issued for use, the Validation Phase will begin and this will include a separate validation review of the DBD. The DBD validation review is designed to provide reasonable assurance that the design bases information is consistently reflected in both the physical plant and in the controlled documents (including operating, surveillance, test, and emergency operating procedures) used to support plant operation. The validation process will be based on SSFI techniques which will focus on the functionality of the system as related to the design bases. The results of the validation will be documented in a report.

t Resolution of Ooen Items Open items are expected to be identified during the development and validation phases.

They may consist of discrepancies such as system, structure, or component design information that conflicts or is inconsistent with either the as-built condition of the plant or with other design documents and may include items such as FSAR errors, procedural omissions, specification inconsistencies, etc. Open items also may result from missing information which is required to support the design bases. The process for resolution of open items is addressed in the Writers Guide and the Validation Guideline. Open items with potential operability or reportability concems will be processed in accordance with the VY ER process.

62 Enclosure 1

Toolca! Reoorts (Plant Level Design Attributes)

VY presently maintains topical reports for the following issues:

Environmental Qualification of Electrical Equipment 10 CFR 50 Appendix R Regulatory Guide 1.97, Post-Accident Monitoring IST/ Component Testing

• Electrical Separation a Generic Letter 89-10/MOVs

• Inservice Inspection

• Appendix J Leak Rate Testing Additional System DBDs and Toolcal Reoorts VY may expand the :urrent scope of system DBD and Topical Reports based upon the results of assessments during DBD development efforts.

9.2 ITS Project VY is in the process of implementing an ITS Project as described in Section 6.0.

Project Scoce and Schedule The project is confirming, identifying, or developing and documenting design bases for the setpoints and specifications in the ITS. In situations where calculations or analyses to adequately document these design bases are not located, the bases are being reestablished through performance of new calculations, analyses, etc.

The ITS development phase began in June,1996, and will continue until the specifications are submitted to the USNRC, which is currently scheduled for September,1997. The implementation phase will continue until the specifications are approved by the USNRC and are in effect at VY.

The bases for a numerical value in the ITS will be verified to have a reference. Other values (numbers) or statements in the Technical Specifications and the Technical Specification Bases will be evaluated to confirm the values or statements are correct and the capability exists for adequate monitoring.

The ITS Project involves several associated initiatives to assemble or verify design bases information. These include:

• Instrument Setpoint Project.

• Improvements in the Technical Specifications Bases.

• Verification of Technical Specifications Bases against the design bases contained in the FSAR, SER, other design documents, and existing design bases documentation.

63 Enclosure 1

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• Documentation of the bases for equipment Allowable Out of Service Times and extended Surveillance Test Intervals. i

• Re-evaluation of limiting and non-limiting transients and associated setpoints.

t

• Recalculation of torus volume and temperature requirements. i t

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f 64 Enclosure 1

10.0 COMMITMENTS The information described in Sections 1.0 through 9.0 of this document is not intended to create new commitments, unless explicitly stated otherwise below, and is not intended to preclude future procedural or process changes that would be made following the normal VY l

procedures and practices. All commitments made in response to the referenced USNRC letter are explicitly stated in this section.

10.1 Design Bases Documentation Program VY is developing and implementing a DBD Program. The currently planned scope and schedule are provided in Section 9.1, but could change as work progresses or as issues emerge.

10.2 ITS Project VY is developing an ITS Project. The currently planned scope and schedule are provided in Section 9.2.

10.3 FSAR Verification VY will perform a verification of the FSAR which is targeted for completion in 1998. This program will be designed to determine and verify the internal consistency of the FSAR and the manner in which the FSAR information conforms to similar information in other plant documentation. The scope and schedule associated with this commitment are intended to be consistent with the USNRC's recently announced discretionary moratorium in its Enforcement Policy.

10.4 Follow-Up Activities to this Response VY will further review insights gained from assessments and reviews made during  !

preparation of this response. A number of potential areas for improvement and further assessment have been identified. VY will evaluate the issues and document findings and recommendations in a formal configuration management self assessment report that is expected to be completed by June 1,1997.

65 Enclosure 1

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I APPENDlX A i l

! RESPONSE TEAM ASSESSMENTS i 4

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l 66 Appendix A

Introduction to Appendix A  ;

The Response Team performed a number of targeted validation assessments. The purpose )

of the process validations is to provide a performance indicator, based on a sample of engineering products, inspections, and reports.

