ML20154S245

Summary of 860312-14 Meetings W/Util in Berlin & Waterford, CT Re Actions to Resolve Isap Topics.List of Attendees & Draft Preliminary Assessment of Isap Topics Encl

ML20154S245
Person / Time
Site:	Millstone
Issue date:	03/24/1986
From:	Boyle M Office of Nuclear Reactor Regulation
To:	Office of Nuclear Reactor Regulation
References
NUDOCS 8604010011
Download: ML20154S245 (90)

Text

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j e a 8 c UNITED STATES 8 o NUCLEAR REGULATORY COMMISSION

, WASHINGTON, D. C. 20555 k . . . . . p# March 24,1986 Docket No.: 50-245 LICENSEE: Northeast Nuclear Energy Company FACILITY: Millstone Nuclear Power Station, Unit No. 1

SUMMARY

OF MARCH 12-14, 1986 MEETING TO DISCUSS THE ACTIONS TO RESOLVE ISAP TOPICS By letter dated March 3, 1986, the staff issued a safety evaluation for the detailed scope of the topics to be considered in the Integrated Safety Assessment Program (ISAP) for Millstone 1. This safety evaluation was based upon the individual topics summaries and Probabilistic Safety Study topic sumaries submitted by the licensee. On March 12-14, 1986, the NRC staff met with representatives of Northeast Nuclear Energy Company (those attending the meeting are identified in Enclosure 1) to discuss the actions proposed by the licensee to resolve the Millstone 1 ISAP topics. The licensee provided a draft copy of their preliminary assessment of ISAP topics (Enclosure 2) to facilitate the discussion.

On Wednesday, March 12, 1986, the staff met with the licensee at their offices in Berlin, CT. At that time, the staff and the licensee discussed the status and proposed actions to resolve the ISAP topics, including the three new topics identified in the staff's March 3,1986, SER. In general, the staff agreed that the actions proposed should resolve the topics. However, the staff noted specific instances where clarification of the proposed action will be necessary before the topic can be resolved. The staff also found the licensee's preliminary decision as to which topics warrant further action and which don't requires further explanation on the decision bases. During this session of the meeting, the staff identified specific aspects of the ISAP topics which the staff felt should be discussed with the plant operations and maintenance personnel or

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could be clarified by a direct examination of existing systems or components at the plant.

On Thursday, March 13, 1986, the staff met with the licensee at the Millstone 1 plant site in Waterford, CT. At that time, the staff and Millstone 1 plant personnel discussed the status of several of the ISAP topics to get an operations /

maintenance perspective on the topics. The staff and the licensee discussed l maintenance and operating procedures to clarify specific aspects of several topics. Also,i.he staff toured the plant to examine the specific components or systems related to the proposed topic rarolution.

At the last session of the meeting on March 14, 1986, the staff met with the license'e at their offices in Berlin, CT. During this session the staff end the licensee discussed the status of the projects that are to be completed independent of the Millstone 1 ISAP and the schedule for the remainder of the Millstone 1 ISAP. The licensee stated that 19 of the independent projects have already been completed and that work was continuing on the remaining projects. The licensee also stated that a formal assessment of tire ISAP topics, which reflect the comments discussed in the meeting, may be submitted 8604010011 860324 DR ADOCK 0 25 ,

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March 24,1986 by the end of April 1986. That submittal will reflect the clarifications requested by the staff during this meeting and will also present the basis and criteria for identifying which ISAP topics warrant further action and which do not. After the licensee's submittal of their formal assessment, the staff will schedule another meeting to discuss the submittal to identify any potential differences between the staff and the licensee with respect to specific topic resolutions.

As soon as possible after that meeting, the staff will issue its draft Integrated Safety Assessment Report for the Millstone 1 ISAP for review by the licensee and a peer review panel. The licensee should wait until after the draft report has been published to formalize the implementation of ISAP topic resolutions in a proposed integrated schedule.

Michael L. Boyle, IAPM Integrated Safety Assessment Project Directorate Division of PWR Licensing - B

Enclosures:

As Stated DISTRIBUTION

& Docket: File p ISAP Reading NBoyle CGrimes JShea NThomasson NRC PDR Local PDR l l

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ISAP:DPL-B ISAP: B ISAP:DPL-MBoyle:lt .JShep CGrimes 03/'l.j/86 03/19/86 03 M /86 i

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l by the end of April 1986. That submittal will reflect the clarifications requested by the staff during this meeting and will also present the basis and .

criteria for identifying which ISAP topics warrant further action and which do '

not. After the licensee's submittal of their formal assessment, the staff will schedule another meeting to discuss the submittal to identify any potential differences between the staff and the licensee with respect to specific topic resolutions.

As ::non as possible after that meeting, the staff will issue its draft Integiotc? Safety Assessment Report for the Millstone 1 ISAP for review by the licensee ans a peer review panel. The licensee should wait until after the draft report has been published to formalize the implementation of ISAP topic resolutions in a proposed integrated schedule.

Michael L. Boy e, IAPM Integrated Safety Assessment Project Directorate Division of PWR Licensing - B

Enclosures:

As Stated

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Mr. John F. Opeka Millstone Nuclear Power Station Northeast Nuclear Energy Company Unit No. I cc:

Gerald Garfield, Esquire Kevin McCarthy, Director Day, Berry & Howard Radiation Control Unit Counselors at Law Department of Environmental City Place Protection Hartfor'd, Connecticut 06103-3499 State Office Building Hartford, Connecticut 06106 Edward J. Mroczka Vice President, Nuclear Operations Richard M. Kacich, Supervisor Northeast Utilities Service Company Operating Nuclear Plant Licensing Post Office Box 270 Northeast Utilities Service Company Hartford, Connecticut 06141-0270 Post Office Box 270 Hartford, Connecticut 06141-0270 State of Connecticut Office of Policy and Management ATTN: Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06106 Regional Administrator, Region I U.S. Nuclear Regulatory Commission 631 Park Avenue King of Prussia, Pennsylvania 19406 Northeast Nuclear Energy Company -

ATTN: Superintendent Millstone Nuclear Power Station P. O. Box 128 Waterford, Connecticut 06385 1

Resident Inspector j

c/o U.S. NRC '

Millstone Nuclear Power Station '

P. O. Box 811 Niantic, Connecticut 06357 First Selectman of the Town of Waterford {

Hall of Records 200 Boston Post Road Waterford, Connecticut 06385 l

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2 MARCH 12-14, 1986 Enclosure 1 Millstone Unit No. 1 ISAP Meeting Name Organization Gerard van Noordennen NUSCO - Licensing Paul A. Blasioli NUSCO - Licensing Mitchell S. Ledennan NUSCO - Licensing Richard M. Kacich NUSCO - Licensing Robert Factora NUSCO - Project Management John M. Quinn NNECO - Unit 1 Engineering Eric A. DeBarba NUSCO - Engineering Clinton Gladding NUSCO - Engineering Tom Starr NUSCO - Engineering John Bickel NUSCO - Engineering John Stetz Millstone 1 Station Superintendent Pete Prescott NUSCO -

Geoffrey Grant NRC - Resident Inspector Theodore Rebelowski NRC - Senior Resident Inspector Christopher I. Grimes NRC - ISAPD Michael L. Boyle NRC - ISAPD James Shea NRC - ISAPD Glenn Kelly NRC - ISAPD l

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ISAP Topic No.1.41 - Flooding of Compartments by Backflow Reference

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21 Proposed Action ISAP Topic No.1.41 addresses concerns related to flooding of safety equipment corapartments by backflow through floor drains as noted in NRC Generic Issue 77. Reference 21 provided the NRC with a review of the applicability of GI-77 to Millstone Unit No.1. The referenced letter provided the results of a NNECO review of the equipment and floor drain systems as well as the plant drainage system designs for Millstone Unit No.1. The review identified areas of the plant drainage system which had the potential for backflow through drain

basins. Following the review, procedural changes were instituted to provide guidance to operators as to which plant drains are to be temporarily plugged in the event of flooding. No plant design modificatiou were required.

Based on the plant drainage review undertaken by NNECO, we believe that this issue has no unresolved safety significance for Millstone Unit No. I and any additional action on this topic is not warranted.

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ISAP Topic No.1.42 - Main Steam Line Leakage Control System Reference

! 17 Proposed Action ISAP Topic No.1.42 addresses NRC concerns and regulations that require, in part, that piping systems penetrating containment be provided with leak detection, isolation and containment capability having redundancy, reliability and performance capabilities that reflect the importance of isolating these piping systems.

In Reference 17, NNECO addressed the applicability of NRC Task C-8(a) to Millstone Unit No.1. In the referenced letter, NNECO noted that Millstone Unit No. I does not employ a main steam isolation valve leakage control system (MSIVLCS), however main steam isolation valve leakage is limited to 11.5 SCFH.

In addition, a review of the performance of the MSIVs indicated no individual MSIV leakage rate in excess of 40 SCFH. This is in sharp contrast to other plants where leakage rates in excess of 3,400 SCFH have been experienced.

l NNECO concluded that since Millstone Unit No. I does not employ a MSIVLCS, concerns regarding adverse impact of system operation on dose consequences in the event of a postulated DBA are not applicable to Millstone Unit No.1. In addition, although excessive MSIV leakage rates have been experienced at some nuclear power plants, that is not the case for Millstone Unit No.1. Existing ~

maintenance practices are adequate to ensure valve operability and limit MSIV leakage.

, Based on the above, NNECO believes that no need exists for addition of a 4

MSIVLCS at Millstone Unit No. I and any further action on this topic is not warranted.

(a)NRC Task C-8 was initiated to investigate whether:

o The MSIV Leakage Control System recommended in Regulatory

! , Guide 1.96, Rev. I is acceptable; and 3

o MSIVLCS should be backfitted to BWRs that do not have such systems. ,

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ISAP Topic No.1.43 - Water Hammer Reference 30 Proposed Action ISAP Topic No.1.43 addresses water hammer concerns which are noted in USI A-

, 1. In Reference 30, NNECO provided the NRC with an update of the status of USI A-1 for Millstone Unit No.1. In the reference letter, NNECO noted the actions which were being contemplated to address water hammer concerns.

These actions were:

a. Implement a feedwater pump trip on high reactor level.

b. Lower the RWCU system isolation setpoint to provide system availability for reactor vessel fill and address associated post-accident radiological concerns.

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c. Evaluate lowering the programmed reactor water level setpoint following reactor trip to avoid vessel overfill.

d. Implement low flow feedwater controller improvements to a!!ow remote system actuation from the control room and maintain constant reactor vessel water level automatically.

Items (a) and (d) have been implemented, item (b) is being evaluated as ISAP Topic 2.20 and item (c) was evaluated and found to be of minimal benefit. The referenced letter also noted the NRC Staff conclusion that NNECO's sfiort-term and long-term corrective actions for water hammer events assure that Millstone Unit No. I can continue to operate without endangering the health and safety of the public.

Based on the above, NNECO believes that any further action (with the exception of ISAP Topic 2.20) on this topic is not warranted.

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ISAP Topic No. l.44 - Asymmetric Blowdown Loads on Reactor Systems Reference Proposed Action Unresolved Safety Issue (USI) A-2, " Asymmetric Blowdown Loads on Reactor Systems" was determined by the NRC to have generic implications for all PWRs.

Although not directly applicable to BWRs, Reference 30 provided the Staff with an update of the status of USI A-2 for Millstone Unit No.1. In the referenced letter NNECO concluded that, based on Section I of NUREG-0609 which states:

"Although similar loads associated with a postulated rupture of piping in primary systems in boiling water reactors (BWRs) are expected to occur, the overall safety significance is considered to be much less because of lower operating pressures in primary systems in BWRs."

USI A-2 has no safety implications and is not applicable to Millstone Unit No.1.

Therefore, NNECO believes any further action on this topic is not warranted.

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I ISAP Topic No.1.46 - Determination of SRV Pool Dynamic Loads Reference

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Proposed Action During the conduct of a large scale testing program for an advanced design BWR pressure suppression containment system (MARK III), new suppression pool hydrodynamic loads associated with a postulated LOCA were identified which had not been explicitly included in the original design of the MARK I containment systems. This item was subsequently determined to be an unresolved safety issue (USI A-39). -

Reference 20 provided an update of the status of USI A-39 for Millstone Unit No. 1. In the referenced letter, NNECO concluded that based on NNECO and NRC reviews of this issue, NNECO considered USI A-39 resolved for Millstone Unit No.1. Therefore NNECO believes any further action on this topic is.not

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ISAP Topic No.1.47 - Containment Emergency Sump Performance Reference roposed Action Unresolved Safety Issue (USI) A-43, " Containment Emergency Sump Performance," safety concerns deals with post-accident conditions which could degrade long-term recirculation capability. Reference 51 provided the status of USI A-43 for Millstone Unit No.1. In the referenced letter NNECO concluded that based on NNECO and NRC reviews of this issue, USI A-43 no longer has any safety significance for Millstone Unit No.1. Therefore NNECO believes further action on this topic is not warranted.

