ML13308A902

SER Re Environ Qualification of safety-related Equipment

ML13308A902
Person / Time
Site:	Haddam Neck, San Onofre, 05000000
Issue date:	12/06/1982
From:	Office of Nuclear Reactor Regulation
To:
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ML13308A897	List: ML13308A898 ML13308A901 ML13308A902
References
TAC-56906 NUDOCS 8601080556
Download: ML13308A902 (22)

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SAFETY EVALUATION REPORT BY THE OFFICE OF NUCLEAR REACTOR REGULATION FOR CONNECTICUT YANKEE ATOMIC POWER COMPANY HADDAM NECK DOCKET NO. 50-213 ENVIRONMENTAL QUALIFICATION OF SAFETY-RELATED ELECTRIC EQUIPMENT INTRODUCTION General Design Criteria 1 and 4 specify that safety-related electrical equipment in nuclear facilities must be capable of performing its safety related function under environmental conditions associated with all normal, abnormal, and accident plant operation.

In order to ensure compliance with the criteria, the NRC staff required all licensees of operating reactors to submit a re-evaluation of the qualification of safety-related electrical equipment which may be exposed to a harsh environment.

BACKGROUND On February 8, 1979, the NRC Office of Inspection and Enforcement (IE) issued to all licensees of operating plants (except those included in the systematic evaluation program (SEP)) IE Bulletin (IEB) 79-01, "Environ mental Qualification of Class IE Equipment." This Bulletin,.together with IE Circular 78-08 (issued on May 31, 1978), required the licensees to perform reviews to assess the adequacy of their environmental qualifica tion programs.

On January 14, 1980, NRC issued IE Bulletin 79-01B which included the DOR guidelines and NUREG-0588 as attachments 4 and 5, respectively.

Subsequently, on May 23, 1980, Commission Memorandum and Order CLI-80-21 was issLued and stated the DOR guidelines ari po tian 88 form the requirements that licensees must meet regarding environmental 8601080556 851223 PDR ADOCK 05000206 P

PDR

-2 qualification of safety-related electrical equipment in order to satisfy those aspects of 10 CFR 50, Appendix A, General Design Criterion (GDC) 4.

Supplements to IEB 79-01B were issued for further clarification and definition of the staff's needs. These supplements were issued on February 29, September 30, and October 24, 1980.

In addition, the staff issued orders dated August 29, 1980 (amended in September 1980)-and October 24, 1980 to all licensees. The August order required that the licensees provide a report, by November 1, 1980, docu menting the qualification of safety-related electrical equipment. The October order required the establishment of a central file location for the maintenance of all.equipment qualification records. The central file was mandated to be established by December 1, 1980. The staff subsequently issued Safety Evaluation Reports (SERs) on enviromental qualification of safety-related electrical equipment to licensees of all operating plants in mid-1981. These SERs directed licensees to "either provide documentation of the missing qualification information which demonstrates that safety-related equipment meets the DOR Guide lines or NUREG-0588 requirements or commit to a corrective action (re-qualification, replacement (etc.))."

Licensees were required to respond to NRC within 90 days of receipt of the SER. In response to the staff SER issued May 29, 1981, the licensee submitted additional information regarding the qualification of safety-related electrical equipment.

-3 EVALUATION The acceptability of the licensee's equipment environmental qualification program was reviewed for the Division of Engineering by the Franklin Research Center (FRC) as part of the NRR Technical Assistance Program in support of NRC operating reactor licensing actions. The consultant's review is documented in the report "Review of Licensees" Resolutions of Outstanding Issues from NRC Equipment Environmental Qualification Safety Evaluation Reports," which is attached.

We have reviewed the evaluation performed by our consultant contained in the enclosed Technical Evaluation Report (TER) and concur with its bases and findings. Our review has also revealed certain discrepancies in the TER which are being corrected by this SER as follows:

o Delete the third paragraph on page 1-9 of the TER.

o Delete the second paragraph on page 1-10 of the TER.

The staff has also reviewed the licensee's justification for continued operation regarding each item of safety-related electrical equipment identified by the licensee as not being capable of meeting environmental qualification requirements for the service conditions intended.

CONCLUSIONS Based on the staff's review of the enclosed Technical Evaluation Report and the licensee's justification for continued operation, the following conclusions are made regarding the qualification of safety-related elec trical equipment.

-4 Continued operation until completion of the licensee's environmental qualification program has been determined to not present undue risk to the public health and safety. Furthermore, the staff is continuing to review the licensee's environmental qualification program. If any ad ditional qualification deficiencies were identified during the course of this review, the licensee would be required to reverify the justification for continued operation. 'The staff will review this information to ensure that continued operation until completion of the licensee's environmental qualificatton program will not present undue risk to the public health and safety. In this regard, it is requested that the licensee do the following:

o Resolve any deficiencies identified in Appendix D of the FRC TER regarding justification for continued operation. If as a result of resolving these deficiencies, the previous justifi cation for continued operation is changed, provide within thirty.