The majority of data reviewed was produced during the past two years and supports the l overall conclusions drawn in the response to USNRC Topic (e) (Section 8.0) regarding the ,

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

Summaries of the following targeted validation assessments are provided in this Appendix: j A.1 Review of Engineering Processes and Products A.2 Review of the Design Change Procedure A.3 EDCR Closecut Documentation Review A.4 LER Review A.5 Corrective Action Review l

)

67 Appendix A

j A.1 Review of Engineering Processes and Products Scooe and Acoroach The purpose of this evaluation was to review the major design engineering processes and a sample of engineering products developed under these processes. The selection of the engineering products for review was accomplished by reviewing the lists of the recent engineering product titles. The reviewers selected the time frame for the products to be reviewed as those produced since 1989. The reviewers exercised their professional l judgement in selecting the engineering products which appeared to have a safety significant  ;

scope.

)

The following processes were reviewed:

• 10 CFR 50.59 Safety Evaluation Process

• Design Change Process

• Calculation Process Each process was evaluated for compliance with the applicable regulatory requirements. The sample of engineering products generated under these processes was reviewed for compliance with the goveming procedure (s) and overall soundness of the product.

Conclusions The conclusions from this evaluation provide reasonable insights into and an appropriate level of confidence in the validity and quality of engineering work and include:

• The review validated the Unreviewed Safety Question (USO) determinations or conclusions regarding impact on safety that had been documented within the individual products. The comments from the review were dispositioned by means of discussions and/or clarifications with Engineering personnel.

1

• The calculation process was considered to be a strength.

• The overall modification process was considered to be adequate. Improvements were recommended to the process.

• The safety evaluation process was considered to be adequate. Improvements also were recommended to this process. i

• The minor and temporary modification process was considered to be weak and to lack

, adequate guidance. However, the minor modification that was reviewed was found acceptable.

I i A.2 Review of the Design Change Procedure Design Engineering Instruction WE-100 (EDCR] was reviewed by the Response Team against key elements based on the criteria for design change control in ANSI N45.2.11. The 68 Appendix A l

.m _ .- . _ _ _ _ _ _ _ _ _ _ . _ . _ . . _ _ _ _ _ _ . _ - _ _ _ ,

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! conclusion from this review is that each of the key elements to provide design control is evident in the procedure.

l Of particular interest was whether the procedure requires:

• Review for conformance with the design bases.

-

• Cumulative changes be reviewed to be consistent with the design bases.

The impact of changes on procedures, operations, maintenance and training l programs be evaluated.

• Changes be reviewed to 10 CFR 50.59 requirements.

l' Each of the above attributes, as well as eleven (11) additional attributes, was considered.

Each of the fifteen (15) attributes was found to be satisfactorily addressed in WE-100. The i assessment concluded that WE-100 provides adequate guidance for development, i

implementation, review, and acceptance of design changes to plant safety related and non-l safety related equipment. It also provides reasonable assurance that controls are such that j design changes will be made and controlled consistently with the design bases.

i 1 A.3 EDCR Closeout Documentation Review Quality Assurance Department (OAD) review is performed as a requirement of WE-100. A i review of eighteen (18) QAD Surveillance Reports (SR) documenting the closeout of EDCR

! packages was performed. Selected EDCR closeout activities performed during the 1993 to l 1996 period were reviewed.

i This review revealed that documentation and administrative issues had been identified during

the closeout review of EDCR packages in 1993 and 1994. Improvements in these areas were j noted in the closeout review findings beginning in 1995.