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I ISAP Topic No. 2.01 - LPCI Remotely Operated Valves 1-LP-50A and B References 1,39,49 ,

Proposed Action ISAP Topic No. 2.01 addresses the Millstone Unit No.1 Technical Specifications which require that the terus water level be. checked once every shift to assure that the level is within allowable limits. When the torus water level exceeds the maximum allowable level, the water is drained to the radwaste system by the opening of a series of 2 normally closed 3 inch manual gate valves 1-LP-50A and B. During certain accident conditions these valves may be inaccessible. As a result NNECO initiated a feasibility study to evaluate:

a. Adding motor operators to the valves; or

b. Adding reach rods to the valves to manually operate them from remote locations. -

NNECO's evaluation of this issue has concluded that adding remote operation capability to these valves will have no impact on core melt or radiation sequences as the valves and their associated downstream piping are too small to be censidered for a potential torus b!eed-and-feed long-term decay heat removal scheme. Additionally, if it were decided to add motor operators to the valves, maintenance of the valves would be affected by the addition of new surveillance testing to ensure the operability of the installed motor operatdrs. The NRC's contractor review of this issue also concluded the benefit of this project was negligible. Based on the above, NNECO concludes that further action on this topic is not warranted.

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I ISAP Topic No. 2.04 - High Steam Flow Setooint Increase l l

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3,38,49 I Proposed Action i

ISAP Topic No. 2.04 addresses weekly surveillance testing of the turbine-stop valves. This weekly surveillance testing requires reducing core power from 100%

to 90% to prevent main steam isolation valve closure due to the resultant high

,1 steam flows in the nonaffected steamliner. The proposed project is to increase the MSIV closure setpoint from 120% to 140% of rated steam flow in order to eliminate the need to reduce power for the weekly turbine stop valve testing. .

I In evaluating the raising of the high steam flow setpoints for MSIV closure it was recognized that the potential exists for certain main steamline breaks (outside containment) not being automatically detected and mitigating actions taken.

This would have the effect of replacing automatic action (for certain size breaks) with a manual action. NNECO's evaluation of this issue indicated a slight increase in core melt frequency and thus a slight increase in public risk.

The NRC contractor's review of this issue also concluded that a slight increase in

risk to the public would result from this project. However implementation of

this project would contribute to an overall higher capacity factor for Millstone 1 Unit No. I by eliminating the need to reduce reactor power on a weekly basis.

l Overall, NNECO concludes that the current 120% setpoint is optimum and that further action on this issue is not warranted.

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6 ISAP Topic No. 2.13 - Turbine Water Induction Modifications ISAP Topic No. 2.19 - DC System Review Reference 13 4 Proposed Action As noted in Reference 13, ISAP Topics 2.13 and 2.19 have been cancelled and thus are not being evaluated within the framework of ISAP.

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ISAP Topic No. 2.14 - Evaluation and Implementation of NUREG-0577 Reference 11 - ,7 Proposed Action As noted in Reference 11, the NRC stated in NUREG-0577 that this issue was not applicable to BWRs. Based upon this information, NNECO considered any further action on this topic to be unnecessary.

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I ISAP Topic No. 2.31 - LPCI Lube Oil Cooler Test Frequency References 9,40 Proposed Action The Millstone Unit No.1 Probabilistic Safety Study identified that one of the major contributors to low pres:ure coolant injection (LPCI) system unavailability is the failure of the solenoid valve controlling the LPCI pump motor bearing tube oil cooling. The LPCI pumps are started on a monthly basis for surveillance testing. However in the tests, the pumps are run only for a short time which does not confirm that the solenoid valves have opened to allow cooling flow 'to the tube oil.

NNECO proposed to change the surveillance testing procedure of the LPCI system in order to enable personnel to confirm opening of. the solenoid valve during LPCI system operation. As noted in Reference 9, in response to the potential safety concern surrounding operation of the LPCI lube oil cooler solenoid valves, NNECO has modified the LPCI pump operability procedures to assure that the solenoid valves operate as intended and LPCI cooling remains available.

Based on the above, the objective of the topic has been satisfied and NNECO believes that no further action on this topic is warranted.

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1 Docket No. 50-245 -

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i Attachment 3 j List of References Noted in Attachments 1 and 2 I

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February 1986 i

Table of References Letter 4

Reference From H Title Number Date i

1 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil641 August 13, 1985 Program

!2 3. F. Opeka C. I. G rimes Integrated Safety Assessment B11638 August 13, 1985 l Program 3 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11642 August 13, 1985 Program 14 3. F. Opeka C. I. G rimes Integrated Safety Assessment Bil664 August 26,1985

Program i

15 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11665 '

September 6,19S5 Program

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j6 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11676 September 9,1985 Program

7 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11695 September 13,198

Program 8 3. F. Opeka C. I. Grimes Integrated Safet'y Assessment Bil715 September 26,198:

Program 9 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil714 September 30,198 -

Program 10 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil716 October 1,1985 Program Il J. F. Opeka C. I. Grimes Integrated Safety Assessment Bil717 October 2,'1985 Program 12 3. F. Opeka C. I. G rimes Integrated Safety Assessment Bil747 October 9,1985 i Program

'I3 3. F. Opeka C. I. Grimes. Integrated Safety Assessment Bil776 October 9,1985 Program i

14 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil795 October 11,1985 Program 15 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11800 October 15,1985 ,

Program - '

$6 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil818 October 16, 1985 i j Program l

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Letter Rrferrnce From lo,o Title Number Date

!17 3. F. Opaka C. f. Grimes Integrated S,afety Assessment Bil822 October 16, 1985 Program 18 3. F. Opeka C. I. G rimes Integrated Safety Assessment B11801 October 16,' l'985 Program

.19 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11808 October 16, 1985 Program 20 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil816 October 16,1985 Program -

21 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil821 October 16, 1985 Program 22 3. F. Opeka 'C. I. G rimes Integrated Safety Assessment B11819 October 16, 1985 Program 23 3. F. Opeka C. I. G rimes Integrated Safety Assessment Bil807 October 16, 1985 Program 24 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11799 October 16, 1985 Program 25 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil820 October 16, 1985 Program 86 3. F. Opeka C. I. G rimes Integrated Safety Assessment B11806 October 16, 1985 Program l87 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil794 October 16, 1985

, Program 28 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11804 October 22, 1985 Program 29 3. F. Opeka C. I. G rimes Integrated Safety Assessment B11809 October 22,1985 Program 30 3. F. Opeka C. I. Grimes Integrated Safety Assessment Bil805 October 23,1985 Program 331 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11696 October 24, 1985 Program l

@2 3. F. Opeka C. I. Grimes Integrated' Safety Assessment B11828 October 24, 1985  :

Program 93 3. F. Opeka C. I. Grimes Integrated Safety Assessment B11838 October 25, 1985 i Program -

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1 Refercnce From To Title Number Date 34 3. F. Opeka C. I. G rimes Integrated Safety Assessment .Bil856 November 3,1985 Program 35 3. F. Opeka ' C. I. G rimes Integrated Safety Assessment B11766 November 25,1985 Program 36 3. F. Opeka C. I. Grimes TS Changes to Address Bil668 August 20, 1985 10CFR50.72 and 10CFR50.73 37 3. F. Opeka C. I. G rimes Summaries of Public Safety Bil633 August 7,1985 Impact Model Project Analyses -

38 3. F. Opeka C. I. G rimes Summaries of Public Safety B11650 August 15,1985 Impact Model Project Analyses 59 3. F. Opeka C. I. G rimes Summaries of Public Safety B11662 August 23,1985 Impact Model Projec't Analyses 60 3. F. Opeka C. I. G rimes Summaries of Public Safety B11666 August 26, 1985 Impact Model Project Analyses il 3. F. Opeka C. I. G rimes Summaries of Public Safety B11682 September 6,1935

. Impact Model Project Analyses 62 3. F. Opeka C. I. G rimes Summaries of Public Safety - B11683 September 6,1985 Impact Model Project Analyses L3 3. F. Opeka C. I. Grimes Summaries of Public Safety B11684 September 6,1985 Impact Model Project Analyses 4 3. F. Opeka C. I. Grimes Summaries of Public Safety Bil699 September 12,1985 Impact Model Project Analyses S 3. F. Opeka C. I. Grimes Summaf.es f Public Safety Bil681 October 17, 1985 ImpM Mn 1 Project Analyses 3 3. F. Opeka C. I. Grimes ISM Topic No.1.19, B11939 January 6,1986 Integrated Structural Analysis F 3. F. Opeka C. I. Grimes Supplement to ISAP Topic A05437 January 13, 1986 i No.1.25 3 3. A. Zwolinski 3. F. Opeka NRC Staff's Evaluation of LS05 September 16, 1985 NNECO's Resolution of ISAP 09-012 Issue 1.20 - MOV Interlocks C. I. Grimes R. M. Kacich NRC Final Report on PRA -

January 3,1986 Review of ISAP issues l m --- r- ,.ww- w w e m - -w-wq -~ -e,-

Letter R-ference From To Title Number Date 50 3. F. Opeka C. I. G rimes / Request for Additional .A05364 January 10, 1936 A. C. Thadani information - Generic Letter 83-28, Jtem 1.2 91 3. F. Opeka C. I. G rimes Integrated Safety Assessment Bil823 October 16, 1985 Program 52 3. P. Opeka C. I. G rimes Environmental Qualification Bil951 January 17, 1986 of Electrical Equipment -

Request for Exemption 93 F. 3. Miraglia 3. F. Opeka ATWS Rule Schedule Required -

December 13,1985 by 10CFR50.62 54 H., L. Thompson 3. F. Opeka Exemption from Submittal LS05 November 22,1985 Date for an Updated Final 011-034 Safety Analysis Repdrt (FSAR) 55 J. F. Opeka H. L. Thompson Exemptions from 10CFR50.71(e), A05282 October 11,1985 FSAR Updates 56 R. M. Bernero BWR Licensees Safety Concerns Associated -

January 3,1936 with Pipe Breaks in the BWR Scram System ,

17 3. F. Opeka C. I. Grimes ATWS Rule Schedule Required A05442 February 15,1986 by 10CFR50.62 58 3. F. Opeka C. I. G rimes 15AP Topic 1.19 -Integrated A05395 February 4,1986 Structural Analysis 59 D. C. Switzer K. R. Goller

Subject:

Millstone Unit -

November 14, 1975 No.1 Compliance with Appendix 3 50 3. F. Opeka C. I. G rimes Pressure-Temperature B11923 December 17,1985 Operating Limits Curves -

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t Docket No. 50-245 B11992 Attachment 4 Summary Table of Northeast Nuclear Energy Company's Preliminary Proposal for Disposition of - -

Millstone Unit No.1 ISAP Topics l

February 1986

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i ISAP Topics Warranting Further Activity ISAP Topic No. Title 4

1.02 Tornado Missile Protection 1.05 Ventilation System Modifications

. 1.06 Seismic Qualification of Safety-Related Piping 1.07 Control Room Design Review 1.08 Safety Parameter Display System 1.09 Regulatory Guide 1.97 Instrumentation 1.10 Emergency Response Facilities Instrumentation 1.12 Control Room Habitability 1.13 BWR Vessel Water LevelInstrumentation 1.14 Appendix J Modifications *

, 1.15 FSAR Opdate 1.16.1 MP1/MP2 Backfeed 1.16.2 Modify CRD Pumps 1.16.3 Alternative Cooling for Shut.down Cooling 1.16.4 Power Cold Shutdown Equipment 1.36 TS Covered by Generic Letter 83-36 1.45 Systems Interaction 1.48 . Safety Factor for Penetration X-10A 1.49 Reactor Vessel Surveillance Program 2.02 Drywell Humidity Instrumentation i 2.03 Process Computer Replacement 2.05 Hydrogen Water Chemistry Study 2.06 Condenser Retube ~

2.07 Sodium Hypochlorite System 2.08 Extraction Steam Piping 2.09 Upgrading of Piping and Instrumentation Diagrams 2.10 Drywell Ventilation System ,

2.11 Stud Tentioners 2.12 l Reactor Vessel Head Stand Relocation 2.15 Torque Switch Evaluation for MOVs 2.16 Reactor Protection Trip System 2.17 4.16 kV,480V and 125Vdc Plant Distribution Protection 2.18 Spent Fuel Pool Storage Racks / Transportation Cask 2.20 RWCU System Isolation Setpoint' Reduction 2.21 480V Load Center Replacement of Oil Filled Breaker

, 2.22 Control Rod Drive System Water Hammer Analysis 2.23 Instrument, Service and Breathing Air improvements 2.24 Offsite Power Systems 2.25 Drywell Temperature Monitoring System Upgrade 2.26 Reliability Equipment 2.27 Spare Recirculation Pump Motor l 2.28 Long Term Cooling Study i 2.29 FWCI Assessment Study

, 2.30 MSIV Closure Test Frequency  !

