ML20213D197

Forwards Summary of 800401-02 Meeting W/Savannah River Lab, Util,Ge & Burns & Roe in Richland,Wa Re FSAR Changes Resulting from Applicant Rewrite of Chapter 7 & to Conduct Site Tour of Facility

ML20213D197
Person / Time
Site:	Columbia
Issue date:	04/15/1980
From:	Szukiewicz A Office of Nuclear Reactor Regulation
To:	Satterfield R Office of Nuclear Reactor Regulation
References
CON-WNP-0301, CON-WNP-301 NUDOCS 8005070951
Download: ML20213D197 (52)

Text

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! MEMORATIDUM FOR: Rodney it. Satterfield, Chief 1 Instrumentation & Control Systems Branch, DSS THRU: Thomas G. Dunning, Section Leader Instrumentation & Control Systems Branch, DSS l

1 FROM: An' drew J. Szuktewicz, Instrumentation & Control Systems Branch, DSS I

EUBJECT:, '

SUMMARY

OF MEETIrlG WITH WASHIttGTON PUBLIC POWER C0f'PAflY (WPPSS Uf1IT 2)

A meeting was held in Richland, Washington at the WPPSS Unit 2 site on April 1 and 2,1980 with the.flRC, Savannah River Laboratory, Washington l Public Power Company, General sElectric, and the Burns and Roe representatives.

[ A list of attendees is enclosed (Enclosure 1).

i The purpose of.the meeting was 1) to familiarize the staff and the Instrumentation l and Control reviewer with the changes in the FSAR resulting from the applicant's re-write of Chapter 7, and 2) to conduct a site tour of Unit 2.

) Enclosure 2 discusses the highlights of the meeting and describes some of the

, concerns identified by the staff. Enclosures 3, 4, and.5 are copies of the transparencies used by the applicant for his presentation. Enclosure 6 is a

preliminary draft of tha applicant's separation criteria for Unit 2.

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i Andrew J. Szukiewicz

! Instrumentation & Control Systems Branch Division of Syst' ens Safety Enclosures : DISTRIBUTI0fl:

As stated Central ICSBReadingIile File '( M cc; V. Moore ~ AJSzukiewicz P. Check T. Dunning D. Sullivan ,

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ENCLOSURE 1 LIST OF ATTENDEES Keener Earle WPPSS W. E. Faiaone WPPSS Robert Green WPPSS Gordon Brastad WPPSS Pat Powell WPPSS Dusty Rhodes WPPSS J. Ellwanger Burns & Roe R. Pifferetti General Electric G. Darmohray General Electric Fred MacLean General Electric Art Hadden E. I. duPont (SRL)

Mel Goosey E.I.dePont(SRL)

Dave Lynch NRC/DPM A. J. Szukiewicz NRC/ICSB S

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ENCLOSURE 2 WPPSS UNIT 2 I MEETING APRIL 1 - 2 -

i RICHLAND, WASHINGTON  :

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1. Enclosure 4 summarizes the changes made to Chapter 7 of the FSAR. The staff expressed concern that Chapter 15 would not be updated and available to the staff when the new Chapter 7 is formally transmitted for NRC review.

Delaying the submittal of a revised Chapter 15 (that will be consistent with the new Chapter 7) could have a significant irrpact on the review process. WPPSS was requested to expedite a revised Chapter 15 and advise NRC of the schedule for its issue. It was also noted that the internal cross referenceswould not be corrected in the initial submittal of the new Chapter 7, but will be corrected on some future amendment. WPPSS agreed to advise the staff on a schedule when these corrections would be incorporated.

2. To expedite the review process and to help the SRL reviewer in resolving minor questions that could be resolved verbally, Mr. Gordon Bastad of WPPSS' instrumentation and control section was appointed as the direct contant liaison.

3. During the plant tour, the staff expressed concern regarding the adequacy of WPPSS' separation criteria which allows mixing IE and non-IE (non-power carrying) cables. Since this criterion represents a significant deviation from currently accepted practice, the design should be reviewed in depth and the adequacy of such a deviation justified, or the design modified to provide an acceptable separation. Enclosure 6 provides the details of WPPSS separation criteria.

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4. There is no redundancy of safety related equipment on the remote shutdown panel. The parameters located on the panel could not be viewed because of construction covers. We expressed concern over the lack of redundancy an'd indicated that the design did not satisfy current requirements and that an indepth review of this design was warranted.

5. WPPSS briefly described the organization that reviews and is responsible for equipment qualification. They plan to verify their qualification programs (both envimnmental and seismic) with a staff of 6 engineers.

WPPSS indicated that by mid 1980 their qualification methodology programs would be finalized and requalification phase (if required) would begin by the end of 1980. It was estimated that the SQRT (Seismic Qualification Review Team) visit could 'be conducted in February 1981.

6. The applicant stated that Chapter 3 of the FSAR (Sections 3.10 and 3.11) is in the process of being revised. WPPSS agreed to submit a schedule for the revision and submittal of the qualification information of Chapter 3.

7. The applicant was given a copy of draft questions that were developed by the SRL reviewer. These questions were also s'ubmitted to the staff via letters from C. P. Ross to.R. Satterfield dated December 19, 1979, January 25, 1980, March 11, 1980, and March 19, 1980, respectively.

WPPSS was advised that some of the questions may have been resolved as a result of the Chapter 7 re-write, however, it was prudent to review those questions and incorporate their resolution to as many as possible in this '

re-write, before the formal transmittal is submitted to the staff.

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WPPSSMIRC I&C MEETING APRIL 2, 1980 INTRODUCTION

. EARLE SR. LICENSING ENGINEER, WNP-2 PURPOSE

1. INTRODUCTION TO NEW CHAPTER 7

2. PLANT IOUR AND FAMILIARIZATION

3. SUMMARIZE STATUS OF SIGNIFICANT l&C OPEN ITEMS

4. QUESTIONS AND ANSWERS OKE 3/31/80

INTRODUCTION TO CHAPTER 7

1. WHAT DID WE D0?

2. WHY?

ANSWER IN THE CHRONOLOGY e

OKE 3/31/80

CHAPTER 7 CHRONOLOGY 3/15/77 FIRST SUBMITTAL OF FSAR  !

6/24/77 NRC REJECTION O'F FIRST FSAR PRIMARY REASON WAS CHAPTER 7 MANY I&C QUESTIONS 3/17/78 SECOND SUBMITTAL OF FSAR ANSWERED ISC QUESTIONS 6/78 FSAR DOCKETED 9/18/78 RECEIVED ADDITIONAL ACCEPTANCE REVIEW QUESTIONS FROM NRC ADDRESSED OUR RESPONSES IN 3/17 SUBMITTAL 1/5/79 WPPSS RESPONDED TO ACCEPTANCE REVIEW QUESTIONS OKE 3/31/80

3/28/79 RECEIVED ROUND 1 I&C QUESTIONS FROM flRC IDENTIFIED CHAPTER 7 PROBLEMS e REPETITION OF GENERIC CONCERNS ON OTHER DOCKETS e CONFLICTING AND VAGUE STATEMENTS e REQUEST FOR WPPSS TO REVISE CHAPTER 7 3/28/79 TMI 5/11/79 .q[ . WPPSS LETTER 0

e REQUEST flRC TO PROVIDE LIST OF GENERIC ISSUES e COMMITMENT TO REMOVE CONFLICTING AND INCONSISTENT STATEMENTS S REQUEST A MEETING TO RESOLVE l&C CONCERNS FALL 1979 WPPSS DECIDES TO REWRITE CHAPTER 7 OKE 3/31/80

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qax if.2, <pta PURPOSE @#'N CHAPTER 7.0 REWRITE M2##5 Gdtsao WM4.

INCREASE TECHNICAL CONTENT ACCURACY, AND READABILITY.

REMOVE DISCREPANCIES--APPARENT AND REAL.

REMOVEREPETITIbuSDISCUSSIONSWITHINCHAPTER7.0.

TAKE ADVANTAGE OF DISCUSSIONS OUTSIDE CHAPTER 7.0 THROUGH REFERENCE.

DECREASE LENGTH OF SUBSECTION NUMERICAL IDENTIFICATION.

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PRE-REWRITE CHAPTER 7.0 SUBSECTION FORMAT SYSTEM TITLE i

1. POWER SOURCE BRIEF STATEMENT ON INSTRUMENTATION POWER SOURCES.

2. SYSTEM IDENTIFICATION BRIEF DESCRIPTION OF SYSTEM SCOPE AND COMPONENTS.

3. EQUIPMENT DESIGN I

BRIEF DESCRIPTION OF EQUIPMENT AND COMPONENTS .

PROVIDED.

4. INITIATING CIRCUITS DESCRIPTION OF SYSTEM INITIATING LOGIC.

5. LOGIC AND SEQUENCING DESCRIPTION OF INITIATING AND OTHER SYSTEM LOGIC.

6. BYPASSES AND INTERLOCKS j DESCRIPTION OF LOGIC WHICH MAY BE CONSTRUED TO BE BYPASSES OR INTERLOCKS.

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7. REDUNDANCY AND DIVERSITY GENERALLY A DESCRIPTION OF INITIATING LOGIC.

8. ACTUATED DEVICES DESCRIPTION OF LOGIC ACTUATING PUMPS, VALVES, ETC.

9. SEPARATION BRIEF DESCRIPTION OF SEPARATION FOR THE SYSTEM. ,

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10. TESTABILITY DESCRIPTION OF COMPONENT AND SYSTEM TESTING PROVISIONS (MECHANICAL INCLUDED).