(30) days of receipt of this SER the new justification for continued operation regarding each affected item.

The major qualification deficiencies that have been identified in the enclosed FRC TER (Tables 4-1, 4-2, 4-3 and 4-4) must be resolved by the licensee. Items requiring special attention by the licensee are summarized below:

o Submission of information within thirty (30) days for items in NRC categories 1B, 2A and 2B for which justification for continued operation was not previously submitted to NRC or FRC,

-5 o Incorporation of vendor maintenance/replacement schedules for specific equipment and replacement schedules for components, sub-components and materials, based on degradation, into Table 1 (surveillance and preventive maintenance) of the FRC TER (Page 4-16),

o Section 4.3.3.3 of the FRC TER identifies a concern regarding radiation inside and outside containment.

The staff has

• reviewed this concern and concludes that the licensee's methodology is acceptable.

o Section 4.3.4 of the FRC TER identifies a concern regarding temperature and pressure conditions outside containment. The staff-has reviewed this. concern and conludes that the'licensee's qualification methodology-and determination of environmental parameters for areas outside containment are acceptable.

The licensee must provide the plans for qualification or replacement of the unqualified equipment and the schedule for accomplishing its proposed correction action.

PROPRIETARY REVIEW Enclosed in the FRC Technical Evaluation Report (TER) are certain identi fied pages on which the information is claimed to be proprietary.

During the preparation of the enclosed TER, FRC used test reports and other documents supplied by the licensee that included material claimed to be proprietary by their owners and originators. NRC is now preparing to publicly release the fRC ii( and it is incumbent on the agency to seek review of all claimed proprietary material. As such, the licensee

-6 is requested to review the enclosed TER with their owner or originator and notify NRR within seven (7) days of receipt of this SER whether any portions of the identified pages still require proprietary protection.

If so, the licensee must clearly identify this information and the specific rationale and justification for the protection from public disclosure, detailed in a written response within twenty (20) days of receipt of this SER. The level of specificity necessary for

-such continued protection should be consistent with the critefia enumerated in 10 CFR 2.790(b) of the Commission's regulations.

0e UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON, D. C. 20555 March 28, 1985 Docket Nos. 50-213/245 LSO5-85-03-035 Mr. W. G. Counsil, Senior Vice President Nuclear Engineering and Operations.

Connecticut Yankee Atomic Power Company and Northeast Nuclear Energy Company Post Office Box 270 Hartford, Connecticut 06141

SUBJECT:

SCHEDULAR EXTENSION FOR EOUIPMENT QUALIFICATION Re:

Haddam Neck Plant Millstone Nuclear Power Station, Unit No. 1 This is in response to your letter of February 28, 1985, wherein you requested an extension of the deadline established in 10 CFR 50.49 for final

.environmental qualification of certain electrical equipment at both the Haddam Neck Plant and Millstone Nuclear Power Station Unit 1. Included in your request for these units are valve motor operators (14 in Haddam Neck and 28 in Millstone Unit 1) and the reactor coolant system loop temperature elements in Haddam Neck.

10 CFR 50.49 established a deadline for final environmental qualification of safety related electrical equipment by the end of the second refueling outage following March 31, 1982, or March 31, 1985, whichever is earlier.

The rule states that extensions beyond March 31, 1985 to a date no later than November 30, 1985 may be granted by the Director, Office of Nuclear Reactor Regulation, for specific pieces of equipment, if the requests are filed on a timely basis and demonstrate good cause for the extension such as procurement lead time, test complications and installation problems.

In your December 28, 1983 letter, you requested a schedular extension until March 31, 1985 from the deadline because of difficulties experienced in the procurementoof the replacement valve motor operators. The replacement valve motor operators for both Haddam Neck and Millstone Unit I are approximately three to four times larger than the existing operators. The larger size and mass would require structural modifications because of space limitations and significant reanalysis of the piping supports due to the eccentric mass effects of the valve operators. In addition, procurement of the new operators is dependent on the productivity of the vendor.

0 0

Mr. W. March 28, 1985 With regard to the loop temperature elements for Haddam Neck, your preliminary engineering analysis concluded that the existing single-element resistance temperature detectors (RTDs) should be replaced with qualified, dual-element RTDs. This conclusion anticipates certain instrumentation design changes necessary for Inadequate Core Cooling and other post-accident monitoring functions. While qualified replacement RTDs are available, you have stated that procuring RTDs with an established system reliability is a difficulty.

You have stated that an engineering evaluation to ensure that the circuit design protects against transients caused by reliability-related failures would be performed, but would require a significant amount of time.