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ISAP Topics Not Warranting Further Activity ISAP Topic No. Title 1.01 Gas Turbine Generator Start Logic Modifications 1.03 '

Containment Isolation - Appendix A Modifications 1.04 RWCU System Pressure Interlock 1.05 Ventilaticn System Modifications 1.11 Post-Accident Hydrogen Monitor 1.17 Replacement of Motor Operated Valves 1.18 ATWS 1.19 Integrated Structural Analysis 1.20 MOV Interlocks 1.21 Fault Transfers -

1.22 Electrical Isolation 1.23 Grid Separation Procedures 1.24 Emergency Power ,

1.25 Degraded Grid Voltage Procedures 1.26 Item 2.1 - Equipment Classification / Vendor Interf ace 1.27 Items 3.1.1 and 3.1.2 - Post-Mairitenance Testing 1.28 Item 3.1.3 - Post-Maintenance Testing TS Changes 1.29 Response to Generic Letter 81-34 1.30 Item 1.2 - Post-Trip Review Data and Information 1.31 Item 2.2 - Equipment Classification / Vendor Interface 1.32 Items 3.2.1 and 3.2.2 - Post. Maintenance Testing Procedures 1.33 Item 3.2.3 - Post-Maintenance Testing TS Changes 1.34 Items 4.5.2 and 4.5.3 - Reactor Trip System Testing 1.35 Item 4.5.1 - Reactor System Functional Testing ,

1.37 TS Affected by 50.72 and 50.73 (Generic Letter 83-43)

, 1.38 Expand QA List 1.39 Radiation Protection Plans 1.40 Bolting Degradation or Failure 1.41 Flooding of Compartments by Backflow 1.42 MSL Leakage Control Systems 1.43 Water Hammer 1.44 Asymmetric Blowdown Loads on Reactor Systems 1.46 Determination of SRY Pool Dynamic Loads 1.47 Containment Emergency Sump Performance 2.01 LPCI Remotely Operated Valve 1-LP-50A and 50B 2.0'4 High Steam Flow Setpoint increase 2.13 Turbine Water Induction Modifications 2.14 Evaluation and Implementation of NUREG-0577 2.19 DC System Review 2.31 . LPCI Lube Oil Cooler Test Frequency

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4 ISAP Topic No.1.17 - Replacement of Motor-Operated Valves l

References l 19,34,45,52 Proposed Action References 19 and 35 provided NNECO's evaluation of this topic. In Reference 52, NNECO formally requested a permanent exemption from 10CFR50.49 for the 11 remaining motor-operated valves which the Staff requires to be environmentally qualified. NNECO considers this issue to be resolved and pending NRC concurrence with our exemption request, no further effort on this topic is warranted. .

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ISAP Topic No.1.18 - ATWS .

References 2,39,49,53,57 Proposed Action  ;

10CFR50.62 was promulgated to require additional protection against an anticipated traspient without scram (ATWS). 10CFR50.62 specified 3 separate requirements that are applicable to BWRs. These are:

1. Each BWR must have an alternate rod insertion system which is diverse from the reactor protection system to provide an alternate path for actuating the scram pilot valves. -

2. A recirculation pump trip to automatically trip the recirculation

! pumps.

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! 3. A standby liquid control system (SLCS) with a minimum flow capacity and boron coolant equivalent of 13 weight i iercent sodium pentaborate solution.

As noted in Reference 2, Millstone Unit No.1 is currently in compliance with the

, requirements of 10CFR50.62 concerning alternate rod ir.sertion and recirculation pump trip systems. Millstone Unit N.o. I does not meet the SLCS capacity requirement of 10CFR50.62.

In Reference 39, NNECO provided the NRC Staff with a pr6babilistic risk-oriented evaluation of utilizing an equivalent 86 gpm flow rate and 13 weight percent sodium pentaborate solution in the Millstone Unit No.1 SLCS. The results of this evaluation concluded that upgrading the SLCS to an 86 gpm and i 13 weight percent sodium pentaborate solution equivalent SLCS would yield a slight decrease in core melt frequency at Millstone Unit No.1. In addition, the NRC contractors review of this issue also concluded that the benefit of upgrading the SLCS system is minimal and is a candidate to be dropped.

In Peference 53, the NRC requested NNECO to:

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a. Provide a date when Millstone Unit No. I would be brought into compliance with 10CFR50.62, or; i

b. Provide a request for an exemption from the schedular requirements of 10CFR50.62 or;

c. Provide the date by which a request for a permanent exemption to 10CFR50.62 will be submitted to the NRC.

In Reference 57 we notified the Staff of our intention to request an exempt!on from 10CFR50.62 for Millstone Unit No. l.- As noted in that letter, we plan to request a permanent exemption by May 30, 1986.

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Based on the' above, NNECO believes that with the exception of NNECO l requesting and receiving an exemption to 10CFR50.62 for Millstone Unit No.1,  !

no further action is necessary on this topic. 1

ISAP Topic No.1.19 -Integrated Structural Analysis 1 &

Refererces  : J4, 18,46,58 Proposed Action In February 1983, the NRC iss ed the Integrated Plant Safety Assessment Report QPSAR) of Millstone Unit No.1. The IPSAR identified a number of open issues resulting from reviews of the fullowing SEP topics, including:

SEP Topic II-3.B, Flooding Poter.tial and Protection Requirements SEP Topic II-4.F, Settlement of Foundations and Buried Equipment -

SEP Topic III-2, Wind and Tornado Loadings SEP Topic III-3.A, Effectr, of High Water Level on Structures SEP Topic III-6, Seismic Design Considerations SEP Topic III-7.B, Design Codes, Design Criteria and Load Combinatiuns Subsequently these structural topics were considered to be addressed via an Integrated Structural Assessment Program. This topic was subsequently identified as ISAP Topic No.1.19.

In References 18,46 and 38, NNECO provided the NRC Staff with information on the remaining structural issues identified as open under the SEP. Based on the information provided in Reference 18 and the additional information provided ~in References 46 and 58, NNECO believes that the pending NRC resolution of the remaining structural issues identified in the IPSAR concludes the Integrated Structural Analysis for Millstone Unit No.1.

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ISAP Topic No.1.20 - MOV Interlocks 3 ,, A Proposed Action 5AP Topic No.1.20 addresses concerns related to thermal overload protection for motors of motor-operated valves. Reference 2 transmitted NNECO's evaluation of this topic for Millstone Unit No.1. In Reference 45, the NRC Staff !ssued a final safety evaluation of NNECO's resolution of this topic and concluded that the issue was satisfactorily resolved.

Based on the above, NNECO concludes that no further effort on this topic is warranted.

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ISAP Topic No.1.21 - Fault Transfers '

l References 24,34,41,49 i Proposed Action SEP Topic VI-7.C.1, " Independence of Reduridar.t Onsite Power - Systems" l l reviewed the AC and DC power systems at Millstone Unit No.1. The safety objective of this SEP topic was to ensure that the onsite electrical power supplies and the onsite distribution systems have sufficient independence to perform their safety functions assuming a single failure. ,

The current Millstone Unit No. I design utilizes automatic bus transfer (ABT) I switches to assure that certain vital electrical loads receive power from redundant sources. Because of this design, concerns have been raised th:2! the

] parallel operation of redundant power sources could result in their common mode r

failure under faulted modes of operation., In particular, if a fault were present on any of the loads which are connected to one sourde of power and this results in a low voltage condition, the ABT could potentially transfer the same fault to the redundant power source. Such a transfer could cause the failure of both sources due to the, subsequent protective breaker isolation function. The worst case that could occur is when each of the redundant supply buses was powered separately by one of the two sources of site emergency power (e.g., gas turbine or diesel generator). At Millstone Unit No.1 there are 7 ABTs which fall into this category.

In R'eferences 24 and 41, NNECO provided the NRC Staff with an evaluation of this issue. Of the 7 ABTs, one is presently disarmed and is not available to

automatically transfer the loads of one load group to the power source of another. The 6 remaining ABTs were evaluated t.nd NNECO concluded that i

disarming or removal of these 6 ABTs was not practical due to the safety function of the devices. Reference 41 provided the NRC Staff with a I probabilistic risk-oriented. evaluation which f.howed either no significant impact j on public risk or an actual slight increase in public risk if the 6 ABTs were

! disarmed. The NRC contractor's review of this issue noted that while the removal of the ABTs would result in the elimination of any possibility of l transferring faults from one electrical power source to another, the benefit )

would be offset by a reduction in the reliability of the LPCI system in both the injection and cooling mode.

i Based on the above, NNECO believes no further action on this topic is warranted. .

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ISAP Topic No.1.22 - Electrical Isolation

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Pro sed Action i

SEP Topic VII-1.A, " Isolation of Reactor Protection System from Nonsafety l Systems 3 Including Qualification of Isolation Devices" and Vll-2, "ESF System i 1

Control Logic and Design," reviewed electrical isolation provisions at Millstone l Unit No. 't. The SEP reviews concluded that the existing plant design met '

current licensing criteria with the following four exceptions:

l 1. ThNe are no isolation devices between the nuclear flux monitoring i systeins and the process recorders and indicating instruments. j

2. Isolation devices are not provided to isolate the APRM system from

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3. The power xsupplies for the RPS channels do not qualify as IE equipment. (solatien between each RPS channel and its respective power supply is inadequate.

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4. Isolation devicel are not provided to isolate the main steam line

radiation monitors from nonsaf ety-related indicators and recorders.

In Reference 23, NNECO prokided the NRC Staff with an evaluation of these 1

4 open items. NNECO noted that resolution of items 3 and 4 was previously provided to and accepted by t!% NRC. For items 1 and 2 NNECO had also previously provided to the Staff a tesolution for the Staff's concerns. As noted in Reference 23, we have not received a response to items I and 2 from the 4

NRC.

Based on the above, NNECO conclude's that all areas of this topic have been

addressed ar.d pending the NRC's final resolution of items 1 and 2 noted above, no further action on this topic is necessary,

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IS AP Topic No.1.23 - Grid Separation Procedures

!! AP Topic No.1.25 - Degraded Grid Voltage Procedures P

Reference ,i

12,47

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Proposed Action I

! ISAP Topics 1.23 and 1.25 address concerns related to a degraded grid voltage l

condition occurring at Millstone Unit No.1. Specific concerns related to the potential for equipment damage to occur if the grid voltage were to degrade and become low enough so that the voltage at the Class IE equipment is less than its qualified operating voltage. .

Reference 12 provided the NRC with interim degraded grid voltage procederes which define operator actions during such conditions. These procedures are to be j used pending completion of the Millstone Unit No. I degraded grid voltage design modifications. These modifications were to have been implemented during the 1985 Millstone Unit No. I refueling outage but were delayed to the next refueling outage currently scheduled for the summer 1987. This delay was based on a study derived from the Millstone Unit No.1 PSS that concluded that the probability of a station blackout would increase by approximately a factor of 2.4 -

if the planned modifications were completed. In Reference 47, NNECO provided

the Staff with details of the probabilistic eva!uation of the proposed degraded grid modifications.

In the interim NNECO believes t' it the interim degraded grid procedures provide technically acceptable means for defining operator actions during degraded grid voltage conditions without concurrent LOCA conditions. As a result, NNECC believes any additional effort on this subject, except for revising and implementing the degraded grid design modification,is not warranted.

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, ISAP Topic No.1.24 - Emergency Power Reference Proposed Action During the SEP review of Millstone Unit No.1, the NRC concluded that failure of the gas turbine generator appeared in approximately one-quarter of the dominant accideat sequences. Based on a review of 31 reported failures of the gas turbine generator over a 12-year period, the NRC concluded that some of these failures could have been prevented by a more effective preventive maintenance program. .

In Reference 2, NNECO provided the NRC Staff with a review of the adequacy of NNECO's gas turbine generator preventative maintenance program. In the referenced letter, NNECO detailed the thorough review of the gas turbine generator preventative maintenance program. In addition, NNECO identified several areas where improvements were warranted 'and the corrective actions which have been implemented.

, Based on the above, and the fact that recent experience with gas turbine 1

generator surveillance indicates that the problems which have arisen in the past have been alleviated to the point where overall reliability is high, NNECO considers this issues to be resolved and believes that no further effort on this topic is warranted at the present time. Any further consideration of the adequacy of the gas turbine generator preventative maintenance program would be in conjunction with efforts on USI A-44, Station Blackout.

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ISAP Topic No.1.26 -Item 2.1, Equipment Classification / Vendor Interface ISAP Topic No.1.27 -Item 3.1.1 and 3.1.2, Post-Maintenance Testing ISAP Topic No.1.28 -Item 3.1.3, Post-Maintenance Testing TS Changes ISAP Topic No.1.30 -Item 1.2, Post-Trip Review Data and Information i

ISAP Topic No.1.31 Item 2.2, Equipment Classification / Vendor Interface ISAP Topic No.1.32 -Item 3.2.1, Post-Maintenance Testing ISAP Topic No.1.33 -Item 3.2.3, Post-Maintenance Testing TS Changes

! SAP Topic No.1.34 -Item 4.5.2, Reactor Trip System Testing ISAP Topic No.1.35 -Item 4.5.1, Reactor System Functional Testing References 6, 50 Proposed Action On July 8,1983 the NRC issued Generic Letter 83-23," Required Actions Based on Generic Implications of Salem ATWS Events." The actions required of l

licensees addressed issues related to trip system reliability and general management capability. -

As noted in the above references, NNECO believes the Millstone Unit No. I responses to Generic Letter 83-28 are complete. At this time there are no outstanding NRC requests for additional information and NNECO concludes that further action on these topics is not warranted at the present time. NNECO is awaiting receipt of a final NRC SER on.this subject. .