11. ENv!RONMENTAL CONSIDERATIONS BRIEF DESCRIPTION OF ENVIRONMENTAL CONSIDERATIONS WITH REFERENCE TO CHAPTER 3.11.

12. REACTOR OPERATOR INFORMATION DESCRIBES SYSTEM FOR PROVIDING OPERATORS WITH INFORMATION (ANNUNCIATORS, COMPUTER).

13. SET POINTS GENERALLY A REFERAL TO A TABLE OF INSTRUMENTATION SET POINTS.

14. DESIGN BASIS INFORMATION CONDITIONS, VARIABLES, LIMITS, MARGINS, FIRES, FLOODS, ETC.-

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15. FINAL SYSTEM DRAWINGS DESCRIPTION OF FINAL DRAWINGS; FCD, P&ID, ETC.

16. ANALYSIS - GENERAL FUNCTIONAL REQUIREMENTS GENERALLY A DESCRIPTION OF SYSTEM FUNCTION, INITIATING CIRCUITS, ENVIRONMENTAL CONDITIONS, AND METHODS OF ASSESSING SYSTEM STATUS.

17. REGULATORY Gu!DE CONFORMANCE

18. FEDERAL CODE CONFORMANCE

19. IEEE STANDARDS CONFORMANCE l

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REWRITTEN CHAPTER 7.0 SUESECTI0tl FORMAT SYSTEM TITLE

1. SYSTEM FUNCTION BRIEF DESCRIPTION OF SYSTEM FUNCTION UNDER VARIOUS PLANT CONDITIONS WITH REFERENCE TO APPROPRIATE WRITEUPS IN OTHER FSAR SECTIONS. -

2. SYSTEM OPERATION DESCRIPTION OF REFERENCES (TABLES AND FIGURES),

INITIATING INSTRUMENTATION AND LOGIC, SYSTEM COMPCNENT RESPONSE TO INITIATING LOGIC, OTHER SYSTEM LOGIC, AND REDUNDANCY / DIVERSITY.

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CHAPTER .7_.0 SUBSECTION CONTENT ADDITIONS AND DRFTIONS SECTION 7.1 - IDENTIFICATION OF SAFETY-RELATED SYSTEMS (

ADDtTIONs: /

ADDEo DISC'USSIONS OF 8EGULATORY GulDES AND

'GDC'S WHICH APPLY EQUALLY TO ALL GAFGTY-RELATED SYSTEMS.

DELETIONS: 1. REMOVED ALL DESCRIPTIONS QF SYSTEMS WHICH ARE NOT S.AFETY-RELATED,

2. REMOVED ALL SYSTEM DESIGN BAS.!S INFOR-MATION - DISCUSSED IN OTHER CHAPTER 7.0 SUBSECTIONS.

3. REMOVED ALL DISCUSSION OF SEPARATION CRITERIA - MOVED 70 SECTION 8.

4. REMOVED DISCUSSION OF REGULATORY (UIDES AND.6DC'S WHICH REQUIRED SYSTEM SPECIFIC DISCUSSIONS IN OTHER CHAPTEA 7.0 SuB-SECTIONS.

5. REM 0vED, FROM TABLE 7.1 ($YSTEM DESIGN AND SUPPLY RESPONSIBILITY), IABLE 7.1-2

($YSTEM SIMILARITY TO LICENSED REACTORS),

AND IABLE 7.1-3 (CODES AND STANDARDS APPLICABILITY), THOSE SYSTEMS WHICH ARE NOT SAFETY-RELATED.

6. REMOVED FRCM IABLE 7.1-3 REFERENCES TO BRANCH TECHNICAL POSITIONS.  !

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7. DELETED TABLES 7.1-4 THROUGH 7.1-9 (SYSTEM COMPONENT CODES AND STANDARDS APPLICABILITY).

8. DELETED IABLES 7.1-10 THROUGH 7.1-13 AND FIGURES 7.1-1 THROUGH 7.1-6 (SEPARA-

! TION CRITERIA) - MOVED TO SECTION 8.3.

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SECTION 7.2 - REACTOR IRIP SYSTEM ~

ADDITIONS: NONE DELETIONS: 1. REMOVED DISCUSSIONS OF GDC'S, IEEE

STANDARDS, AND REGULATORY Gu! DES WHICH APPLY EQUALLY TO ALL SAFETY-RELATED SYSTEMS AND ARE DISCUSSED IN SECTION 7.1.

2. REMOVED DISCUSSIONS PER REVISED FORMAT.

3. DELETED FIGURES 7.2-8 (LOCATION OF SENSORS FOR RPS), 7.2-12 (CONTROL ROD DRIVE PIPING), AND 7.2-13 (TURBINE GENERATOR SUILDING CONCRETE WALL SECTIONS).

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SECTION 7.3 - ENGINEERED SAFETY FEATURE SYSTEMS ADDITIONS: 1. ADDED A DISCUSSION FOR THE RHRS -

SUPPRESSION POOL COOLING MODE

2. ADDED TABLES " CHANNELS REQUIRED FOR PROTECTIVE ACTION COMPLETION" AND

" MINIMUM CHANNELS REQUIRED" FOR SYSTEMS NOT PREVIOUSLY ADDRESSED.

3. ADDED HPCS POWER SUPPLY FCD TO FIGURES.

DELETIONS: 1. REMOVED DISCUSSIONS OF GDC'S, IEEE I STANDARDS, AND REGULATORY GUIDES WHICH APPLY EQUALLY TO ALL SAFETY-RELATED SYSTEMS AND ARE DISCUSSED IN SECTION 7.1.

2. REMOVED DISCUSSIONS PER REVISED FORMAT.

3. DELETED IABLES 7.3-6 (PRM SYSTEM CHARACTERISTICS), 7.3-12 (PRM INSTRUMENTS SPECIFICATIONS), AND 7.3-16 THROUGH I

7.3-46 (VARIOUS BOP SYSTEM ANNUNCIATOR ,

AND COMPUTER LOGGING INFORMATION) -

DISCUSSED ELSEWHERE OR UNNECESSARY.

4. DELETED FIGURES 7.3-1 (ECCS NETWORK MODELS),

7.3-4 (MSIV SOLENOID MECHANICAL ARRANGE-MENT), 7,3-6 (CONTROL ROOM PANELS),

7.3-9 (ECCS SEPARATION SCHEME), 7.3-30 AND 7.3-31 (REACTOR BUILDING FLOOR PLANS) -

ALL SHOWN ELSEWHERE.

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SECTION 7.4 - SYSTEMS REQUIRED FOR SAFE SHUTDOWN ADDIT'ONS:

. 1. ADDED A DESIGN BdSIS DISCUSSION.

2. ADDED TABLES " CHANNELS REQUIRED FOR PROTECTIVE ACTION COMPLETION" AND

" MINIMUM CHANNELS REQUIRED" FOR THE RCIC SYSTEM.

DELETIONS: 1. REMOVED DISCUSSIONS OF GDC'S, IEEE STANDARDS, AND REGULATORY Gu! DES WHICH APPLY EQUALLY TO ALL SAFETY-RELATED SYSTEMS AND ARE DISCUSSED IN SECTION 7.1.

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2. REMOVED DISCUSSIONS PER REVISED FORMAT.

3. DELETED IABLE 7.4 (REACTOR SHUTDOWN COOLING VALVE ISOLATION SIGNA'_S) .

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SECTION 7.5 - SAFETY-RELATED DISPLAY INSTRUMENTATION ADDITIONS: NONE DELETIONS: 1. DELETED ALL ANALYSIS--WILL BE ADDED LATER FOLLOWING ISSUE TO REGULATORY GUIDE 1.97, REVISION 2.

2. DELETED FIGURES 7.5-1 (REACTOR CONTROL BENCHBOARD ARRANGEMENT); AND 7.5-2 (REACTOR CORE' COOLING BENCHBOARD ARRANGE-MENT).

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i SECTION 7.6 - ALL OTHER INSTRUMENTATION SYSTEMS REQUIRED FOR SAFETY ADDITIONS: 1. ADDED DESIGN BASIS DISCUSSION.

2. ADDED IABLES " CHANNELS REQUIRED FOR PROTECTIVE ACTION COMPLETION" AND

" MINIMUM CHANNELS REQUIRED" FOR HIGH/ LOW PRESSURE INTERLOCKS, LDS, AND NMS.

DELETIONS: 1. MODIFIED SECTION TITLE BY DELETING "AND/OR POWER GENERATION" TO ALIGN WITH REGULATORY GUIDE 1.70, REVISION 2.

2. REMOVED DISCUSSIONS OF GDC'S, IEEE STANDARDS, AND REGULATORY GUIDES WHICH APPLY EQUALLY TO ALL SAFETY-RELATED SYSTEMS AND ARE DISCUSSED IN SECTION 7.1.

3. REMOVED DISCUSSIONS FOR REVISED FOR'4AT.

4. REMOVED DISCUSSIONS OF NON-SAFETY-RELATED SYSTEMS (REFUELING INTERLOCKS, N[, j REACTOR MANUAL CONTROL SYSTEM, ROD SEQUENCE CONTROL SYSTEM, ROD WORTH

/, MINIMIZER, REACTOR BUILDING CLOSED i j L COOLING WATER, SRM, RBM, TIP, RECIRCU-l \ N/ LATION PUMP LEAK DETECTION, SRV LEAK DETECTION, AND REACTOR VESSEL HEAD LEAK DETECTION) - DISCUSSIONS EXIST EITHER IN SECTION 7.7 OR ELSEWHERE IN FSAR.

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5. REMOVED DISCUSSIONS OF RPT (RECIRCULATION t

PUMP IRIP). REFERENCED TO APPENDIX H.

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6. REMOVED DISCUSSION ,0F CMS (CONTAINMENT ATMOSPHERE MONITORING SYSTEM). INFOR-MATION MOVED TO SECTION 7.5.