Discussion with cognizant personnel revealed that a weakness existed in the interface between WE-100 and AP 6022 (Job Order Files) conceming the information to be contained  !

j in the Job Order (JO) package for Job Order closecut. Procedure AP 6022 was later modified l 1

to provide additional guidance to VY personnel concerning the documentation to be retained  !

i within the Job Order file. Currently, the specific items are identified that are to be contained in I

a Job Order package that is considered ready for closeout. As noted above, closeouts of

- EDCR Job Order packages have disclosed fewer documentation problems during 1995 and i 1996. Based on the above, it was evaluated that implementation of the EDCRs that were 3

reviewed is in accordance with the guidance documents. Thus, there is reasonable i assurance that, while certain discrepancies and areas for improvement had been identified  :

1 and were being addressed through the corrective action process, these processes and  !

programs are effective in maintaining the configuration of the plant in accordance with the  !

design bases.

i

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69 Appendix A

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4 A.4 LER Review The Response Team wpled eighty seven (87) LERs from 1986 to 1997 to review the extent to which even? that potentially may be design bases-related are identified and reported. It was determined from this review that VY programs, procedures, and self-assessment initiatives emphasize a questioning attitude in the investigation of events as

well as in normal day-to-day operations, engineering, and other support activities. As an j example in 1995 and 1996, twenty two (22) LERs were considered potentially to be related to design bases or configuration control. The preparers of these twenty two (22) LERs determined that the LERs did not have a safety significance in regard to the health and safety of the public. These LERs also were evaluated in this Response Team review not to have a safety significance. They originated from degraded er potentially degraded conditions as opposed to failures of SSCs. Twenty (20) of the twenty two (22) LERs were self-discovered

through assessment activities rather than having been initiated as a result of an occurrence j or event.

l Additionally, fifteen (15) of these twenty twa (22) LERs were self-discovered through comprehensive reviews of three major programs or initiatives. Seven (7) were related to fire protection and Appendix R programs; five (5) to ISI and IST programs; and three (3) to the 4 Appendix J leak rate testing program. These three major programs and initiatives have undergone extensive recent review, evaluation, and upgrade that contributed to the identification and correction of these issues.

A.5 Corrective Action Review The Response Team sampled aspects of the VY corrective action process. The evaluation included the: i

• ER process described in Administrative Procedure AP 0009.  !

USNRC Systematic Assessment of Licensee Performance (SALP) Report No. 50-271/95-99.

1995 and 1996 USNRC Inspection Reports.

• 1995 and 1996 internal audits and assessments.

A total of sixty six (66) documents were reviewed. Fifty (50) of the sixty six (66) documents included references to corrective actions, typically involving follow-up evaluation or inspection of corrective actions resulting from a previous inspection or assessment report.

It was determined that thirty three (33) documents had indicated previous corrective actions to be appropriate. Seventeen (17) documents to some degree had reported negatively on previous corrective actions. Common threads in the negative assessments included timeliness of implementation of corrective actions, tracking of commitments, and, in certain instances, adequacy of previous corrective actions. Although these concerns are supported in part by the results of USNRC Inspection Report No.96-200, the corrective action process 70 Appendix A

. . .~ . . _ , - . - . - . ... - - . _ - . . .. - - . . . . -. - _

l l

. is evaluated to be adequate when considered in combination with VY's continued emphasis I to aggressively identify, evaluate, respond to, and resolve issues.  ?

l l

l l

l I

\

l l

l 71 Appendix A

l ACRONYMS  !

ACS Alternate Cooling System BMO Basis for Maintaining Operation CDC Critical Design Characteristic DBD Design Bases Document i DC Direct Current j ECCS Emergency Core Cooling Systems ]

EDCR Engineering Design Change Request j EDG Emergency Diesel Generator EDSFl Electrical Distribution System Functional Inspection  !

EOP Emergency Operating Procedure ER .

Event Report FSAR Final Safety Analysis Report HPCI High Pressure Coolant injection HVAC- Heating, Ventilation, and Air Conditioning l&C . Instrumentation and Controls i

IASD Input Assumptions Source Document  !

IE Inspection and Enforcement l lST Inservice Testing ITS Improved Technical Specification I LCO Limiting Conditions for Operations '

LER Licensee Event Report LOCA Loss-of-Coolant Accident MCC Motor Control Center MOV Motor-Operated Valve PORC Plant Operations Review Committee RHRSW Residual Heat Removal Service Water SER Safety Evaluation Report SSCA Safe Shutdown Capability Analysis SSCs Systems, Structures, and Components SSCs Structures, Systems, and Components SSFl Safety System Functional Inspection SWOPl Service Water Operational Performance Inspection USI Unresolved Safety Issue USNRC U.S. Nuclear Regulatory Commission VY Vermont Yankee Nuclear Power Station 72 Acronyms

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