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ISAP Topic No.1.29 - Response to Generic Letter 81-34

Reference 7, 56 Proposed Action Generic Letter 81-34 provided guidance for ut.'lities to review and ascertain on a plant-specific basis that their scram discharge volume design was acceptable and

! that safety concerns associated with a postulated scram discharge system piping failure do not represent a dominant contribution to core melt risk.

In Reference 7, NNECO provided the NRC Staff with an evaluation of the '

Millstone Unit No.1 SDV design against the guidelines of Generic Letter 31-34 (and its accompanying NUREG-0303). In summary, the letter concluded that for Millstone Unit No.1, an SDV break accident is not a major core uncovery or risk concern. Additionally, the letter concluded that the concerns enuraerated in ~

NUREG-0803 have been completely addressed for Millstone Unit No.1.' ,

j In Reference 56, the NRC issued an SER satisfactorily resolving this issue for all

.: BWRs. Based on the above, NNECO coaciudes no further action on this topic is necessary, i

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ISAP Topic No.1.37 - TS Changes to Address 10CFR50.72 and 10CFR50.73 Reference 36 f.

g g. 8 Proposed Action

- As noted in Reference 36, NNECO has proposed revisions to the Millstone Unit

. No.1 Technical Specifications to address 10CFR50.72 and 50.73 which meet the guidance of Generic Letter 83-43. To date, however, they have not yet been issued by the NRC Stafi. Based on this, NNECO considers any further NNECO action on the this topic is not warranted and issuance of the proposed changes by the Staff will fully resolve this issue for Millstone Unit No.1. -

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ISAP Topic No.1.38 - Expand QA List Reference 22 L Proposed Action The TMI Action Plan identified that several systems important to the safety of TMI Unit 2 were not designed, fabricated and maintained at a level equivalent to their safety importance. The NRC noted this condition existed at other plants and resulted primarily from the lack of clarity in NRC guidance for graded protection. The NRC proposed to develop guidance for licensees to expand their QA lists to cover equipment important to safety and rank the equipment in order of its importance to safety.

In Reference 22, NNECO provided an assessment of the Staff's proposal on QA.

In the referenced letter we noted examples of how NNECO routinely applies management measures to nonsafety-related equipment as a matter of good engineering, construction and operation practice's. In addition NNECO emphasized our position that the terms "important to safety" and " safety-related" are in fact synonymous. We also noted that notwithstanding the definitional problem that currently exists between the NRC Staff and the industry in general, we believe that more than adequate management controls and measurer. are in place for equipment beyond the traditional safety-related i

set of systems and components utilized to adequately maintain the equipment and to adequately protect the public health and safety. As a result, NNECO believes that additional action on this topic is not warrante.d. -

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ISAP Topic No.1.39 - Radiation Protection Plans  ;

Reference 29 Proposed Action The purpose of TMI Action Plar. III.D.3.1 is to improve nuclear power plant worker radiation protection programs by better defining the criteria and responsibility for such programs. Proposed guidance and acceptance criteria for radiation protection plans have been published in draft NUREG-0761.

In Reference 29, NNECO provided the Staff with a comparison of the radiatica protectica program at Millstone to the draf t NUREG-0761. NNECO noted that although some differences do exist, all of the basic guidance of draf t NUREG-0761 is addressed in the Millstone radiation protection program. In addition, NNECO noted that the Millstone Radiation Protection Plan which is applicable ,

to the three Millstone. units was reviewed by the NRC as part of the licensing '

review of Millstone Unit No. 3 and found to be satisfactory.

Thus, NNECO considers that the Radiation Protection Plan in place at the Millstone station meets, and in many instances, exceeds the licensing criteria set forth by the NRC and any further action on this topic is unwarranted.

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l ISAP Topic No.1.40 - Bolting Degradation or Failure l Reference **

25

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1 Prowed Action ISAP Topic 1.40 addresses NRC Generic Issue 29 concerns related to bolting degradation at nuclear power plants which might proceed unnoticed to the point where botting might fail and jeopardize the safe operation of nuclear power plants.

In Reference 25, NNCGO provided the NRC with a review of the applicability and status of Generic Issue 29 to Millstone Unit No.1. In the referenced letter, NNECO addressed the recommendations contained in INPO SOER 84-5, " Bolt Degradation or Failure in Nuclear Plants." NNECO noted that the recommendations in the INPO SOER had been addressed for Millstone Unit No. I via corporate, plant or individual departmental procedures. NNECO also concluded that based on the response of Millstone Unit No. I to the recommendations outline in the INPO SOER, Generic Issue 29' was resolved for Villstone Unit No.1.

Based on the above, NNECO believes no further action on this topic is warranted, i

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ISAP Topic No. 2.25 - Drywell Temocrature Mc.nitoring System Upgrade Reference 3

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• Proposed Action ISAP Topic No. 2.25 addresses a project to upgrade the existing drywell temperature monitoring system ir. order to more accurately measure the drywell

%1k air temperature. Drywell bulk air temperature is an input into various cuntainment response analyses (including environmental qualification and design basis calculations) which must be verified on a periodic basis. _

As noted in Reference 5, it is anticipated that implementation of this project

. will have a positive effect on public safety as well as plant performance by alding in the monitoring of safety-related equipment in the drywell.

Implementation of this project is expected to facilitate verification of J compliance with 10CFR50.49 and may' allow NNECO to eliminate some unnecessary conservatism in aging calculations. This could possibly lengthen the service life of the affected 50.49 equipment.

Based on the above, NNECO believes that further effort on this tcpic is warranted.

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e ISAP Topic No. 2.26 - Reliability Equipment Reference 4

2 Proposed Action As noted in Reference 4, ISAP Topic No. 2.26 consists of the procurement of computerized UT instrumentation, vibration monitoring and diagnostic equipment, closed circuit TV cameras for visual inspections, and miscellaneous accessories for this equipment.

As the ability to perform localized tests is expected to aid NNECO/NUSCO's

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Reliability Engineering Group in assessing the performance of and determining methods for improvement of Millstone Unit No. 1, NNECO concludes that further action on this topic is warranteo.

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. 1 ISAP Topic No. 2.27 - Spare Recirculation Pump Motor l Reference i

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kl Proposed Action As noted in Reference 5, this project consists of the engineering services to repair the damaged recirculation pump motor stator and rebuild and provide a spr.re rotor for the recirculation pump motors for Millstone Unit No.1.

As implementation of this project will enable NNECO/NUSCO to expedite replacement of a recirculation pump motor, as well as give the personnel involved in these efforts experience in working with recirculation pump motors, NNECO believes that further action on this topic is warranted.

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ISAP Topic No. 2.28 - Long-Term Cooling Study j References . u a

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4,38 '

i Proposed Action The Millstone Unit No.1 Probabilistic Safety Study '(PSS) determined that

approximately 64% of the total calculated core melt frequency at Millstone Unit No. ! is due to a failure to maintain adequate long-term decay-heat ,re noval

capability. By reviewing the dominant core melt sequences, it can be seen that

, failures of shutdown cooling and alternate shutdown cooling are the major l contributors to core melt frequency. -

{ In References 4 and 38, NNECO provided the Staff with the scope of the planned long-term cooling study. The scope included:

o A review of all potential decay-heat removal schemes (including those not directly considered in the Millstone Unit No.1 PSS).

l o Plant-specific thermal hydraulics analyses of long-term core cooling and containment cooling utilizing systems and containment codes.

o Identification of potential operator actions to improve on the existing decay-heat removal capability (such as torus flooding from external sources to prolong injection) until permanent hardware improvements are incorporated. -

i o Identification of decay-heat removal systems requiring hardware modifications.

i The desired results will includes j o A refinement of system success criteria.  !

o Recommendations on possible emergency operating procedure changes.

o Identification of weaknesses of existing long-term decay-heat removal systems and recommendation of possible hardware modifications.

Based on the above, and the large contribution of calculated failure of long-term decay-heat removal to the calculated Millstone Unit No. I core melt frequency, i

NNECO concludes that further effort on this topic is warranted. '

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9-ISAP Topic No. 2.29 - FWCI Assessment Study Reference 9

Proposed Action ISAP Topic No. 2.29 addresses concerns related to the feedwater control system and the FWCI system. During the past few years, several changes to the feedwater system have been made. While implementing several of these changes concerns surrounding the accuracy of system drawings with the as-wired system have surfaced. Additionally, some previously unexpected questions related to '

feedwater control system have arisen as noted in Reference 9. -

At the present time, ISAP Topic No. 2.29 is in the form of a study with no proposed hardware modifications. The study is intended to identify discrepancies between the system drawings and as-wired system, the limits cf operation of the feedwater control system and a review of the assumptions and impact of the feedwater control system on design basis transients.

• The results of the study are expected to provide confirmation as correct or modifications required to current system wiring drawings, assessment of the impact of different modes of feedwater control system operation in the current design basis, identification of the feedwater control system /FWCI assumpt3ns utilized in the design basis of Millstone Unit No.1, and recommendations for improvements or additional documentation on the feedwater system.

As the project is in the study phase at the present time, the exact benefits cannot be quantified. Howeyer, it is ext ected that completion of this study will provide NNECO with assurances of the adequacy of the feedwater control /FWCl system or modifications that may be required to increase the utility of these systems. Thus, NNECO believes that further activity on this topic is warranted.

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! l ISAP Topic No. 2.30 - MSIV Closure Test Frecuency References -

9, 3s, 49 l-  !

Proposed Action

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As noted in Reference 9, the Millstone Unit No. I reactor protection system j (RPS) design incorporates a MSIV closure-trip function which initiates reactor

trip when one out of two taken twice coincidence logic senses closure greater than 10%, based on valve travel limit switches. This trip function is an

anticipatory trip which provides for reactor trip before the reactor ptessure and j neutron flux respond to the collapse in reactor coolant voids that accompanies

MSIV closure., ,

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At present, Technical Specification 4.1.A requires monthly surveillance testing ,

j of this trip function. The actual test performed requires 10% closure of the valve to ensure that a reactor trip signal is generated. On several occasions while performing these tests at 100% power, the individual valve being tested

, over travelled and closed causing high steam flow in the remaining 3 steamlines resulting in the generation of a main steamline isolation signal and the closure of the remaining MSIVs.

At Millstone Unit No. I several MSIV closure events have been caused by over

travelling of the valve during testing. As noted in Reference 33, the Millstone

! Unit No. 1 PSS evaluated MSIV closure events and concluded that they 1 contributed to 93% of the causes of reactor transient events with the main condenser unavailable, which account for 2.44% of the predicted core melt

} frequency. Therefore, by reducing the frequency of 10% MSIV closure testing, i the frequency of such MSIV closure events can be reduced.

i t ISAP Topic No. 2.30 addresses a change in the surveillance test procedures by i requiring that the surveillance test for 10% MSIV closure be conducted in i conjunction with the quarterly MSIV closure stroke' test required by Technical 1

Specification 4.7.D.I.C. 'In performing the MSIV closure stroke test, reactor

, power is reduced to approximately 60% to avoid high steam flows in the j steamlines not being tested.

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Increasing the testing interval may slightly reduce the reliability of the RPS j signal generated by MSIV position. However, this slight reduction in the

( reliability of the RPS signal generated by MSIV position is more than offset by a j reduction in the frequency and potential safety hazards of inadvertent MSIV

closures during testing. As noted in Reference. 33, implementation of this j procedural change will result in a net decrease in risk to the public.

i Based on the above, and the fact that implementation of a change to the

frequency of performing MSIV closure testing requires a Technical Specification

, change., rather than a hardware modification, NNECO concludes that -further action on this topic is warranted.

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Attachment 2 Preliminary Assessment of Millstone Unit No.1 ISAP Topics Not Warranting Further Activity .

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ISAP Topic No.1.01 - Gas Turbine Generator Start Logic Modifications References r 2,40

• Proposed Action ISAP Topic 1.01 concerns the protective trips utilized on the Millstone Unit No. I emergency gas turbine generator. Specifically there are several trips associated with start-up and steady-state operation of the gas turbine generator which do not use coincident logic or are not bypassed in emergency conditions.

NNECO's evaluation of this concern focused on the potential for bypassing the gas turbine generator protective trips that are not presently bypassed during emergency operation. Modifications to the gas turbine generator's start-up trip, operational trip and generator trip logic schemes are being evaluated within this topic.