7. MOVED IABLES 7.6-1 (REFUELING INTERLOCK pd EFFECTIVENESS), 7.6-3 (SRM SYSTEM IRIPS),

)f t , 7.6-7 (RBM SYSTEM IRIPS) TO SECTION 7.7.

[ 8. DELETED TABLES 7.6-2 (PRM INSTRUMENT CHARACTERISTICS), 7.6-8 (RPT SYSTEM IIME RESPONSE), 7.6-9 (RPT SYSTEM INSTRUMENT SPECIFICATIONS), 7.6-10 AND 7.6-11 (SUPPRESSION POOL MONITORING SYSTEM ALARMS AND COMPUTER LOGS) - DISUCSSED IN OTHER TABLES, APPENDIX H, OR ARE UNNECESSARY.

9. M0vsD FIGURE 7.6-1 (PRM IED) TO SECTION 7.2.

10. DELETED FIGURES 7.6-4 (RECIRCULATION PUMP LEAK DETECTION BLOCK DIAGRAM), 7.6-5 (RHR/RCIC AREA IEMPERATURE MONITORING BLOCK DIAGRAM), AND 7.6-19 (RECIRCULATION SYSTEM P&ID) , DISCUSSED IN APPENDIX H OR ON LDS IED.

11. MOVED FIGURES 7.6-14 (SRM BLOCK DIAGRAM),

7.6-21 (LPRM INPUT TO RBM), 7.6-23 AND 7.6-24 (RBM RESPONSE TO CONTROL ROD MOTION) TO SECTION 7.7.

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.SECTION 7.7 - CONTROL SYSTEMS NOT REQUIRED FOR SAFETY ADDITIONS: 1. ADDED DISCUSSIONS FOR NEUTRON MONITORING SYSTEM SRM AND RBM, ROD SEQUENCE CONTROL SYSTEM, ROD WORTH MINIMIZER, AND REFUELING INTERLOCKS - FROM SECTION 7.6.

2. ADDED TABLES AND FIGURES FROM SECTION 7.6.

3. ADDED TABLES TO IDENTIFY SYSTEM DESIGN SIMILARITY TO OTHER PLANTS.

DELETIONS: 1. DELETED DISCUSSIONS OF GDC'S, IEEE STANDARDS, AND REGULATORY GUIDES. THESE DO NOT APPLY TO NON-SAFETY-RELATED CONTROL SYSTEMS.

2. REMOVED DISCUSSIONS PER REVISED FORMAT.

3. DELETED DISCUSSION ON RWCU - NOT A MAJOR PLANT CONTROL SYSTEMJ INCLUDED IN CHAPTER-5.0. -

4. DELETED DISCUSSION ON GASEOUS, LIourD, AND SOLID RADWASTE - NOT A MAJOR PLANT CONTROL SYSTEMJ INCLUDED IN CHAPTER 11.0,

5. DELETED TABLES 7.7-1 (CRD SYSTEM INDICA-TORS), 7.7-2 (GASEOUS RADWASTE INSTRUMENTS),

7.7-3 (AREA RADI ATION MONITORING SENSORS) -

ALL SHOWN ELSEWHERE IN FSAR.

6. DELETED FIGURES 7.7-7 (RECIRCULATION SYSTEM FCD), 7.7-11 (DETECTOR DRIVE SYSTEM SCHEMATIC), 7.7-13 (RWCU P&ID), 7.7-14 (RWCU FCD), 7.7-15 (FILTER /DEMINERALIZER P&ID), AND 7.7-16 (AREA PADIATION MONITOR-ING BLOCK DIAGRAM) - ALL ARE SHOWN ELSE-WHERE IN FSAR.

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WilAT CilANGED IN Tile REQUIREMENTS? ..)

IEEE 3'll-1971 i VERSUS SQRT IEEE 323-1971 VERSus NUREG-0588

SEISMIC QUALIFICATION 1

SEQUENCE OF EVENTS e NRC P0sITION STATEMENT FEBRUARY 23, 1979 TO LEAD BWR PLANTS

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e NRC QUESTION #110,'37 & #110,36 MARCH 22,.1979 e INFORMAL MEETINGS WITH NRC APRIL-JULY, 1979 ON SQRT & HYDRODYNAMIC LOADS l

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. A. S.Q.R.T. Requirements

1. For older plants having components qualified under previous criteria (IEEE 344-1971); that components have adequate margin to perform their intended design function during the seismic event.

2. For new plant applications; that there has been uniformity and consistency in implementing the new criteria.

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ACCEPTANCE CRITERIA IEEE CRITERIA COMPARISON FOR QUALIFICATION BY TEST IEEE 344-1971 SQRT Requirements e Single Axis Test e Justify the equipment response 3 is adequately represented by one node. No rodal participation or cross coupling.

e Single Frequency e Justify characteristics of seismic motion dominated by one frequency.

(i.e., structural filtering, pipe mounted).

e Can Certify by Comparing e Justify test input motion has sufficient N

PeagFloorACC to Test intensity and duration to excite all ACC nodes to the required amplitude such that the test response spectra envelopes the corresponding response spectras of the individual modes.

e Single freque.ncy & Single o Consider the effects of hydrodynamic load coupled with seismic loads by

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Axis Test Accepted Without Justification BWRs.

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! ?( )o SEISMIC EXCITATIONS GENERALLY HAVE A BROAD FREQUENCY CONTENT. MULI-FREQUENCY INPUT MOTION SHOULD BE USED IN THE TESTING.

HOWEVER, SINGLE FREQUENCY INPUT, SUCH AS SINE " BEATS," MAY BE APPROVED PROVIDED ONE OF THE FOLLOWING CONDITIONS ARE MET:

1. THE CHARACTERISTICS OF THE SEISMIC IN-PUT MOTION INDICATE THAT THE MOTION IS DOMINATED BY ONE FREQUENCY (e.g. BY

~ ~ STRUCTURAL FILTERING EFFECTS).

2. THE ANTICIPATED RESPONSE OF THE EQUIP-MENT IS ADEQUATELY REPRESENTED BY ONE MORE.

. 3. THE TEST INPUT MOTION HAS SUFFICIENT INTENSITY AND 00 RATION TO EXCITE ALL MODES TO THE REQUIRED AMPLITUDES, SUCH THAT THE TESTING RESPONSE SECTRA WILL ENVELOPE THE CORRESPONDING RESPONSE SPECTRA 0F THE INDIVIDUAL MODES.

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SEISMIC RESPONs2 SPGCTRA COM PAR I son

TO IMPLEMENT REEVALUATION OF EQUIPMENT QUALIFICATION PROGRAM TO SQRT AND HYDRODYNAMIC LOADS e ESTABLISH WPPSS POSITION ON COMPLIANCE e ESTABLISH CLASS 1E LIST (NSSS & B0P) e ESTABLISH SAFETY RELATED MECHANICAL LIST (NSSS 8 BOP) e RETRIEVE QUALIFICATION DOCUMENTS e ESTABLISH REEVALUATION CRITERIA:

8 NEW LOADS DEFINITION o ACCEPTANCE CRITERIA BASED ON SQRT e FINITE ELEMENT SEISMIC ANALYSIS VERSUS LUMPED MASS STICK MODEL e ESTABLISH ORGANIZATION TO PERFORM REEVALUATION e NSSS e BOP -

e PERFORM REEVALUATION AND PREPARE A REPORT ON RESULTS 8 IDENTIFY ITEMS THAT CAN NOT BE SHOWN QUALIFIED OR CON-SERVATIVELY TESTED AND ANALYZED TO SQRT REQUIREMENTS.

e COMMENCE REQUALIFICATION p

ENVIRONMENTAL QUALIFICATION SEQUENCE OF EVENTS e IE CIRCULAR No. 78-08 MAY 31, 1978 e IE BULLETIN No. 79-01 FEBRUARY 8, 1979 e FIRsT ROUND QUESTIONS MARCH 22, 1979 I

i IE BULLETIN No.79-01B JANUARY 14, 1980 OPERATING REACTOR ONLY e NRC Issues NUREG-0588 FEBRUARY 5, 1980  :

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ACCEPTANCE CRITERIA IEEE CRITERIA COMPARISON FOR QUALIFICATION BY TEST IEEE 323-1971 NUREG-0588 Applicable to equipment qualified in accordance with IEEE 323-1971.

e Aging Not Mentioned e Requires aging be addressed for motors j and valve actuators (qualified life) and a life estimate of all other IE equipment.

e Margin Not Mentioned e Requires margin be provided but will evaluate on a case by case basis. Requires minimum of one hour accident test for in-containment equipnent.

e Test Sequence Not Addressed e Requires radiation and LOCA tests be

,. Separate Effects Testing used sequential not separate.

e No preference made on method of e Preferred method is identified as Type Test.

quali fica tion.

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e Identifies levels of Documentation e Allows IEEE-1971 level of document and including Test Procedures & Results does not accept C of C's without procedures  ;

and results.

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T0 IMPLEMENT NUREG-0588 l I o REQUIRES REEVALUATION TO ESTABLISH DEGREE OF COMPLIANCE WITH NUREG-0588 FOR EACH ITEM OF SAFETY RELATED ELECTRICAL EQUIPMENT. INCLUDES B0P AND NSSS.

e JUSTIFY DEVIATIONS e DOCUMENT REEVALUATION EFFORT IN SEPARATE REPORT OR THE FSAR.

e IDENTIFY ITEMS NOT QUALIFIED e REQUALIFY TO CURRENT STANDARDS.

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( ,- Page 1 of 25 Q. 031.100 [ V M u 4. c h -

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u In Table 7.1-2 of the FS AR, you indicate that many of your instrumentation and control systems are identical to those of LaSalLe and Zimmer. During the cou se of our review'of these facilities, which are similar to the WNP-2 facility, we encountered a number of errors in the implementation of the basic GE design. Our concern is that these same errors, or similar errors, could occur in implementing the electrical design of the WNP-2 facility. In particular, we find that your analyses in 7.3.2.1.2.3.1 and 7.3.2.2.2.3.1.1 of the FSAR, to determine compliance with the requirements of IEEE Std. 279-1971, are too general in content. We provide guidance for the inforamtion we need in Section 7.2 of the Standard Review Plan, especialLy 'In Appendix 7.2.A. Specific examples of areas where we require additional information are presented in Items 031.081, 031.084, 031.091, and 031.092 of this enclosure. Accordingly, provide more specific analyses of how you have implemented, in detail, the basic GE electrical design in the WNP-2 facility. References to other sections of the FSAR are acceptable in lieu of repeating this information in 7.3.2.1.2.3.1.

Response

The GE separation criteria has been integrated into the WNP-2 separation criteria. A copy of the criteria is attached.

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. n WNP-2 Page 2 of 25 Cable Separation Criteria Objective

• The instattationof electrical cables shall be in accordance with the foltowing design. criteria. The purposes of these criteria are as folLows:

a. To preserve the independence of redundant safety related electrical systems.

b. To minimize the influence of a non-safety '

related cable on safety related cables.

i

c. To minimize the influence of various types of cables (instrumentation, power, etec) on each other.

d. To give design and installation guidance to assure that separation and identification >

requirements are met.

Definitions and General Requirements - Balance of Plant and Nuclear Steam Suppty Safety Related Systems Definitions Power Cable '

Power cables are defined as those cables that provide electrical energy for equipment motive power and heating requiring 14.4 kv, 6.9 kv, 4.16 kv, 480 volts, 240 volt, 120/208 volt, a-c, 250 and 125 volts d-c. (See Page 031.100-24 for further information.) .

Power cabtes of difference voltage ratings must be routed

• in different cable trays except as folLows: (a) Common tray is permitted for 480 volt, 120/208 volt ac,125 volt  ;

and 250 vold de of compatible divisions; (b) Common tr'ya is permitted for 4160 and 6900 volt power cables of compatible divisions. 480, 4160 and 6900 volt power cables are not to be installed in cable trays in the spreading area beneath the control room. If a run through this area is unavoidable, the power table shalL be installed in conduit.  ;

Power cables shalL be installed in raceways separate from control cables and low-level signal cables and where verti-

• calLy stacked, the power cables shalL be placed in the tray with the highest position in the tray tier. Stacking,of multiple power trays shalL be such that the voltage levels decrease sequentially from the top to the bottom tray in the stack.

B 9

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WNP-2 Page 3 of 25 Control Cable Control cables are those cables using voltages 120 volts ac (or below) or 125 volts de (or below), with normal current not in excess of 30 amperes, whose circuits are designed to supply control power for the plant systems. Included in the category of control cables are those cables used for inter-mittent operation to change the operating status of a t utilization device of the plant system. Control cables include atL cables which have any of the following functions:

(See page 031.100-24 for further information.)

a. 125 volts de or 120 volts ac feeds to switch-gear, panel and local panel control buses.

Wire types are to be power cabler type G2.

b. 125 volt de or 120 volts ac f eeds to solenoids.

c. 125 volts de or 120 volts ac control and interlock circuits.

d. Annunicator circuits.

Instrument Cable (Low-level signals)

Instrumentation cables are those cables used to carry low-Level analog or digital signals. Low-level signal cables require a specific degree of separation or segregation to perserve the accuracy of the transmitted signal. Low-level signal cables are run in raceways separate from alL power and control cables, except within the Control Room Power Generation and Control Complex (PGCC) and as noted below. -

Instrument (signal) trays shalL be of the enclosed (solid bottom and coverse) type.

Analog and digital signal input cables shalL be routed as folLows:

Digital computer signals in the reactor building shalL be run in Divisional control trays as applicable by the device being serviced. Non-Class 1E digital signals in other areas shalL be run in instrumentation trays of Division Br unless they are routed through the reactor building.

• Analog computer signals in the reactor building shaLL.be run in Divisional instrumentation trays as applicable by the device being served. Non-Class 1E analog signals in ohter areas shalL be run in instrumentation trays of Division As unless they are routed through the reactor building.

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WNP-2 Page 4 of 25 Safety Related Electrical and Instrumentation Systems and Equipment Those electrical and instrumentation systems and equipment

'which are relied upon to prevent or mitigate the consequences of accidents and malfunctions originating within the reactor coolant pressure boundary. Safety related systems and equipment are limited in this document to the Reactor Pro-tection System and the Engineered Safeguards Systems.

Reator Protection System (RPS)

The Reactor Protection System is the overalL complex of instrument channelse trip system and trip actuators, and wiring which generates a reactor trip (scram) signal to

-initiate a reactor trip when a monitored parameter (or group of parameters) exceeds a setpoint value indicating the approach of an unsafe condition. The complete RPS is a Class 1E safety related system.

The Reactor Protection System Power Systems consisting of MG sets, distribution panels, etc., is a separater non-safety related system which supplies power to the RPS itself.

Engineered Safeguards Systems (ESS)

This includes that combination of subsystems which take automatic action to provide the cooling necessary to limit or prevent the effects of fuel cladding melting, maintain the integrity of the containment, and insure that the exposure of the public to radiation wilL be below the limits of .

10C FR100 in the event of a design basis reactor accident.

Nuclear Steam Suppty Shutoff System The instrument channels (except those common to RPS), power supplies, trip systems, manual controls, and interconnecting wiring involved in generating a NSSS system function.

Instrument channels for the isolation functions which are shared with the Reactor Protection System are considered a part of the RPS as far as separation is concerned.

Instrument Channel An arrangement of sensory and intermediate components *as required to generate a single trip signal related to a particular plant parameter and introduce this trip signal into a trip system. The channel loses its identifty upon combination of its trip signal with others.

. l

WND-2 Page 5 of 25 Trip System n interconnected arrangement of components making use of instrument channel outputs in the generation of a trip function when appecpriate Logic is satisfied.

Trip Actuator The mechanism which carries out the final action of the protection system.

Redundant System A system or sub-system whose function can be provided by another system or sub-system.

Standby Pouer Sources Emergency "on-site" power sources designed for use when offsite power is not available. These include engine-driven generators and station batteries.

Single Failure A single failure is an occurren~ce which results in the loss-of capability of a component to perform its intended safety functions. Multiple failures resulting from a single occur-rence are considered to be*part of the single failure.

Systems providing safety functions are considered to be designed against an assumed single failure, if a single failure of any component does not result in a loss of capa-bility of the systems to perform their safety function.

a. Active Failure: An active component failure is defined as the malfunction or loss of function of a component of electri' cal or fluid systems.

The failure of an active component of a fluid system is considered to be a loss of component function as a result of mechanical, hydraulic, pneumatic, or electrical malfunction, but not the loss of component structural integrity.

b. Passive Failure: A pas'sive component failure in the sense . utilized i n S e c t i o n 3. 6.1. 21 a

• referes to the failure of:

1) passive electrical equipment such as shorts in cables, .

2) pump or valve seats for long term cooling requirements. ,

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WNP-2 Page 6 of 25 Isolation Device A device in a circuit which limits the effects of events in one section of a circuit from causing unacceptable conse-quences in other sections of the circuits or other circuits.

Some examples of isolation devices.are relays, buffer amplifierse isolation transformers, fuses, circuit breakers and fire stops. '

Raceway Any channel that is designed and used expressty for supporting wire, cables or bus bars. Raceways consist primarily of, but are not restricted to, cable trays, wireways and conduits.

Potential Hazardous Area This is any area in the vicinity of potential missile and '

external fire risk, pipe whip, and jet impingement.

General Area This is an area from which potential hazards of missiles, external fires and pipe whip are excluded.

General Requirements Segreation of Cables outside of the Main Control Room separate cable trays shatt be installed for the five types of cables, i.e., high voltage ,

powere controle low-level signal and RPS, with not more than one of these types of cable permitted in any tray. ,

Separation Details for Raceways The degree of isolation and/or separatation between raceways  :

varies with the potential hazards within a particular area of the station. These areas are classified as folLows:

a. General Areas ,

b. Mechanical Damage (Missile) Area

c. Fire Hazard Area .

d. Cable Spreading Room

e. Control Room e >

9

em -

WNP-2 Page 7 of 25 Minimum separation distances are for open ventitated trays '

providing the fotLowing is observed:

a. Cable splices in raceways are to be prohibited.

b. Cables and raceways are to be flame retardant.

c. Design basis is that cable trays will not b6 filled such that cables extend above tray side raits (this approximates a 50% tray fitL on a '

random basis),

d. Hazards to be Limited tc failures or faults internal to electric cables.

General Areas

a. The minimum separation distance betyeen open cable trays cf redundant divisions or between an open tray of cne division and a conduit of a redundant diyision routed aBove the tray snalL be three feet free air space (hoeirontalLy) and five feet free air space (vertically). Howevers if no automatic area fire detection and extinguishing system exists, and the Lower tray is the highest tray in a tier of more than threef the minimue vertical free air space for separation shalL be eight feet.