Modifications to the emergency' gas turbine generator trip logic schemes are not expected to significantly increase the reliability and/or availability of the gas turbine. Proposed modifications to the trip logic schemes were found t_o provide a small reduction in core melt frequency and thus provide minimal benefit to public safety. Based on the above NNECO believes that no further action is warranted on this topic. Any further consideration of improvements to the gas turbine generator would be in conjunction with efforts on USI A-44, Station Blackout.

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ISAP Topic No.1.03 - Containment Isolation - Appendix A Modifications Reference g 16 w"i "

Proposed Action General Design Criteria (GDC) 54 through 57 of Appendix A to 10CFR50 require '

isolation provisions for the lines penetrating the primary containment to maintain an essentially leak-tight barrier against the uncontrolled release of radioactivity to the environment. GDC 54 establishes design and test requirements for leak detection provisions, the isolation function and the containment capability of the isolation barriers in lines penetrating the primary reactor containment. GDC 55, 56 and 57 establish explicit requirements for isolation valving in lines penetrating the containment. Specifically, they address the number and location of isolation valves (e.g., redundant valving with one located inside containment and the other located outside containment), valve actuation provisions (e.g., automatic or remote manual isolation valves), valve position (e.g., locked closed, er the position of greater safety in the event of an accident or power failure), and valve type (e.g., a simple check valve is not a permissible automatic isolation valve outside containment).

In Reference 16, NNECO provided the NRC with the status of Millstone Unit No. I conformance with current containment isolation licensing criteria. As noted in the reference letter, NNECO considers.this issue to be resolved with the exception of potential modifications to penetration X.204. Penetrati'on X-204 is being addressed as part of ISAP Topic No.1.14. Based on the-above, NNECO believes that no further action on this topic is warranted, s

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ISAP Topic No.1.04 - RWCU System Pressure Interlock References 2,39,49 DRF.T I Proposed Action

ISAP Topic 1.04 addresses the Staff's concern that failure of the p essure control valve in the RWCU, followed by a single failure of the one pressu. interlock in j the system, would result in a LOCA which bypasses the conta

ment. The i proposed project which is being evaluated involves installation o a second, independent pressure interlock to ensure system isolation in the eve t that the '

pressure regulating valve f alls in the wide open position. -

NNECO's evaluation of the proposed design change fccused on the sensi ivity of the RWCU system isolation function to the addition of a second ;.fessure

! Interlock Specifically, the reduction in the probability of a small break \LOCA j

and a RWCU interfacing system LOCA were evaluated. Failure of the e.isting l pressure and temperature Interlocks on the RWC0 system were evalu ted.

Should they both fall, a relief valve which discharges to the torus would ope.. to i

protect the RWCU low pressure piping. Successful opening of the relief va ve l

would result in an isolable small break LOCA to the torus. Due to the reliabilt y and sizing of t.he relief valve, an isolable LOCA is 500 times more likely to occ' than an interfacing system LOCA where the relief valve failed to open and th q

break bypasses containment. -

\

1 The results of our analyses concluded there would be a minimal reduction in the core melt frequency at Millstone Unit No. 1 if this modification was implemented. The frequency of a RWCU interfacing system LOCA is very low '

and the resultant change in small break LOCA frequency due to these modifications is also very low. In addition, the NRC contractor's review of this issue conclud e d "the addition of a redundant pressure interlock would not alter -

the probability of a failure to isolate the RWCU line in the case of a failure of the pressure regulating valve." Based on the above, NNECO concludes that no further action is warranted on this topic.

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i ISAP Topic No.1.05 - Ventilation System Modifications References 2,44,49 i Proposed Action s

ISAP Topic No.1.05 addresses physical modifications t'o the emergency power capabilities for the FWCI area space coolers and for the intake structure exhaust fans. The proposed design change consists of two parts:

o Modification of the electrical suppliers for the feedwater/FWCI area coolers so that all six area coolers are automatically sequenced onto a gas turbine-powered bus following as LNP.

o Modification to the power supply for the intake structure exhaust fans to allow automatic sequencing of one fan onto a gas turbine bus and the other fan onto a diesel generator bus.

NNECO's evaluation of the latter design change focused on the ventilation requirements of the intake structure. Our analyses showed that during a LNP event with no ventilation in the intake structure and with two service water pumps operating, the heat buildup in the intake structure would not affect the performance of the service water pumps. Additionally, it was determined that the emergency service water pumps would not be needed unless feedwater failed and manual depressurization was required. Even in this scenario, the operator would have several hours to manually connect power to the exhaust fans before activating the emergency service water pumps. As a result, sur e/aluation deterininea that an LNP load shed of the intake structure exhaust fans will not affect operation of emergency service water should it be required. The NRC contractor's review of this issue concluded that this issue had no impact on the Millstone Unit No. I core melt frequency or public and should be dropped. Based on the above, NNECO concludes that further action on this issue is not warranted.

Our evaluation of the former design change is discussed in Attachment 1.

l ISAP Topic No.1.11 - Post-Accident Hydrogen Monitor ,

Reference 35 Proposed Action Following a postulated design basis accident, limited quantitles of hydrogen gas are generated by the reaction of steam with the zircaloy fuel cladding in the core. If adequate quantitles of oxygen and hydrogen exist in the containment, there is the potential for combustion or deflagration. To reduce or eliminate the potential for combustion or deflagration,10CFR50.44 requires Mark I BWRs to have inerted conta!nments. Inerting reduces the potential for combustion or deflagration by reducing the oxygen content in the containment atmosphere to the level that it is not capable of reacting with hydrogen produced by an accident. A Combustible Gas Control Evaluation (CGCE) was performed for Millstone Unit No. I and has shown that if the initial oxygen concentration in the containment is 4% or less (the Technical Specification limit), there is no potential for combustion or deflagration during, or following a DBA.

Nevertheless, fo!!ow-up actions resulting from the TMI-2 accident have required that the ability to monitor containment hydrogen concentrations be provided (reference NUREG-0737, Item II.F.1.6). ISAP Topic No.1.11 address this issue for Millstone Unit No.1.

As noted in Reference. 35, Millstone Unit No. I has a single channel hydrogen monitor which meets all NRC criteria except redundancy, and a Post-Accident Sampling System that can also provide information on centalamtat atmosphere hy&osen concentration. Since a combustible gas mixture will not exist following a postulated design basis accident at Millstone Unit No. I and NNECO relies on the control of oxygen rather than hydrogen to prevent a combustible mixture, no operator actions are currently predicated on the ability to monitor the containment hydrogen concentration while the drywell is inerted. The NRC Staff interprets current criteria to require a two-channel containment hydrogen monitoring system.

In the referenced letter, NNECO committed to provide the Staff with additional information on this topic. We plan to submit this information by March 31,1986.

Pending submittal of this additional information, NNECO believes sufficient hydrogen monitoring capability exists at Millstone Unit No. I and further action on this topic is not warranted.

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Docket No. 50-245 B11992 .

i h's Attachment 1 Preliminary Assessment of Millstone Unit No.1 ISAP Topics Warranting Further Activity .

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February 1986

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l 5AP Topic No.1.02 - Tornado Missile Protection '

g References 1 i

18,37  !

Proposed Action ISAP Topic 1.02 concerns provision of a tornado missile protected source of make-up water to the Millstone Unit No.1 isolation condenser (IC) and the reactor pressure vessel (RPV)in order to ensure that tornado-generated missiles do not inhibit the ability to achieve and maintain a safe shutdown condition. The project under consideration proposes a design change to provide for a portable engine-driven pump and a tornado missile protected offsite water supply .to provide makeup to the IC and RPV following a LNP event caused by a tornado.

The water supply to the RPV may be needed to restore the RPV water level during depressurization as the vessel level drops due to shrinkage or possible leaks.

The IC is a valuable system for depressurizing the RPV and removing decay heat.

Following a tornado-induced LNP, makeup to the IC needs to be initiated within 40 minutes, injection to the RPV within 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. The normal sources of make-up water to the IC are the fire water system and the condensate transfer system.

However neither the fire water tanks nor the condensate storage tank are protected from tornado missiles. As a result, the proposed design change calls for a portable engine'-driven pump which can be used for IC makeup and RPV injection,if necessary. .

Use of an offsite supply of make-up water to the IC and RPV necessitates that the supply not be susceptible to the effects of the rame tornado. The source of city water for the Millstone site is located approximately 7 miles from the site, thus it is unlikely that the source will be hit by a tornado at the same time as the Millstone site. In the event of a loss of primary power to the city water system, back-up power is provided by a diesel generator. Therefore, this water supply is -

not susceptible to the effects of the same tornado. Hookup to the city water system would require installation of an underground water hydrant and a fitting into the IC make-up line.

NNECO's evaluation of the proposed design changes indicate a reduction in core melt frequency of approximately 5% In addition, the ability to utilize the IC to depressurize the RPV and remove decay heat following a LNP event is a major contributor to the overall safety of Millstone Unit No.1. Based on the above, NNECO concludes that further action on this topic is warranted, se

ISAP Topic No.1.05 - Ventilation System Modifications References 2,44,49 1

Proposed Action i l

ISAP Topic No.1.05 addresses physical modifications to the emergency power  ;

capabilities for the FWCI area space coolers and for the intake structure e xhaust  !

fans. The proposed design change consists of two parts:

o Modification of the electrical suppliers for the feedwater/FWCI area coolers so that all six area coolers are automatically sequenced onto a gas turbine-powered bus following as LNP.

o Modification to the power supply for the intake structure exhaust fans  ;

to allow automatic sequencing of one fan onto a gas turbine bus and the other fan onto a diesel generator bus.

NNECO's evaluation of the former design change focused on the FWCI system cooling requirements during normal conditions and LNP conditions. Our evaluations concluded that there was a very high risk associated with failure of the feedwater/FWCI area coolers following a LNP event. The public risk was calculated to be on the order of 10,000 man-rem due to loss of FWCI area coolers. In addition,.the NRC contractor's review agreed with our evaluation of this issue and concluded that the safety significance of this issue should be rated high. Based on the above, NNECO concludes that further action on this issue is warranted.

Our evaluation of the latter design change is discussed in Attachment 2.

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ISAP Topic No.1.06 - Seismic Qualification of Safety-Related Piping References

'T 14,39,49 i i

Proposed Action ISAP Topic No.1.06 addresses concerns related to the ability of plant structures and equipment to function following an earthquake. I&E Bulletin 79-14 required field verification of as-built safety-related piping to compare the pipe configuration and support with the design assumed in the analysis performed to show seismic qualification of the plant. NNECO's review of Millstone Unit No. I determined that several piping supports were found to be in need .of  ;

modification. NNECO divided the modifications into two categories:

o Priority modifications which were needed to qualify the piping for the operating basis earthquake. These modifications have all been completed.

o Upgrading modifications which are needed to qualify the piping for the safe shutdown earthquake.

Our evaluation of this topic addresses the remaining modifications necessary to i

qualify the piping for a safe shutdown earthquake. Approximately 1,100 modifications were identified of which approximately 300 remain. Our evaluations of these remaining modifications concluded that no overall significant improvement in public safety would be' gained through implementation of the remaining modifications. Our analyses were based on seismic experience data collected by the Seismic Qualification Utility Group which thowed that piping is not susceptible to failure due to seismic inertia i

loads. Additionally, recent seismic PRA analyses of Millstone Unit No. I vintage and newer plants have also shown that aboveground piping is generally not the dominant contributor to seismic risk. In addition, the NRC contractor's review of this issue concluded that further modification of piping- supports is not warranted and is unlikely to have any impact on plant safety. The contractor's review did recommend that a cursory review of the remaining modifications be performed to determine if any of the modifications were included due to potential displacement failure modes, particularly due to differential motions of buildings. NNECO agrees with the contractor's recommendations and concludes that a review of potential displacement failure modes and any corrective actions which may result from the review are the only aspects of this topic which warrant further effort.

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ISAP Topic No.1.07 - Control Room Design Review References

• 26,39 ~

Proposed Action Following the accident at Three Mile Island Unit No. 2, the NRC developed a number of proposed requirements to be implemented on operating reactors and on plants under construction. One of these requirements was for the operators of nuclear facilities to pnform a control room design review (CRDR). The safety issue which led to the desire to perform systematic CRDRs was the recognition that the control rooms in many nuclear power plants contain significant human engineering deficiencies.

In Reference 26, NNECO provided the NRC with a review of the status of the Millstone Unit No.1 CRDR. In this letter we noted that we intend to provide a program plan and schedule for implementing a CRDR by March 2,1987. We also noted that the expected safety benefit and optional plan for conducting this review should be considered in the integrated assessment. In Reference 39, NNECO pro.ided the NRC with a probabilistic risk-oriented evaluation of the CRDR issue which concluded that any further improvements to control room layout is not expected to provide any significant im in human reliability of an operator during an emergency. However, provementNNECO believes that conduct of a CRDR study in accordance with NUREG-0737, Supplement 1, requirement is warranted. Upon completion of the study we will evaluate any proposed modifications which may arise and disposition them accordingly. By that time, the ISAP process is expected to be sufficiently mature to adequately evaluate the plant-specific CRDR findings.