The minimum separation distance hetween an open cable tray of one division and a conduit of a redyndant division where the conouit is routed below the open tray shall be one inen. Where equipment arrangement precludes maintaining the minimum separation distancer covers or barriers

. are to be provided between trays of redundant

• divisionse as shown on pages 031.100-15 thru 031.100-18. Circuits of redundant divisions can atso be run in.solto enclosed racewayse such as totally enclosed trays or rigid steel conduits where the minimum established distance for open trays is not maintained.

b. In cases of crossover of one open tray over another of a redundant division where the 4tinimun vertical separati6n criteria established in a.

l above is not maintained, barriers consisting of solid steel covers on bottom trays and solto bottom in top trays shalt be provided. These

. covers shalL extend to tach side of both tray edges by a mininum distance equal to three times the width of the widest tray involved in either l

r, WilP-2 Page 8 of 25 division. The length of the protective covers is taken along the tray centerline. At cross-overs, a minimum vertical separation of one inch is to be provided between the top of the bottom tray and the bottom of the top tray.

c. In cases of crossovers of enclosed raceways and open trays of a redundant division, the minimum separation distance shalL be one inch when the enclosed raceway is below the open tray. Other-wise, vertical separation established in a. above shalL be maintained.

d. Fire stops shall be used where any raceway.pene-trates the slab into the control rooms where any raceway penetrates designated fire arease or where any raceway penetrates areas where an ambient pressure difference exists. In additions fire stops shalL be provided where any open vertical raceway penetrates floor or ceiling slabs. Both the penetration and the trays themselves shalL be seated with fire resistant material.

Mechanical Damage (Missile) Area

a. An analysis shalL be performed to assure that r ou t in g s arrangement and/or protective barriers are such that no credible locally generated missile pipe whip or jet impingement can damage a sufficient amount of safety related cabling or equipment to cause loss of safe shutdown / accident mitigation capability when taken with a single active or passive failure. ,

, b. Class I Electrical Systems Cables shalL not be

• routed through other than Class I structures to protect against earthquake damage and exposure to tornado or flooding conditions unless analysis is i performed to demonstrate that loss of such cables  !

does not negate a safety function.

c. Installation of non-Seismic Category I equipment in areas containing Seismic Category I equipment should be avoided where practicables or adequate barriers shalL be provided to protect Category I equipments or analysis shalL be performed to demonstrate that failure of the non-Category I equipment wilL not Lead to degradation of a plant

. safety function.

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l WNP-2 Page 9 of 25 Fire Hazard Areas

)

l

a. Routing of cables and conduits for safety related redundant systems through an area where there is potential for accumulation of large quantities of oil or other combustible material shalL be pro-hibited.

b., Fire stops shalL be provided for alL tray and conduit (at the first available junction) pene-trations passing through fire rated barriers at fire hazard area boundaries (both sides).

Cable Spreading Room

a. The cable spreading room is the area under the control room where cables l'eaving the panels are dispersed into their various raceways for routing to all parts of the plant.

b. The minimum separation distance between open trays of redundant divisions is to be one foot between trays separated horizontally and three feet between trays separated vertically assuming a fire detection and extinguishing system is present. If these distances cannot be maintained, fire barriers shalL be installed.

c. The minimum separation clearance between conduits and open trays of redundant divisions is 6 inches free air space when the conduit is below or to the side of the open tray and 3 feet free air space when the conduit is located above the open trays.

Control Room

a. In general, no sinle control panel should include wiring essential to more than one safety related redundant function. If cabling of redundant functions must be terminated in the same panel or if cables of redundant divisions run through the same panet, a minimum separation of 6 inches shalt be maintained between cables and components to prevent common damager unless separated by a barrier or an isolation device. A sheet metal enclosure and/or conduit around the intruding division wiring or component is an adequate barrier.

The enclosure (s) shalL include the cabless terminal blocks and the actual device (e.g., switch, light) if required.

, j ,

9 WNP-2 i Page 10 of 25

b. For PGCC see G.E. NEDO 10466.

c. In the ars. behind the PGCC termination cabinets i and near the Control Room walts, cables wilL be routed in grounded flexible conduit and the area provided with a silicone foam fitL or halon fire suppression system, or an alternate method of '

providing electrical separation / fire protection shalL be furnished. .

Identification of Panelse Racks, Junction and PutL Boxes, Cabler Csole Trays and Conduit

a. General Equipment associated with the RPS, NSSS and ESS shalL be identified so that two facts are physi-calLy apparent to the operating and maintenance personnett first, that the equipment is part of nuclear safeguards system; and seconds the grouping (or division) of enforced segregation with which the equipment is associated.

b. Panels and Racks Panels and racks associated with the nuclear safeguards system shall be labelled with marker plates which are conspicuously different in color or color of engraving-fitL from those for other similar panets. The marker plates shalL include identification of the division of the equipment included.

c. Junction or PutL Boxes Junction and/or putL boxes enclosing wiring for the Nuclear Safeguards System shalL have identi-fication similar to and compatible with the panels and racks considered in b. above.

d. CLble Safety related cables (Divions 1 thru 7) shalL be uniquely identified by number and color code.

. Each cable Listed in the cable s:hedule shalL be assigned a number for identification purposes.

The number shalL appear on the electrical instal-Lation drawing and on the wiring diagrams on which the terminations of the cable are shown.

- l' WNP-2 P a g e 11 ao f 25 Cable identification tags shalL be made of a permanent material and permanentty attached to alL cables. Tags shalL indicate the individual cable numbers and the particular separation division to which the cable is assigned according to the marking characteristics showr. in f. below.

Cables shalL be tagged at fifteen foot intervals and at their terminations. This identification' requirement does not apply to individual con-doctors or to cables which run in conduit only.

e. Identification of Cable Tray and Conduit Each cable tray section shalL be assigned an identification node number which is made of a plastic material and applied to the sides of the tray. Moreover, those sections that are assigned a separation code corresponding to the codes assigned to each safety system cable grouping shalL have their respective color numbers marked on their sides in color.

Conduits shalL be tagged in a manner similar to that used for cable identification.

AlL trays and conduits shalL be identified at entrance and exit points of each room they pass through. Conduits shalL be identified at the beginning and at the end, at alL boxes, and at atL discontinuities.

Tray / conduit marking characteristic code 's i shown

• in f. below.

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WNP-2 Page 12 of 25

f. Marking Characteristic Code Tray / Conduit Inscription Divisions Application Characters Characters Background

{

'1 PrCrI Div. 1 Black Yellow 2 P,C,I Div. 2 Black Orange 3 P,CiI Div. 3 Black Red 4 RPS-A1 R Ch. A1 Red Lt. Blue NSSSS-A1 NMS-A 5 RPS-A2 R Ch. A2 Red Green l NSSSS-A2 NMS-C 6 RPS-B1 R Ch. B2 Red Dk. 9Lue NSSSS-B1 NMS-B 7 RPS-B2 R Ch. 92 Red Brown NSSSS-B2 NMS-D A PiC, I Div. A Black Silver or Silver / Yellow Stripe B PrCrI Div. B Black Gold or Go(d/ orange Stripe P - Power C - Control I - Instrumentation Non-Class 1E circuits receiving power from Class 1E power sources which are not shed by an accident signal shalL be identified by the addition of checkered black / silver or black / gold markers indicating the Class 1E division

, (Division 1 or 2 respectively) from which the circuit i

receives its power and identfied as A'1 or B'2 (respec-tively) in the computerized cable schedule.

l 9

WNP-2 Page 13 of 25 l

l Specific Requirements for Separation of Cables for Nuclear Safeguards Systems Reactor Protection System (RPS, NSSSS and NMS)

Reactor Protection System (RPS, NSSSS and NSSSS, and NMS fail-safe wiring:

a. Fails afe wiring outside of the main protection-system cabinets shalL be run in rigid or flexible conduits a'nd/or totally eneLosed trays used for no other wiring and shalL be conspicuously identified -

at atL junction or pull bc<ts. IRM, LPRM inputs and RPS Scram Group output rabtes may be combined in the same wireway provided tKst the foer divi-sional separation is maints+3ed.

b. Wires from both RPS trip system trip actuators to a single group of scram sotenoids may be run in a single conduit; howevers a single conduit shalL not contain wires to more thar. one group of scram solenoids. Wiring for two solenoids on the same control rod may be run in the same conduit.

c. Cables through the primary containment penetra-tions shalL be so grouped that failure of alL cabling in a single penetration cannot prevent a scram. (This applies specificalLy to the neutron monitoring cables and the main steam isolation valves position switches.)

d. Power supplies to systems which de-energize to operate (so called " fail-safe" power supolies) -

require onty that separation which is deemed prudent to give reliability (continuity of oper-ation). Therefores the pro.tection system flywheel motor generator (MG) sets and load circuit breakers are not required to comply with the separstion requirements of this Specification for safety reasons even though the load circuits go to separate panels.

,g

e. Wiri'g for the four RPS scram group outputs and the NSM LPRM. inputs must be routed as four separate divisions.

• . T ) . /

WNP-2 Page 14 of 25 Non-Class 1E Circuit s Non-Class 1E circuits which receive power from Class 1E power sources shalL be uniquely identified and compty with the same separation requirements placed on Class 1E circuits. For example, a Division A non-Class 1E circuit whose power source origin is a Division I critical bus must be separated from a Division B non-Class 1E circuit whose power source origin is a Division II critical bus. '

AlL other non-Class 1E circuits require no separation.

See Table IV for a description of acceptable non-Class 1E circuit routing.

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Page 19.of 25 TABLE"I '

CA3LE RCt;T!UC- CRITIRIA -

(Ixcluding of Redundanc C.*.annels) .

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Differenc Grcup n"-ker indicates sspar;ca raceway.