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ISAP Topic No.1.08 -Safety Parameter Display System Reference 26 Proposed Action Following the accident t.t Three Mile Island Unit No. 2, the NRC developed a number of proposed requirements to be implemented on operating reactors and on plants under construction. One of these requirements was for license,es to provide a plant safety parameter display (SPDS) console that will display to operators a minimum set of parameters defining the safety status of the plant, capable of displaying a fu!I range of important plant parameters and data trends on demand and capable of indicating when process limits are being approached or exceeded.

In Reference 26, NNECO provided the NRC with a review of the status of Millstone Unit No. I's activities on SPDS. In this letter we noted that we intend to provide a safety analysis, and a plan and schedule for operator training and implementation of a fully operational SPDS by April 9,1987. We also noted that the expected safety impact, need and priority for implementation of an SPDS should be determined in the integrated assessment. Our preliminary assessment of this topic indicates that the SPDS does not provide any new information to the operator and will only be beneficial in rare events such as ATWS and large break LOCA. As a result NNECO believes that further activity to study the benefits of installing an SPDS at Millstone Unit No. I should continue. Following completion of- the study an overall assessment of the value of an SPDS should be conducted to determine the implementation plan and schedule.

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.l ISAP Topic No.1.09 - Regulatory Guide 1.97 Instrumentation .

.g Reference 33 Proposed Action As required by 10CFR50, Appendix A, General Design Criterion (GDC) 13,

" Instrumentation and Control," instrumentation shall be provided to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences, and for accident conditions, as appropriate, to assure safety. The requirements of GDC 13 are supplemented by GDC 19 and GDC 64. Criterion 19, " Control Room," includes a requirement that a control room be provided from which actions can be taken to maintain a plant in a safe condition under accident conditions and that equipment, including necessary instrumentation, be provided at locations outside the control room with the capability to effect prompt hot shutdown. Criterion 64, " Monitoring Radioactivity Releases," requires that means be provided for monitoring radioactivity which may be released as a result of a pbstulated accident.

Regulatory Guide (RG) 1.97, Revision 2, " Instrumentation for Light Water-Cooled Nuclear Power Plants to Assess Plant and Environs Conditions During and Following an Accident," was issued in December 1980 to provide guidance concerning implementation of the requirements described above. RG 1.97, t

Revision 2 identifies parameters to be monitored and ranks these parameters, and their corresponding equipment qualification levels, according to the importance of their function. Further, ranges for these parameters are recommended.

In Reference 33, NNECO provided the NRC with the status of Millstone Unit.

1 No. I's conformance with RG 1.97, Revision 2. At present we have completed our review of the RG 1.97 requirements and have identified the following modifications would be required to be implemented in order to bring Millstone Unit No.1 into conformance with RG 1.97, Revision 2:

1. RPV Pressure Provide redundant instrumentation

• for the pressure range 0-2,500 psig.

2. RPV Water Level Provide redundant instrumentation

• for the level range 340 - 316".

3. Neutron Flux Provide environmentally qualified instruments with redundant power sources.

4. Drywell Temperature Provide environmentally qualified instruments.

• With separate power sources.

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5. Drywell Spray Flow Provide drywell spray flow measuring instruments which are environ-mentally qualified and redundant.

6. IC Shell-Side Valve Position Provide environmentally qualified instruments.

7. IC System Valve Position Provide environmentally qualified position indication.* *

8. Ventilation Damper Position Provide environmentally qualified and redundant position Indication.

9. Primary Containment Provide redundant power source for Isolation Valve Position valve position indication.

Most recently, we performed a probabilistic risk-oriented evaluation of the benefits of implementing these modifications. Our evaluation has shown that most of the modifications provide little or no benefit (i.e., decrease in public j

risk), while some might provide a significant benefit. As a result, NNECO believes that further effort on this topic is warranted to definitively determine which modifications provide significant benefits and thus merit implementation.

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• • Redundant power source issue is addressed in item 9. -

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ISAP Topic No.1.10 - Emergency Response Facilities Instrumentation Reference 31 Proposed Action On October 31, 1980, the NRC Staff issued NUREG-0737, which incorporated into one document all TMI-related action items approved for implementation.

Subsequently, the NRC issued Supplement I to NUREG-0737," Requirements for Emergency Response Capability," on December 17, 1982 via Generic Letter (GL) 82-33. GL 82-33 provided additional clarification regarding safety parameter display systems, detailed control room design reviews, Regulatory Guide 1.97 (Revision 2), Application to Emergency Response Facilities, Upgrade of Emergency Procedures, Emergency Response Facilities and Meteorological Data.

ISAP Topic 1.10 addresses the conformance of Millstone Unit No. I with the requirements and guidance of that portion of GL 82-33 related to Emergency Response Facility (ERF) instrumentation. The proposed project which is being i

evaluated concerns the types of instrumentation and plant status information that should be provided to ERFs following an accident event.

j Following a major plant accident, personnel are assembled to monitor the events occurring at the plant. These personnel will require data from the control room in order to analyze the incident and assess changes in the plant situation. A full complement of operations shift personnel would be called onsite to man the Technical Support Center (TSC) and Emergency Operations Facility (EOF). The TSC and EOF serve as emergency operations work areas providing back-up support to the control room by evaluating post-accident events as they occur.

At present, the information required by personnel within the TSC and EOF would be available through existing instrumentation and communications channels.

However, additional instrumentation in the TSC might help operators to mitigate events that lead to degraded long-term cooling scenarios.

Because the calculated ccre melt frequency associated with such events is relatively high, NNECO concludes that additional activity in evaluating the adequacy of the data retrieval and transmittal system should be undertaken.

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ISAP Topic No.1.12 - Control Room Habitability M

References 32,43,49 Proposed Action NUREG-0737, Item III.D.3.4, " Control Room Habitability Requirements" was issued to assure that licensees adequately protect control room operators against the effects of an accidental release of toxic or radioactive gases and that nuclear power plants can be safely shutdown under design basis accident conditions.

In Reference 32, NNECO provided the Staff with an assessment of ISAP Topic No.1.12 for Millstone Unit No.1. This assessment noted to the Staff that NNECO believed an alternate method of protecting the control room operators should be explored because of the very high cost of control room HVAC desig, modifications. Dose assessments have shown that control room operators only need additional radiation protection from thyroid-doses. Additionally, any reductions in the radioactive source term, as are being evaluated by the NRC Staff, would result in lower thyroid doses. As an alternative, self-contained breathing apparatus could protect control room operators from airborne iodine and chlorine gas and provide approximately the same benefit as the HVAC modifications at a significantly lower cost. Additionally, we are planning to remove the two chlorine tank cars (utilized in marine biofouling control) which are presently stored onsite and represent a significant toxic gas sa,fety concern.

As noted in Reference 43, we have performed a probabilistic risk-oriented evaluation of this issue and have concluded that implementation of an upgraded HVAC system at Millstone Unit No. I would provide minimal benefit in reducing the calculated core melt frequency at Millstone Unit No.1. In addition, the NRC contractor's review of this issue concluded that the benefits of this topic were insignificant a'nd recommended that this issue be dropped.

Based on the above, NNEC'O concludes that further evaluation of alternatives to the proposed HVAC modifications should be performed to identify the appropriate sc7pe of modifications which will provide an equivalent level of protection for control room operators in order to satisfy the intent of Item III.D.3.4. We currently believe that the appropriate scope of work to be implemented will be significantly less than that originally identified in response to TMI Action' Plan Item III.D.3.4.

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ISAP Topic No.1.13 - BWR Vessel Water-Level Instrumentation '

References 14,42 Proposed Action The water level instrumentation in a BWR is relied upon for controlling feedwater, actuating emergency systems and for providing the operators information which is used as a basis for actions to assure adequate core cooling.

Many of the actions in the emergency procedures guidelines are keyed to reactor water level. NUREG-0737 Item II.F.2 and Generic Letter 84-23 require that licensees modify or supplement existing equipment to assure accurate indication of vessel water level.

The reactor pressure vessel (RPV) water-level indication which is used to mitigate accidents is based on a number of level gauges which utilize one of two reference legs located within the drywell. Unusually high drywell temperature in conjunction with RPV depressurization can affect reactor water level instrumentation by causing water in the reference legs to flash when the drywell temperature reaches RPV saturation condition. This results in erroneous readings that may indicate a higher reactor water level than the actual level l which is present in the vessel. If the operator places undue reliance on the indicated water level, the potential exists that makeup could be throttled to the extent that the core is uncovered. Such scenarios are explicitly considered in .

the Millstone Unit No. I emergency procedures. .

In References 14 and 42, NNECO provided the Staff with assessments of this topic for Millstone Unit No.1. A probabilistic risk-oriented evaluation of this topic concluded that eliminating the potential for reference leg flashing at Millstone Unit No. I would yield a significant decrease in risk to the public.

NNECO is currently investigating several approaches to resolve' the reference leg flashing issue for Millstone Unit No. I and based on the above, believes that further effort on this topic is warranted.

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ISAP Topic Nc.1.14 - Appendix 3 Modifications Reference 16,59

Proposed Action Appendix 3 to 10CFR50, " Primary Reactor Containment Leakage Testing for Water-Cooled Reactors," was published on February 14, 1973. Many nuclear plants, including Millstone Unit No.1, had either received an operating license or their containments had reached advanced stages of design or construction at that time. The NRC expressed concern that these plants might not be in full compliance with the requirements of this regulation and in an August 7,1975 -

letter, the NRC requested NNECO to determine if containment leakage testing at Millstone Unit No. I was in full compliance with Appendix 3.

Specifically, NNECO was requested to identify any design features that do not permit conformance with its requirements or existing Technical Specification requirements which are in conflict with Appendix 3.~ NNECO was requested to provide a summary of planned remedial actions, i.e.,

design / Technical Specification modifications or exemption requests, as well as schedules for such actions. In Reference 39, NNECO responded to the Staff, noting the areas where Millstone Unit No.1 did not conform to 10CFR50, Appendix 3 requirem'ents.

As noted in Reference 16, NNECO is currently evaluating the overall status of Millstone Unit No. I compliance with the requirements of 10CFR50, Appendix 3.

Pending the results of this evaluation, NNECO will identify modifications necessary to bring the plant into compliance with Appendix 3 and/or submit exemption requests where necessary. As a result, NNECO believes further effort on this topic is necessary to fully resolve this issue.

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ISAP Topic No.1.15 - FSAR Update

• References 2,54,55 .

Proposed Action In a letter dated November 22,1985 (Reference 54), the NRC Staff granted NNECO im exemption until March 31,1987 for submittal of an updated FS AR for Millstone Unit No.1. This exemption was granted in response to NNECO's October 11, 1985 letter (Reference 55) which provided the NRC with schedular milestones in NNECO's FSAR update effort.

As a result, NNECO does not plan any further evaluation of the " benefits" of this topic within the framework of the ISAP. However, NNECO will consider the manpower and resource burden associated with this topic as part of the resource

. management aspect of ;he ISAP.

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l ISAP Topic No.1.16.1 - Appendix R, MPl/MP2 Backfeed '

References 27,37,49 Proposed Action This topic resulted from the review of proposed exemption requests and modifications submitted for Millstone Unit No. I to comply with 10CFR50.48 and Appendix R to 10CFR50. This topic addresses modifications which would enable the Millstone Unit No. 2 diesel generator to supply emergency power to Millstone Unit No.1 during a control room, reactor building or turbine building fire.

Similarly, the backfeed capability would allow Millstone Unit No. I to supply Millstone Unit No. 2 with shutdown power for a Millstone Unit No. 2 control room, reactor building or turbine building fire.

In Reference 27, NNECO provided the NRC with a review of the proposed modification and its applicability towards bringing the plant into compliance with 10CFR50.48 and Appendix J to 10CFR50. In Reference 37, a probabilistic risk-oriented evaluation of the proposed modification was provided to the Staff.

The results of this analysis concluded that implergen a very significant decrease in risk to the public.t!)Intation of this addition, theproject yielded NRC Staff )

contractor's review of this proposed modification also concluded that the benefits to implementing this project were significant. As a result, NNECO concludes that further activity on this topic is warranted. .

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(1) This is based on the assumption that ISAP Topic No.1.16.2 is implemented concurrently. For more information on this linkage, please refer to ISAP Topic No.1.16.2.

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i ISAP Topic No.1.16,2 - Appendix R, Modify CRD Pumps References 27,40 l

Proposed Action i

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This topic resulted from the review of proposed exemption requests and modifications submitted for Millstone Unit No. I to comply with 10CFR).43 and Appendix R to 10CFR50. This topic addresses a proposed project to modify the CRD pumps to operate without an external source of cooling water during a control room or turbine building / intake structure fire. The modification would allow manual realignment of the CRD motor cooling piping to the pump discharge flow to permit self-cooling. The CRD pump will be used to i compensate for any decrease in reactor vessel water level due to leakage and shrinkage during cooldown.

In Reference ~27, NNECO provided the NRC with a review of the proposed modification and its applicability towards bringing'the plant into compliance ,

with 10CFR50.48 and Appendix R to 10CFR50. In Reference 40, a probabilistic risk-oriented evaluation of the proposed modification was provided to the Staff.