2. T.I.P. Cables =ay be ec=bined with centrol cables..

3. See Par. 3.6.1.2 for control cable definition.

4. Ncn Ic Digital C%:s. inside the reaccc: b1dg. =ay be mixed with Centrcl Ckts. '

6 031.100-19

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WNP-2 Page 20,of 25 TABLE II ASSIGNMENT OF SYSTEMS TO DIVISION OF' SEPARATION

.I' Division 1 Division 2 Division 3 RHR A RHR B HPCS LPCS RHR C HPCS Diesel-Generator Outboard Isolation Inboard Isolation 125 VDC Valves Valves Battery 3 Standby Emergency Standby Emergency Standby Service i

Power 1 Power 2 Water C RCIC Safety Related Display Instr. 3 Automatic Depressuriza- Automatic Depressuriza-tion Div. 1 controls tion Div. 2 controls Standby Gas Treatment Standby Gas Treatment (Loop 1) (Loop 2) 250 volt de Battery 125 vol.t d e B a t t e r y .1 125 volt de Battery 2

~

24 volt. de Battery 1 24 volt de Battery 2 Standby Service Water Standby Service Water Pump A Pump B MSIV-LCS (Inboard) MSIV-LCS (Outboard)

Leak Det. System 1 Leak Det. System 2 CAC 1 CAC 2 Cont. Inst. Air Cont. Inst. Air 2 ,

SLCS 1 SLCS 2 Mn. Cont. Rm. HVAC 1 Mn. Cont. Rm. HVAC 2

, Reactor Shutdown 1 Reactor Shutdown 2 RPT 1 Output RPT 2 Output Safety Related Safety Related Display Instr. 1 Display Instr. 2 Suppression Pool Suppression Pool Temp. Monit. 1 Temp. Monit. 2

^

3 ' N, WNP-2 Page 21 of 25 4

.i ASSIGNMENT OF RPS, NSSSS AND NMS TO DIVISIONS OF SEPARATION f (FAIL-SAFE WIRING)

. i,

, Division 4 Division 5 Division 6 Division 7 (PGCC Div. 1) (PGCC Div. 2) (PGCC Div. 1) (P.GCC Div. 2)

RPS A1 RPS As RPS B RPS B2 NSSSS A1 NSSSS A2 NSSSS B1 NSSSS B2 NMS A NMS C NMS S NMS D 9

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- u, MEETitiG SUMMARf DISTRIDUTIO:s

& NRC ketPDRFile APR 21 ;ggo Local PDR TIC - NSIC NRR Reading LWR #4 File H. Denton E. Case H. Berkow W. Russell D. Ross D. Vassallo S. Varga J. Stolz R. Baer

0. Parr L. Rubenstein C. Heltemes L. Crocker F. Williams R. Mattson R. DeYoung Project Manager M. D. Lynch Attorney, ELD Licensing Assistant M. Service IE (3)

ACRS (16)

R. Denise NRC

Participants:

M. D. Lynch W. Lovelace A. D. Toth J. Rothfleisch

• T. Abell G. Matthews S. Boyd L. Schaub

• • • e

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. s.

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j APR 2 4,1980 l -

Docket tio.: 50-397 .

APPLICANT: WPPSS (' Washington Public Power Supply-System)

! , FACILITY: WNP-2

SUBJECT:

CASELOAD FORECAST PAtlEL SITE VISIT TO WitP-2 Off FEBRUARY 26, 1980 The Caseload Forecast Panel visited the site of the WNP-2 facility on Februah/ 26, i

1930. The purpose of this visit was to discuss a number of factors affecting the construction schedule of this facility and to conduct a site tour including j

observation of construction activities. A listing of those participating in

' the morning discussion is attached. .The CFP's most recent site visit was June 14,

1978.

l The morning discussion was primarily concerned with an overview of the construction progress since the Udne 1978 visit, including an update on construction progress,

~

a discussion of major milestones completed, current problems and any anticipated problems. The' major areas that were considered in the discussion were the con-l tinuing labor productivity problems and the effect of rework on the sacrificial I

shield on the schedule.' WPPSS personnel indicated that they did not anticipate any delays associated with material ' supply inasmuch as they had either shipped the required material and components to the site in advance of need or had ready access to required materials and comporants. UPPSS personnel also indicated l they did not anticipate any delays associated with the ongoing program to define l the pool dynamic loads of the Mark II containment. WPPSS management concluded, however, that its estimated fuel load date of March 1981 would slip to July 1932.

i During the' afternoon tour of the site, the CFP went through all the significant j portions of the facility. It was observed that' additional work remains to be

completed in 'the wet well and a considerable amount of work remains to be done in the control-room. About 80 percent of the electrical and instrumentation cables have been pulled; almost all of the cable trays appear to have been installed.

j While most of the piping systems have been installed, a considerable amount of piping is still being installed. One of the management techniques recently adopted for coordinating construction activities of different crafts appears to be working i

very effe:tively. In this approach, a single individual is responsible for i

scheduling and coordinating all crafts in each area. These area managers, in I

turn,.are part of an integrated team composed of WPPSS and AaE personnel.

At the exit interview on February 28, 1980, the CFP stated its belief that WPPSS' t

estimate of July 1982 for a fuel load date appears reasonable at this time. -

However,theS't}gscheduledueto:RC the co g staff expressed its concerns about the vs

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• (1) the inspection, rework and documentation of the flaus in the sacrificial shield; (2)'the tight confinement in the wet well, especially welding of the omega seal; (3) the time to conduct geological, seismological and structural reevaluations in light of recent geologic phenomena south and scuthwest of the site; and -

!- (4) the impact of the. Action Plans (NUREG-0660) on 1itiP-2. -

I -

I The staff finished its exit intarview by complimenting the !IPPSS management and i

personnel for a well coordinated, straightforward presentation in the mornino F

i discussion and for a thorough well-guided tour in the afternoon, i

' Original signed by:

M. D. Lynch, Project ;'anager Light Llater Reactors Branch No. 4 Division of Project W n gement

Attachment:

.LI stated

cc w/ attachment

See next page _

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'go UNITED STATES J h NUCLEAR REGULATORY COMMISSION J' a WASHING TON, D. C. 20555

\ *****

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APR 2 41980 Docket No.: 50-397 APPLICANT: WPPSS (Washington Public Power Supply System)

FACILITY: WNP-2

SUBJECT:

CASELOAD FORECAST PANEL SITE VISIT TO WNP-2 ON FEBRUARY 26, 1980 The 1980.Caseload Forecast Panel visited the site of the WNP-2 facility on February 26, The purpose of this visit was to discuss a number of factors affecting the construction schedule of this facility and to conduct a site tour including observation of construction activities. A listing of those participating in the morning discussion is attached.

1978. The CFP's most recent site visit was June 14, The morning discussion was primarily concerned with an overview of the construction progress since the June 1978 visit, including an update on construction progress, a discussion of major milestones completed, current problems and any anticipated problems.

The major areas that were considered in the discussion were the con-tinuing labor productivity problems and the effect of rework on the sacrificial shield on the schedule. WPPSS personnel indicated that they did not anticipate any delays associated with material supply inasmuch as they had either shipped the required material and components to the site in advance of need or had ready access to required materials and components. WPPSS personnel also indicated they did not anticipate any delays associated with the ongoing program to define the pool dynamic loads of the Mark II containment. WPPSS management concluded, however, that its estimated fuel load date of March 1981 would slip to July 1982.

During theofafternoon portions tour of the site, the CFP went through all the significant the facility.

It was observed that additional work remains to be completed in the wet.well and a considerable amount of work remains to be done in the control room. About 80 percent of the electrical and instrumentation cables have been pulled; almost all of the cable trays appear to have been installed.

While most of the piping systems have been installed, a considerable amount of piping is still being installed. One of the management techniques recently adopted for coordinating very effectively. construction activities of different crafts appears to be working In this approach, a single individual is responsible for scheduling and coordinating all crafts in each area. These area managers, in turn, are part of an integrated team composed of WPPSS and A&E personnel.

At the exit interview on February 28, 1980, the CFP stated its belief that WPPSS' estimate of July 1982 for a fuel load date appears reasonable at this time.

However, the NRC staff expressed its concerns about the potential for delay in-the construction schedule due to:

APR t 4 E20 (1) the inspection, rework and documentation of the flaws in the sacrificial shield; (2) the tight confinement in the wet well, especially welding of the omega seal; (3) the time to conduct geological, seismological and structural reevaluations in light of recent geologic phenomena south and southwest of the site; and (4) the impact of the Action Plans (NUREG-0660) on WNP-2.

The staff finished its exit interview by complimenting the WPPSS management and personnel for a well. coordinated, straightforward presentation in the morning discussion and for a thorough well-guided tour in the afternoon.

f.

V <~~ .

M. D. Lynch, oject Manager Light Water Reactors Branch No. 4 Division of Project Management

Attachment:

As stated cc w/ attachment:

See next page t

t,PR 2 4 l'd0 Washington Public Power Supply System

• CCs:

Joseph B. Knotts, Jr., Esq. Mr. Neil 0. Strand Debevoise & Liberman Washington Public Power Supply System 1200 Seventeenth Street, N. W. P. G. Box 968

. Washington, D. C. 20036 Richland, Washington 99352 Richard Q. Quigley, Esq.

Washington Public Power Supply System P. O. Box 958 Richland, Washington 99352 Nicholas Lewis, Chairman Energy Facility Site Evaluation Council 820 East fifth Avenue Olympia, Washington 98504 Mr. O. K. Earle Licensing Engineer P. O. Box 968 Richland, Washington 99352 Resident Inspector /WPPSS-2 NPS c/o U.S. Nuclear Regulatory Cannission P. O. Box 69 Richland, Washington 99352 4

e e

. o ATTACHMENT List of Attendees February 26, 1980 WNP-2 Facility Name Organization K. Earle WPPSS M. D. Lynch NRC W. Lovelace NRC A. D. Toth NRC Resident C. Taylor WPPSS Pat Powell WPPSS G. C. Sorensen WPPSS J. D. Martin WPPSS H. L. Perkins WPPSS D. L. Renberger WPPSS J. P. Thomas WPPSS W. A. Goodman WPPSS W. C. Bibb WPPSS J. R. Lewis BPA C. R. Bryant BPA G. I. Wells WPPSS A. D. Kohler WPPSS L

50-197

g. S

• M.AY 9 1960 ME!!0RANDUM FOR: Paul Fine, Technical Assjstant , Technical Support Branch, NRR ,,

FROM: , B. J. Youngblood, Chief. Licensing Branch. No.1,. DL

SUBJECT:

FOIA REQUEST 80-233,.- WASHINGTO NUCLEAR PROJECT NO. 2 (WPPSS-2),N PUBLIC POWER SUPPLY. S Information pertaining to the Freedom of Information Request FOIA-80-233 . '

requested by George Du Vall is attached.. . A. reply to items 2, 3, 4, 5, 6 and 3 are attached. , ..

NUREG-0487 " Mark II Containment Lead Plant Program Load Evaluation and Acceptance Criteria" has been provided. to. correspond with the reply for Item 1.

Item i should be answered by Farouk Eltawila of.the Containment Systems Branch of the Division of Systems Integration. ._ ,

Items 7 and 9 should be answered by Inspection.and Enforcement.

ORIGINAL SIGE Y B. J. Younablood. Chief Licensing Branch No.1 Division of Licensing

Enclosures:

As stated .

DISTRIBUTION: w/en osure M cket File (50-397 PDR LPDR l

LB#1 Reading .

l DLynch MRusk.orook BJY9ungblood q

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Nec F:mm sie (9 76) NRCM 0240 #U.S. GOVERNMENT PRINTING OFFICE: 1979-289 369

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1764.2-The OL c;?licati;n for W::P-2 .cas da:Leted in 1.ne :373 - ;:: i:.2n c; f. c

., thrcush the first econd of ques tions.

I i.E'4. 3-C.,nservative bounding load criteria have bcan required by tha ',;C t taff as a result of the ongoing load definition progran for the pool d . . c ir S is associated with the ark 11 vapor suppression contair ent tys%n. TP. s e conservative lead criteria nacessitated r..,cificat.icns to th? s tract ras sub,iected to pool dynai.ic loads.

.I.T. _EM_.4 We tnew of no rcasen to suspend or raveke the cor.s trac tica cr c.it 'sr ..:,?-2.

I.T EM, _5_

There are a total of tan units which have Sek II v. por so;pr:ssion ccr.ti. 3nt systems. Their ccnstruction status as of Cace-$ar 31, 19 79, '..a s :

Unit. Paf. cent Conplate,

1. 7.i ..a r 9 5';

2. LaSalle 1 935 '

3. Skrcham 60%

4 W:;P-2 80%

5. I.aSalle 2 74%

6. Susqu'-hanna 1 74%

7. Li.:erick 1 56%

8. Susquehar.na 2 46%

9. Li:.arick 2 32%

10. Eailly 1%

This infch.ation cas taken feca .'," REG-0330, '.'clure III, ' o.1, dated February P::.0,

' Construction Sta tus port .*bclear Pcc.ar Plants. "

ITEM 6 Ziaer is a BNP-5/ Mark II plant similar but not idantical to '.:,3-2. "il e b.ith ..

plants have General Elactric as the vendor of the nuclear s te.in s r.nly jst3n, there are two different architect-engineers res:,ansible for tha iotal :: sign of the plant.

r ..,

.* ! r.p 8, ,

Thcre :as no specific infor. aticn received by the .';RC in :975 .hich lad t;a '.?C to rc-avaluate the Mark 11 pool dynanic leads. Mc'.2 var, GE p resentad da ta to the

'.RC in late 1974 relating to the paol s< call leads that ' mld occur in tha . . int Of a fasign basis less-of-coolant accient. 'lhile those ic3 fs ..are r. 5. rid in a

.as t f cility re;:. csantative of tha " irk III c: ti: int, it is Lali .id t'it ti.e sama ;!.cnar inan ...v..ld occur in beth tha " irk I and " irk !! ar < .. si;n

. On'ain .mts.

. The ::;C staff also 5 ci .a ire in 1974 of sf.nifi. .nt ul

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MEETING fl0TICE DISTRIBUTION N b. Y b f,

c Doeket File J. Stolz ' '

NRC POR S. Hanauer Loc &l POR P. Collins TIC

0. Vassallo NSIC .

D. Ziemann Branch File R. Mattson H. Denton R. Schroeder E. Case K. Kniel D. Eisenhut O. Skovholt R. Purnle r,. Knighton T. Novak M. Ernst S. Varga R. Baer T. Ippolito C. Berlinger R. A. Clark S. Israel R. Reid ACRS (16)

R. Tedesco Attorney, OELD J. Youngblood -

OIE (3)

A. Schwencer 050 (7)

Chief, Licensing Branch #3 Profect Manager M. D. Lynch J. R. Miller G. Lainas Licensino Assistan't _M. Rushbrook Receotionis t D. Crutchfield TERA Chief, Systemic Evaluation Program Branch J. LeDoux, I&E Chi.ef, Operating Reactors Assessment Branch R. Vollmer ISE Headnuarters R. Bosnak IAE Region I F. Schauer  !&E Recion It R. E. Jackson I&E Region !!!

G. Lear  !&E Region IV V. Noonan  !&E Region V S. Pawlicki V. Benaraya NRC Particioants:

Z. Rosztoczy

• "d M. D. Lynch

, u r R. Satterfield R. Ballard W. Regan J. D. Saltzman D. Ross P. Check R. Satterfield O. Parr bec: Aoplicant & Service List F. Rosa W. Butler W. Kreger R. W. Houston T. Murphy W. Gammill L. Rubenstein T. Soeis Chief, Core Performance Branch

. . ~

i pN :

MAY 2 81980 1

Docket No.: 50-397 . ,

i MEPORANDUM FOR: B. J. Youngb1_ood, Chief ., .

j Licensing Branch No. 1, DL ,

FROM: M. D. Lynch, Project Manager Licensing Branch No. 1. DL

SUBJECT:

FORTHCOMING MEETING WITH WPPSS TO DISCUSS THE REVISED CHAPTER 7 -

.; 0F THE WNP-2 FSAR Date: - Thursday, May 29, 1980 Time & Location: 8:30 AM - 11:30 AM Suite 203, Landow Building ,

7910 Woodmont Avenue, Bethesda, Maryland 1:00 PM - 3:30 PM Room 6110, Maryland National Bank Building 7735 Old Georgetown Road, Bethesda, Maryland .

Purpose:

To discuss the recently revised Chapter 7 of the WNP-2 FSAR and to discuss any apparent conflicts between the revised Chapter 7 and Chapter 15

Participants:

NRC M. D. Lynch, R. Satterfield, et al .

, SRL A. Hadden, et al.

WPPSS K. Earle, et al .

G_E_ ,

R. Johnson, et al.

ORIGINAL SIGNED GY ,

i M. D. Lynch, Project Manager

Licensing Branch No.1 Division of Licensing

cc

See A xt,page ,, ..

orrice) . .(Bf.b.? . DL ;h d)f. , . . , ,,,,,, ,,, , ,, ,,,,, ,,,,,, ,,,,, ,,,,.,,,,,,,, ,_,, ,, .......... ..

5URNA"bMDLynch/ls BJYbu'noblood -" "" " " -""- - "- "

5/ 28/80 """ S/17/ 80 " "

onE> ............. ... .d..J..... . . .... .... . .. ........ ....... . ... .......... .. ... . . ....

i MAC FORM 318 (9 76) N#CM 0240 -

. W U.S. GOVERfmENT PRINTING OrNCE: 1979-201 369

- _ s __ _ .. . ,

Washington Public Power Supply System CCs: '

Joseph B. Knotts, Jr., Esq.

Debevoise & Liberman 1200 Seventeenth Street, N. W.

Washington, D. C. 20036 Richard Q. Quigley, Esq.

Washington Public Power Supply System P. O. Box 968 .

Richland, Washington 99352 Nicholas Lewis, Chairman '

Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504 .

Mr. O. K. Earle Licensing Engineer P. O. Box 968 <

Richland, Washington 99352 ,

Resident Inspector /WPPSS-2 NPS c/o U.S. Nuclear Regulatory Commission P. O. Box 69 '

Richland, Washington 99352 O

g' kg UNITED STATES

• y NUCLEAR REGULATORY COMMISSION 5 E WASHINGTON, D. C. 20S55 pM

"$g JUN 9 1980 Docket Nos.: 50-341 50-373 '

50-322 50-387 50-358 50-397 MEMORANDUM FOR: Robert L. Tedesco, Assistant Director for Licensing, DL THRU: B. J. Youngblood, Chief, Licensing Branch No'.1, DL FROM: Jerry N. Wilson, Project Manager Licensing Branch No. 1, DL

SUBJECT:

FORTHCOMING MEETING WITH B0ILING WATER REACTORS -

LICENSING RESOLUTION GROUP (BWR-LRG)

DATE & TIME: June 13, 1980, Friday 10:00 AM LOCATION: G. E. Office - Landow Building Bethesda, Maryland PURPOSE: To discuss the progress of the BWR-LRG PARTICIPANTS: NRC T. Speis, R. Satterfield, F. Rosa, J. Wilson, I. Peltier, R. Stark, M. Lynch, T. Bournia, and L. Kintner BWR-LRG L. DelGeorge, R. Boyd, et al.