The results of this analysis concluded that by itself the proposed CRD pump modification would provide insignificant benefit to Millstone Unit No. 1.

Conversely, the backfeed project (ISAP Topic No.1.16.1) will not accomplish its intended purpose without implementation of the CRD pump self-cooling project.

Thus the two topics are implicitly linked, together they provide a significant benefit (i.e., decrease in public risk); individually the benefits are minimal. As a result, as noted in the evaluation of ISAP Topic No.1.16.1, NNECO believes that further activity on this topic (concurrent with Topic No.1.16.1)is warranted.

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, ISAP Topic No.1.16.3 - Appendix R Alternative Cooling for Shutdown Cooling Reference

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27 1 Proposed Action This topic resulted from the review of proposed exemption requests and >

modifications submitted for Millstone Unit No. I to comply with 10CFR50.43 and Appendix R to 10CFR50. This topic addresses a proposed project to modify portions of the shutdown cooling (SDC) system to allow the plant to be cooled to cold shutdown conditions without the use of service water. This will entail modifying the reactor building closed cooling water (RBCCW) connections on the SDC pumps and heat exchangers to accept fire hoses. The fire protectio' n system t

will provide an alternate means of supplying cooling water.

In Reference 27, NNECO provided the NRC with a review of the proposed modification and its applicability towards bringing the plant into compliance with 10CFR50.48 and Appendix R to 10CFR50. - At the present time, the expected benefit of implementing this modification has not been explicitly I

addressed within the framework of the ISAP. As a result, NNECO believes that further activity on this topic is warranted.

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ISAP Topic No.1.16.4 - Appendix R, Power Cold Shutdown Equipment Reference '

6 27 Proposed Action This topic resulted from the review of proposed exemption requests and modifications submitted for Millstone Unit No. I to comply with 10CFR50.48 and Appendix R to 10CFR50. This topic addresses a proposed modification to protect the 3 drywell penetrations in the shutdown cooling (SDC) pump cubicle that contain power and control cables to the SDC isolation valve (1-SD-1), the primary containment inner isolation steam inlet valve (1-IC-1) and the primary containment inner i olation condensate return valve (1-IC-4). Repair proce.dures will be developed to provide a source of emergency power to the MOVs after a reactor building fire.

In Reference 27, NNECO pro'vided the NRC with a review of the proposed '

modification and its applicability towards bringing "the plant into compliance with 10CFR50.48 and Appendix R to 10CFR50. At the present time, the expected benefit of implementing this modification has not been explicitly addressed within the framework of the ISAP. As a result, NNECO believes that further activity on this topic is warranted.

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5AP Topic No.1.36 - TS Covered by Generic Letter 83-36 Reference i 15 i

Proposed Action l 1 Generic Letter 83-36 requested licensees to propose changes to their Technical l Specifications to address each of the TMI Action Plan items identified in l Enclosure 1 (to the Generic Letter) which were applicable to their facilities. In Reference 15. NNECO provided the NRC Staff with a list of proposed Technical Specifications for TMI Action Plan items which were applicable to Millstone Unit No. 1. NNECO plans to submit proposed Technical Specifications, consistent with Generic Letter 83-36, for the items identified in Reference 15.

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ISAP Topic No.1.45 - Systems Interactions Reference 28 Proposed Action Unresolved Safety Issue (USI) A-17, " Systems Interactions in Nuclear Power Plants," addresses the development of a systematic process to review plant systems to determine their impact on other plant systems. The purpose of the task is to identify where the present design, analysis and review procedures may not acceptably account for potentially adverse systems interactions.

Reference 28 provided an update of the status of USI A-17 for Millstone Unit No.1. In the referenced letter NNECO concluded:

"There is clearly no compelling need for NNECO to take action unique to this issue at this time. . Through identification of potential systems interactions via a) the Millstone Unit No.1 PSS, b) an assessment of the potential safety significance of this USI to Millstone Unit No.1 in the i

context of the Integrated Safety Assessment Program phase of ISAP and c) the actions noted above; NNECO feels that the appropriate actions to resolve this USI for Millstone Unit No. I are being taken."

Based on the above, NNECO believes further action on this issue may be necessary at the present time in order to fully resolve this issue. -

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ISAP Topic No.1.48 -Safety Factor for Penetration X-10A .

Reference l

30 Proposed Action ISAP Topic 1.48 addresses concerns regarding presumed degradation of the isolation condenser supply line anchor at containment penetration X-10A which  !

resulted from a water hammer incident in December 1979. Specifically, NNECO has been evaluating the safety factors for the anchor bolts of the support, the potential for degradation of the original anchor design bases and anchor improvements that would meet the required safety factors. -

Reference 30 provided the NRC with an update of the status of this evaluation.

The analytical evaluation of the factor of safety associated with the design of  ;

containment penetration anchor X-JOA is ongoing and will be submitted to the NRC upon completion. In the interim NNECO believes this issue does not represent a significant safety concern and conclud'es that completion of the analytical evaluation is the only activity necessary on this topic at the present 4

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ISAP Topic No.1.49 - Reactor Vessel Survaillance Program -

References 25,60 Proposed Action Appendix H to 10CFR50 requires a reactor vessel material surveillance program to monitor changes in the fracture toughness properties of ferritic materials in the reactor vessel belt-line region of reactors resulting from exposure to neutron irradiation and the thermal environment.

References 25 and 60 updated the status of NNECO's reactor vessel surveillance program. Reference 60 indicated that the pressure-temperature (PT) curves now i in effect for Millstone Unit No. I are valid for another 1.62 effective full power years. It was also noted that NNECO plans to evaluate the PT curves prepared by the General Electric Company and propose an amendment to the operating license for Millstone Unit No. I on or before July 1,1986. Thus, NNECO

, concludes that further action on this issue is warranted.

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I ISAP Topic No. 2.02 - Drywell Humidity Instrumentation .

Reference '

8 Proposed Action l i

Presently, Millstone Unit No. I employs 2 methods for detecting primary system leakage in the drywell. They are:

a. Calculation of an average leak flow rate through the reactor coolant pressure boundary based on the amount of liquid that is pumped to radwaste via the drywell equipment and drywell floor drain sump; and

b. Analysis of air samples obtained during venting of the drywell for increased contaminant activity.

ISAP ' Topic No. 2.02 addresses NNECO's propnsed project to perform an engineering evaluation to determine the best method for continuous monitoring of the humidity, and airborne, gaseous and particulate contamination in the drywell. The scope of the project does not include implementatirn of the study recommendations.

Neither of the 2 methods for leak detection presently being utilized at Millstone Unit No.1 provide for continuous leak detection monitoring. Due to the lack of a continuous monitoring system, there is a possibility that piping cracks might propagate to a greater degree before detection, as compared to the leakage detection sensitivity associated with a continuous monitoring system. Thus, implementation of continuous leak detection could result in a decrease in the probability of a LOCA, yielding a decrease in risk to the public. As a result, NNECO concludes that further effort on this issue is warranted.

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ISAP Topic No. 2.03 - Process Computer Replacement Reference DRAFT

. 5 Proposed Action The present Millstene Unit No. I process computer is more than 15 years old and is considered to be nearing the end of its useful life (typically the useful life of a computer is 10 to 15 years). At present the Millstone Unit No. I process computer is necessary during start-up (for rod worth minimizer system operation) ramp-up and andpowerfor monitoring ramp-down and performing conditions. NSSS During normal operation calculations above during (powe 25% power) the process computer is utilized for daily fuel surveillance; however, its failure does not affect the power generation of the plant.

As noted in Reference 5, replacement of the process compu:er is a one-for-one, plant hardware exchange. The new process computer will have a completely redundant processor with built-in diagnostics capability which could automatically switch over to the redundant processor if the operating processor ,

, should fail, and at the same time inform the operator of the failed processor.

Replacement of the process computer is expected to result in an increase in the reliability of the plant as the existing process computer is becoming less reliable due to aging. Additionally, the existing process computer maintenance costs have begun to show dramatic increases in man-hours as well as parts support. An additional benefit provided by a new process computer will be the capability to integrate a safety parameter display system (SPDS) which the existing process computer would be unable to support.

Based on the above, NNECO concludes that further effort on this activity is

warranted.

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ISAP Topic No. 2.05 - Hydrogen Water Chemistry Study 7

ysuds r Reference 1

Proposed Action ISAP Topic No. 2.05 addresses the feasibility and/or benefits to implementing a hydrogen water chemistry program at Millstone Unit No. 1. Addition of hydrogen into the primary coolant systems of BWRs has been found to be an effective method of reducing intergranular stress corrosion cracking (IGSCC) in BWR primary piping systems.

The proposed NNECO project consists of two phases; a pre-implementation test to determine the plant's response to operation on hydrogen water chemistry and pending the outcorr.e of the pre-implementation test, evaluation of a permanent hydrogen injection system.

The pre-implementation test will provide the necessary plant-specific data needed for NNECO to implement a properly functioning, permanent hydrogen water chemistry system at Millstone Unit No.1. Without the information provided by a pre-implementation test, hydrogen injection could be ineffective in the prevention of IGSCC.

NNECO believes that the pre-implementation test should be conducted in order to provide plant-specific data to use in the evaluation of installing a' permanent '

hydrogen water chemistry program. Following the pre-implementation test, a cost benefit analysis will be performed to weigh the costs of installing a permanent hydrogen water chemistry system with the associated shielding against the savings resulting from reduced in-service inspection requirements and IGSCC repairs.

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EAP Topic No. 2.06 - Condenser Retube ,

References 1 l

I, 37 Proposed Action 5AP Topic No. 2.06 addresses the possibility of completely retubing the j Millstone Unit No. I main condenser. A major aspect of the retubing effort would be to change the condenser tube material from the 70/30 copper / nickel alloy presently in use to titanium tubing.  ;

NNECO's evaluation of this project reveals that retubing of the condenser is l expected to

a. Increase plant efficiency through an increase in heat transfer  ;

capability from the primary side to the secondary side.

b. Decrease radwaste control costs through a reduction in tube leakage.

c. Decrease plant personnel exposures while performing maintenance on the condenser.

d. Reduce the times the plant has to reduce power to condenser maintenance. .

e. Result in a slight decrease in core melt frequency and p'eblic risk.

Based on the above, NNECO concludes that further action on this issue is warranted.

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ISAP Topic No. 2.07 - Sodium Hypochlorite System Referencess -

I,37,49 Proposed Action ISAP Topic No. 2.07 addresses concerns surrounding the use of and onsite storage of liquid chlorine. Presently, liquid chlorine is used as the means of marine biofouling control at Millstone Unit No.1. Liquid chlorine is stored onsite in 55 ton rail tanks cars and transported via an underground double pipe to the

, plant's intake structure where it is vaporized for injection into individual cooling 1

water systems. If the catastrophic failure of a tank car and subsequent chlorine release were to occur, the potential exists for significant impacts on both plant personnel and the public surrounding the plant.

To prevent such an accident NNECO proposes to replace the existing chlorine system with an onsite bulk storage and distribution system of sodium hypochlorite which is a minimal safety hazard. Our evaluation of this issue concluded that this modification will result in a large decrease in risk to both plant personnel end the general public living around the plant by:

a. Eliminating the risk of a chlorine-leak accident which could injure or disable plant personnel and prevent the plant from being operated in a safe manner, and -

b. Eliminating the risk of direct exposure of the public ti) poisonous gas during a chlorine-leak accident.

In addition, the NRC contractor's review of this issue identified this as a significant safety issue. Based on the above, NNECO concludes that further action on this topic is warranted.

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ISAP Topic No. 2.08 - Extraction Steam Piping gg URfL References 3,38,49 1 Proposed Action 5AP Topic No. 2.08 addresses replacement of the extraction steam piping at Millstone Unit No.1. The nuclear industry has noted a number of cases of severe erosion and failures of extraction steam piping. Additionally, NNECO undertook extensive inspections of Millstone Unit No. I to determine if steam piping erosion was occurring. As a result of identified erosion of extraction steam piping at Millstone Unit No. I a project was initiated to replace the extraction steam piping to prevent further erosion.

Replacement of eroded extraction steam piping is considered necessary to preclude injury to plant personnel which could result from steam leakage and to insure that the plant does not have to be removed from service to repair or replace degraded piping. In addition, NNECO perfor~med a safety evaluation of this proposed modification and concluded there would be a slight decrease in public safety if these modifications were not implemented . The NRC contractor's review of this issue also noted that implementation of this project would have a positive benefit to public safety. Based on the above, NNECO concludes that further action on this issue is warranted.

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1 ISAP Topic No. 2.09 - Upgrading of P& ids

, Reference 4

i Proposed Action l

ISAP Topic No. 2.09 addresses upgrading of the Piping and Instrumentation l Diagrams (P&ID) for Millstone Unit No. i. The planned effort represents a l complete upgrading of current plant information to make it consistent and more j retrievable and is expected to be a long-term project (4 - 5 years).

The benefit of upgrading the P&lDs is that existing information will be consolidated into a single accessible base of information (for example, a specific

' component may be identified by three different numbers depending on whether the drawing originated with the NSSS vendor, the architect-engineer or within NU). Such a change will increase the efficiency in what may be broadly termed the backfit design and day-to-day engineering of the plant by cutting down the time needed to chase down/ confirm information. It l's noted that plant hardware changes, plant procedures or Technical Specification changes are not part of this effort. -

i As this effort is expected to result in the creation of a consolidated, accurate i and accessible base of information, NNECO believes that further effort on this topic is warranted. .

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ISAP Topic No. 2.10 - Drywell Ventilation System - -

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1 Proposed Action During the 1982 refueling outage, cleaning and maintenance was performed on the eight drywell ventilation units and associated duct work in an attempt to reverse a trend of increasing drywell air temperature. Following the cleaning and raaintenance activities, testing was performed to compare the actual capacity and design capacity for each unit. This testing identified that a majority of the coolers were still operating significantly below their design capacity. The resultant elevated drywell temperatures could result in premature equipment degradation and could adversely impact many of the evaluations relating to equipment qualification. Equipment aging calculations would be most directly affected.

ISAP Topic No. 2.10 addresses NNECO's perfortnance of an engineering evaluation of drywell air circulation patterns, ventilation duct work and coil cooling water systems to determine methods for minimizing temperature " hot spots" and for improving air circulation. It is expected that modifications to the drywell ventilation system will result in an increase in plant reliability by slowing the rate of degradation of equipment located in the drywell and will have a positive effect on environment qualification of equipment located in the drywell. As a result, NNECO believes further activity on this topi,c is warranted.

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ISAP Topic No. 2.11 - Stud Tensioners Reference ,

3 Proposed Action ISAP Topic No. 2.11 addresses replacement of the manual vessel-head stud tensioners with an automated system. The current practice at Millstone Unit No.1 is to tension the studs on the reactor manually by screwing the tensioner to the studs. The new quick disconnect tensioners will center automatically on the individual studs and will utilize a hydraulic system to tension the studs.

The new stud tensioners will impact the duration of a refueling outage and any unplanned outage which requires reactor vessel-head removal. According to NNECO estimates the new stud tensioners will result in a large economic benefit by resulting in an approximate 24-hour decrease in the duration of a . refueling outage.

Based on the above, and the fact that the project is in the purchasing phase, NNECO concludes further activity on this topic is warranted.

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ISAP Topic No. 2.12 - Reactor Vessel-Head Stand Relocation Reference 3

Proposed Action In 1975 the reactor building crane was modined to provide redundant lift capability. This modification included the installatien of a new trolley, with a redundant crane hook, which limited the travel of the crane with respect to the refueling area floor. As a result of this modification, the centerline of the

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reactor vessel-head stand can not be lined up with the centerline of the crane.

The present method of replacing the reactor vessel-head utilizes a chain-fall and a lift pad bolted to the south wall of the reactor building, allowing personnel to drift the vessel-head into position on the vessel-head stand.

Utilizing this method to locate the vessel head could result in damage to the vessel head br personnel injury if either the chainfall or lift pad fails. As a result NNECO is evaluating the feasibility of relocating the centerline of the reactor vessel-head stand to line up the centerlines of the crane and vessel-head stand. Based on the potential for a serious injury to plant personnel involved in replacing the vessel head and damage to the reactor vessel head during refueling activities, NNECO believes further activity on this topic is warranted.

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5AP Topic No. 2.15 - Torque Switch Evaluations for MOVs ~. .

Reference 11-Proposed Action On February 21, 1984, the NRC issued I&E Information Notice 84-10, " Motor-Operated Valve Torque Switches Set Below the Manufacturer's Recommended Value" to notify all licensees of a recent related experience at GPU Nuclear Corporation. Based on this notice, NNECO initiated ISAP Topic No. 2.15 to investigate all Millstone Unit No. I safety-related MOVs to determine if their existing torque switch setpoints e.re within the manufacturer's recommended setpoint range. NNECO also intends to develop an inspection procedure for use in monitoring MOV torque switch setpoints during refueling outages.

The review conducted as part of this project is expected to result in increased assurance of proper operation of safety-related motor-operated valves during postulated accident conditions. Further, the devdlopment of an inspection procedure for use during future refueling outages will provide additional assurance that torque switch setpoints are maintained at appropriate values. As a result, NNECO believes further effort on this topic is warranted.

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. s ISAP Topic No. 2.16 - Reactor Protection Trip System Reference R0 8

i Proposed Action

, Historically, the reactor protection trip system (RPS) has demonstrated setpoint '

drift problems which have led to difficulties in maintaining setpoint calibration and accuracy. ISAP Topic No. 2.16 addresses:

a. Investigation of the possible replacement of the 120-second automatic depressurization system (ADS) timer with a current state-of-the-art timer circuit, to alleviate the RPS setpoint drif t problem.

b. Removal of two low reactor pressure permissive switches (PS 54A and B) from the ECCS pump start logic to allow ECCS pump start on either high drywe'll pressure or low-low water level, and automatic '

ADS actuation when required.

Replacement of the present 120-second ADS timer with a more accurate state-of-the-art timer circuit will reduce the risk of operation of the ADS outside of operating setpoint guidelines. Removal of the two low reactor pressure permissive switches would alleviate potential safety impact problems due to their 120 psig drift. Additionally, removal of these switches will make manual -

depressurization of the reactor (per the EOPs) easier in that t!)e ECCS pumps will already be running when the EOPs direct the operator to manually depressurize the reactor coolant system.

Thus it is expected that implementation of these two modifications will result in a decrease in risk to the public 'and NNECO believes that additional effort on this topic is warranted.

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ISAP Topic No. 2.17 - 4.16 kV,480 kV and 125 VDC Plant Distribution Protection Reference 10 Proposed Action As part of NNECO's Appendix R rereview completed in early 1985, the Breaker Coordination Study was updated to ensure that the power supplies for components required for safe shutdown in the event of an Appendix R-type fire, would be available despite the failure of circuits not required for safe shutdown.

Based on the results of the Breaker Coordination Study, NNECO initiated ISAP Topic No. 2.17. .

In Reference 10, NNECO provided the Staff with the scope of an extensive review of 4.16 kV, 480 V and 125 VDC plant distribution systems. This review exceeds the scope of work associated with the Breaker Coordination Study performed as part of the Appendix R review. The review being conducted is expected to verify the existing de:!gn adequacy of the 4.16 kV, 480 V and 125 VDC plant protection systems. An additional benefit provided by- this evaluation will be an expected increase' in the overall reliability of these protective systems. Based on the above, NNECO believes that further effort on this topic is warranted.

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l ISAP Topic No. 2.18 - Spent Fuel Pool Storage Racks / Transportation Cask j Reference 13 $

Proposed Action Under the Nuclear Waste Policy Act of 1982, it is the responsibility of industry to provide interim storage for its spent fuel until long-term spent fuel storage becomes available. At Millstone Unit No.1, if the spent fuel storage capacity is not increased, full-core reserve capacity will be lost in 1987 and reload discharge capability will be lost in 1991.

Loss of full-core reserve could reduce the options available for fuel storage and possibly impact maintenance operations during a scheduled or unscheduled shutdown.

Reload discharge capability is necessary to provide fuel storage to discharge the spent fuel from each cyc,the le. Loss unitcapability of this with enough spent could prevent the unit from returning to service.

As indicated in Reference 13, NNECO is currently evaluating several options to enable Millstone Unit No. I to maintain sufficient spent fuel storage capacity to safely discharge spent fuel through the end of its' design lifetime. As a result, of the potential implications of not having sufficient spent fuel storage capacity, '

NNECO believes that continuing effort on this topic is warranted.,

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ISAP Topic No. 2.20 - RWCU System Isolation Setpoint Reduction Reference y

hN Proposed Action The reactor water clean-up (RWCU) system is important in minimizing the amount of radioactivity released to the environment, in the unlikely event of an accident, and in keeping occupational doses low by purifying the reactor coolant during normal operation. Presently, the RWCU system is set to isolate on a low reactor vessel water level signal to prevent draining the reactor and uncovering

fuel during a LOCA event. .

In Reference 5, NNECO provided the Staff with a description of a proposed project to lower the RWCU isolation setpoint to the low-low water level, in order to have the RWCU system available for decay-heat removal and reactor water cleanup following a reactor scram. Lowering the RWCU system isolation setpoint is also addressed in ISAP Topic 1.43 (water hammer) as a means of providing system availability for reactor vessel fill.

The proposed project is to evaluate the feasibility of lowering the isolation setpoint. All potential radiological consequences and adverse EEQ impacts following a break in the RWCU piping with the reduced isolation setpoint will be evaluated. A feasibility study will also be performed to evaluate alternate methods for detecting leakage from the RWCU system. ,

As the project is presently in the study phase, the benefits or risks associated with reducing the RWCU system isolation setpoint should be evaluated to determine the feasibility of such change. Thus, NNECO believes that further activity on this topic is warranted.

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l EAP Topic No. 2.21 - 480 V Load Center Replacement of Oil-Filled Breakers i

i Reference

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10 Proposed Action The over-current trip device is an integral part of the 430 V power circuit i

breakers. The function of the over-current trip device is to sense an over-1 current condition and initiate the breaker trip, if warranted, with a predetermined time delay.

In Referenca 10, NNECO provided the NRC with a description of the condition of and design of the existing over-current trip device. At the time NNECO noted that the existing devices have worn with age and are of less than desirable reliability. In order to rectify this situation, NNECO proposes to replace the

existing electro-mechanical devices with solid-state design devices which perform the same function as the old design with greater reliability and accuracy. It is expected that replacement of the existing breaker trip devices will result in a highly improved level of confidence in the reliability of the

safety-related systems served by these breakers. A side benefit resulting from the replacement of the electro-mechanical devices will be a decrease in the mai.mance and repairs required on the over-current trip devices. Based on the j above, NNECO believes that further effort on this topic is warranted.,

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, s ISAP Topic No. 2.22 - Control Rod Drive System Water Hammer Analysis Reference 3

Proposed Action ISAP Topic No. 2.22 addresses concerns related to the potential for water hammers occurring in the scram inlet lines of control rod drive (CRD) systems at BWRs. These concerns were disseminated to the licensees of BWRs through an unissued draft NRC !&E Bulletin. The draft bulletin would have required all operating BWR units to address CRD system hydraulic loads resulting from fast i actuation of scram inlet valves under the worst case loading condition. As a result, the BWR Owner's Group (BWROG) formed a CRD Scram Valve Water-Hammer Analysis Committee to address concerns on this matter.

As noted in Reference 3, NNECO does not consider thir b:"- to be a safety cdncern at Millstone Unit No. I as many scrams have occurred over the years with no evidence of CRD pipe motion or support damage. However, based on the recommendations of the BWROG Committee, NNECO is currently evaluating the

! water-hammer loads in the CRD piping generated due to scram events to assure closure of this issue. Bascd on the above, and the fact that the evaluation is in its final phase, NNECO believes further action on this topic is warranted.

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ISAP Topic No. 2.23 -Instrument, Service and Breathing Air Improvements Reference - r.

1 Proposed Action ISAP Topic No. 2.23 addresses the station air systems for Millstone Unit No.1.

The station air systems provide clean dry air for pneumatic controls. The air system also provides utility air for pneumatic tools, filter systems, tanks, and emergency breathing air. As noted in Reference 1, this topic involves an engineering review of the Millstone Unit No.1 instrument, service and breathing air systems to improve system reliability and Integrity. .

Improvements to the station air systems will increase the reliability of air-operated equipment needed during both normal and emergency operating conditions. As a result, implementation of this topic is expected to have a positive effect on plant reliability and safety. Thus NNECO believes that further activity on this topic is warranted. -

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ISAP Topic No. 2.24 - Offsite Power Systems Reference 10 i

Proposed Action

! As part of a review of the plant designs of the offsite power systems for i Millstone Unit Nos. I and 2, several areas of potential weakness and the means

for improving offsite power system reliability, capacity and availability for

! Millstone Unit No. I have been identified. As noted in Reference 10, these two items are: -

l 1. Installation of a slow speed bus transfer scheme to afford plant l

personnel both an immediate and a delayed chance of reconnecting to l the switchyard.

j 2. Installation of a generator disconnect device (circuit breaker) for the

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Implementation of the first item could be very useful on buses supplying pump

! motors by affording the plant a second opportunity to reconnect to the switchyard before transferring over to emergency power sources. In addition,

! implementation of the first item would improve plant reliability by allowing time for the pump motors to slow down to a point where a second attempt could be made to restart the pump motor from the offsite supply. ,

I Implementation of the second item would enable plant personnel to keep the i NSST-1 transformer in service by backfeeding through the GSU. In addition, since the majority of the faults associated with the main generator /GSU/NSST area are main generator mechanical faults, the reliability of keeping the NSST in i

service would be increased.

l As a result, based on the. above items, NNECO believes that further action on this topic is warranted.

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