I v 1.42 %D y . Wilson, Project Manager

'fi e sing Branch No. I sion of Licensing cc: See next page

, . s,

. ~.  ;

}

Mr. D. Louis Peoples Director of Nuclear Licensing -

Commonwealth-Edison Company P. O. Box 767 "

Chicago, Illinois 60690 cc: Richard E. Powell, Esq.

Isham, Lincoln & Beale One First National Plaza 2400 Chicago, Illinois 60670 ,

s.

Dean Hansel 1., Esq.  ;

Assistant Attorney General State of Illinois 188 West Randolph Street Suite 2315 Chicago. Illinois 60601 ' ,

Mr. Roger Walker, Resident Inspector U. S. Nuclear Regulatbry Commission P. O. Box 737 Streator, Illinois 61364 i

+

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Long Island Lighting Company ccs:

Howard L. Blau, Esq.

Blau and Cohn, P.C.

217 Neubridge Road Hicksville, New York 11801 Jeffrey Cohen, Esq.

Deputy Commissioner and Counsel New York State Energy Office Agency Building 2 Empire State Plaza Albany, New York 12223 Energy Research Group, Inc.

400-1 Totten Pond Road Waltham, Massachusetts 02154 Irving Like, Esq.

Reilly, Like and Schnieder 200 West Main Street Babylon, New York 11702 J. P. Novarro Project Manager Shoreham Nuclear Power Station P. O. Box 618 Wading River, New York 11792 W. Taylor Reveley, III, Esq.

Hunton & Williams '

P. O. Box 1535 Richmond, Virginia 23212 Ralph Shapiro, Esq.

Camner & Shapiro 9 East 40th Street New York, New York 10016 Edward J. Walsh, Esq.

General Attorney Long Island Lighting Company 250 Old Country Road Mineola, New York 11501 Resident Inspector /Shoreham NPS c/o U.S. Nuclear Regulatory Commission l P. C. Box 8 l Rocky Point, New York 11778

Long Island' Lighting Company ces (continued)

Honorable Peter Cohalan Suffolk County Executive County Executive / Legislative Building Veteran's Memorial Highway Hauppauge, New York 11788 David Gilmartin, Esq. .

Suffolk County Attorney County Executive / Legislative Building Veteran's Memorial Highway Hauppauge, New York 11788 MHB Technical Associates 1723 Hamilton Avenue - Suite K San Jose, California 95125 Stephen Latham, Esq.

Twomey, Latham & Schmitt P. O. Box 398 33 West Second Street Riverhead, New York 11901 Joel Blau, Esq, t!ew York Public Service Commission The Governor Nelson A, Rockefeller Bldg.

Empire State Plaza Albany, New York 12223

r-Mr. Norman W. Curtis Vice President - Engineering and Construction Pennsylvania Power and Light Company 2 North Ninth Street Allentown, Pennsylvania 18101 cc: Mr. Earle M. Mead Mr. Robert J. Shovlin Project Engineering Marager Project Manager Pennsylvania Power & Light Company Pennsylvania Power and Light Co.

2 North Ninth Street 2 North Ninth Street Allentown, Pennsylvania 18101 Allentown , Pennsylvania 18101 Jay Silberg, Esq.

Shaw, Pittman, Potts & Matias F. Travieso-Diaz, Esq.

Trowbridge Shaw, Pittman, Potts &

Trowbridge 1800 M St reet , N. W. 1800 M Street, N. W.

Washington, D. C. 20036 Washington, D. C. 20036 Mr. William E. Barberich, Dr. Judith H. Johnsrud Nuclear Licensing Group Supervisor Co-Director Pennsylvania Powr & Light Company Environmental Coalition on 2 North Ninth Street Nuclear Power Allentown, Pennsylvania 18101 433 Orlando Avenue State College, PA 16801 Edward M. Nagel, Esquire General Counsel and Secretary Mr. Thomas M. Gerusky, Director Pennsylvania Power & Light Company Bureau of Radiation Protection 2 North Ninth Street Department of Environmenta-1 Allentown, Pennsylvania 18101 Resources Connonwealth of Pennsylvania Bryan Snapp, Esq. P. O. Box 2063 Pennsylvania Power & Light Company Harrisburg, PA 17120 901 Hamilton Street Allentown, Pennsylvania 18101 Ms. Colleen Marsh Box 538A, RD#4 Robert M. Gallo Mountain Top, PA 18707 Resident Inspector P. O. Box 52 Shickshinny, Pennsylvania 18655 Mrs. Irene Lemanowicz, Chairperson The Citizens Against Nuclear Dangers Susquehanna Environmental Advocates P. O. Box 377 c/o Gerald Schultz, Esq. RO#1 500 South River Street Berwick, PA 18503 Wilkes-Barre, PA 18702 John L. Anderson Oak Ridge National Laboratory Union Carbide Corporation Bldg. 3500, P. O. Box X Oak Ridge, Tennessee 37830

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Washington Public Power Supply System ces:

Joseph B. Knotts, Jr., Esq.

Debevoise & Liberman .

1200 Seventeenth Street, N. W.

. Washington, D. C. 20036 Richard Q. Quigley, Esq.

Washington Public Power Supply System P. O. Box 968 Richland, Washington 99352 Nicholas Lewis, Chairman '

Energy Facility Site Evaluation Council 820 East Fifth Avenue Olympia, Washington 98504 Mr. O. K. Earle Licensing Engineer P. O. Box 968 Richland, Washington 99352 Resident Inspector /WPPSS-2 NPS c/o U.S. Nuclear Regulatory Commission P. O. Box 69 Richland, Washington 99352 e

Dr. Wayne H. Jens Assistant Vice President Engineering & Construction Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 cc: Eugene B. Thomas, Jr. , Esq. Mr. Jeffrey A. Alson LeBoeuf, Lamb, Leiby & MacRae 772 Green Street, Building 4 1333 New Hampshire Avenue, N. W. Ypsilanti, Michigan 48197 Washington, D. C. 20036 David E. Howell, Esq.

, Peter A. Marquardt, Esq. 21916 John R Co-Counsel Hazel Park, Michigan 48030 The Detroit Edison Company 2000 Second Avenue Mrs. Martha Drake Detroit, Michigan 48226 230 Fairview Petoskey, Michigan 49770 Mr. William J. Fahrner Project Manager - Fermi 2 William J. Scanlon Esq.

The Detroit Edison Company 2034 Pauline Boulevard 2000 Second Avenue Ann Arbor, Michigan 48103 Detroit, Michigan 48226 i Mr. Larry E. Schuerman Licensing Engineer - Fermi 2 Detroit Edison Company 2000 Second Avenue Detroit, Michigan 48226 Charles Bechhoefer, Esq., Chairman Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Comission Washington, D. C. 20555 .

Dr. David R. Schink Department of Oceanography Texas A & M University College Station, Texas 77840 Mr. Frederick J. Shon Atomic Safety & Licensing Board Panel U. S. Nuclear Regulatory Comission Washington, D. C. 20555

(

Mr. Earl A. Borgmann Vice President - Engineering Cincinnati Gas & Electric Company P. O. Box 960 -

Cincinnati, Ohio 45201 cc: Troy B. Conner, Jr. , Esq. David B. Fankhauser, PhD Conner, Moore & Corber 3569 Nine Mile Ptad 1747 Pennsylvania Avenue, N. W. Cincinnati, Ohio 45230 Washington, D. C. 20006 Dr. Frank F. Hooper Mr. William J. Moran School of Natural Resources General Counsel Ann Arbor, Michigan 48109 Cincinnati Gai & Electric Company P. O. Box 96C fis. Augusta Prince,. Chairperson Cincinnati, Ohio 45201 601 Stanley Avenue Cincinnati, Ohio 45226 Mr. William G. Porter, Jr.

Porter, Stanley, Arthur Charles Bechhoefer, Esq. . Chairman and Platt Atcmic Safety & Licensing Board 37 West Broad Street Panel Columbus, Ohio 43215 U. S. Nuclear Regulatory Comission Washington, D. C. 20555 Mr. Steven G. Smith, Manager Engineering & Project Control fir. Glenn 0. Bright Dayton Power & Light Company Atomic Safety & Licensing Board P. O. Box 1247 Panel Dayton, Ohio 45401 U. S. Nuclear Regulatory Commission Washington, D. C. 20555 J. Robert Newlin, Counsel

' Dayton Power & Light Company Leah S. Kosik. Esq.

P. O. Box 1247 3454 Cornell Place Dayton, Ohio 45401 Cincinnati, Ohio 45220 Mr. James D. Flynn, Manager W. Peter Heile, Esq.

Licensing Environmental Affairs Assistant City Solicitor Cincinnati Gas & Electric Company Room 214, City Hall' P. O. Box 960 Cincinnati, Ohio 45220 Cincinnati, Ohio 45201 Atomic Safety & Licensing Board Mr. J. P. Fenstermaker Panel Senior Vice President-0perations U. S. Nuclear Regulatory Comission Columbus & Southern Ohio Washington, D. C. 20555 Electric Company 215 North Front Street Atomic Safety & Licensing Appeal Columbus, Ohio 43215 Board U. S. Nuclear Regulatory Comission David Martin, Esq. Washington, D. C. 20555 Office of the Attorney General 209 St. Clair Street First Floor Frankfort, Kentucky 40601

Mr. Earl A. Borgmann cc: Resident Inspector /Zimmer

.U. S. Nuclear Regulatory Commission .

P. O. Box 58

) New Richmond, Ohio 45157 Dale D. Brodkey Assistant Attorney General Division of Environmental Law Office of the Attorney General 209 St. Clair Street Frankfort, Kentucky 40601 ..

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