On April 5, 1984, the staff granted a schedular extension to March 31, 1985 for the replacement of the valve motor operators in both plants and the loop temperature elements at Haddam Neck. In the Safety Evaluation granting the deferral of equipment qualification items from the 1984 outage also dated April 5, 1984) the staff concluded that if the licensees could not demonstrate adequate qualification of the subject equipment during the course of the staff program review, or otherwise demonstrate that the subject equipment is not required for safe plant shutdown and accident mitigation, the licensees would have to either replace the equipment by March 1985 or provide a basis for continued plant operation beyond that date, with a suitable justification for any further delay, and request a schedular exemption. However, the staff also acknowledges that the environmental qualification for this equipment would be evaluated in concert with other issues to provide a sound technical basis for an integrated implementation schedule for all necessary plant modifications. That evaluation was to be conducted as part of an Integrated Safety Assessment Program (ISAP), following Commission approval of a plan to implement such a program. A policy statement regarding ISAP was published on November 15, 1984 (49 FR 45112) following Commission approval and Congressional review. Subsequently, ISAP was deferred because of budgetary constraints.

In your February 28, 1985 letter you subsequently requested a schedular extension for the valve motor operators at both plants and for the loop temperature elements at Haddam Neck until November 30, 1985. You also stated that possible extensions beyond the November 30, 1985 date are likely and that as a minimum an extension to January 1986 would be necessary to reach the next scheduled refueling outage for Haddam Neck. An extension until November 30, 1985 for Millstone Unit I would permit you to reach your next refueling outage which is presently scheduled to begin in October 1985. We recognize that, while you have continued the procurement and analysis activities related to this issue, the uncertainties associated with implementation of ISAP have hampered your ability to complete plant modification designs and precluded the forum by which you would have otherwise resolved the safety significance of the specific modifications required.

9 Mr. W. March 28, 1985 The staff concludes that you have shown good faith by your attempts to procure new valve motor operators and loop temperature elements for replace ment of equipment whose qualification could not be completely established and by the replacement of one valve motor operator at Millstone Unit 1. We recognize the significant problems you have had with your contract supplier in receiving qualified valve operators for replacement at both the Haddam Neck and Millstone Unit 1 plants and that to date about 40 percent of the operators ordered have been actually received by the utilities.

Further, the staff is currently reallocating resources to support a limited ISAP effort specifically for Haddam Neck and Millstone Unit 1. It is the staff's intent to begin working on the safety significance of the unqualified motor operators and loop temperature elements which will otherwise have to be replaced by November 30,1985, as a part of this effort. We anticipate this effort will be completed by September 1985. The timely completion of our efforts by September 30, 1985 is contingent upon staff receipt of your submittal currently scheduled for June 1985.

Based on the above, we find that your request for extension was filed on a timely basis, within the scope of 10 CFR 50.49(g), and demonstrates good cause for an extension of time to complete final environmental qualification of valve motor operators (14 in Haddam Neck and 28 in Millstone Unit 1) and the loop temperature elements in Haddam Neck. Therefore, an extension is granted until Nov mber 30, 1985 for both plants for the final environmental qualification of the above electrical equipment.

In granting the extensions for Millstone Unit 1 and Haddam Neck, we are aware of the potential for future schedular extension requests beyond November 30, 1985. Millstone Unit 1 will be able to reach its next scheduled outage prior to November 30, 1985 but Haddam Neck would need an extension until January 1986 to reach its next scheduled outage. Some potential therefore exists at both Haddam Neck and Millstone Unit 1 for extension reauests beyond 1986. However, under Section 50.49(g) the Director of NRR does not have the discretion to grant extension beyond November 30, 1985 and therefore any such requests could only be granted by the Commission.

Sincerely, HarovDenton, Director Office of Nuclear Reactor Regulation cc: See next page

Mr. W. March 25, 1985 cc Gerald Garfield, Esquire Kevin McCarthy, Director Day, Berry & Howard Radiation Control Unit Counselors at Law Department of Ehvironmental City Place Protection Hartford, Connecticut 06103-3499 State Office Building Hartford, Connecticut 06106 John F. Opeka Vice President, Nuclear Operations Board of Selectmen Northeast Utilities Service Company Town Hall Post Office Box 270 Haddam, Connecticut 06103 Hartford, Connecticut 06141 Superintendent State of Connecticut Haddam Neck Plant Office of Policy and Management RFD #1 ATTN: Under Secretary Energy Post Office Box 127E Division East Hampton, Connecticut 06424 80 Washington Street Hartford, Connecticut 06106 Resident Inspector Haddam Neck Nuclear Power Station U.S. Environmental Protection Agency co U.S. NRC Region I Office East Haddam Post Office ATTN: Regional Radiation Representative East'Haddam, Connecticut 06423 JFK Federal Building Boston, Massachusetts 02203 Dr. Thomas E. Murley, Regional Administrator Nuclear Regulatory Commission, Region I 631 Park Avenue King of Prussia, Pennsylvania 19406 Northeast Nuclear Energy Company ATTN: Superintendent Millstone Plant P. 0. Box 128 Waterford, Connecticut 06358 Resident Inspector c/o U.S. NRC Millstone Plant P. 0. Box 811 Niantic, Connecticut 06357 First Selectman of the Town of Waterford Hall of Records 200 Boston Post Road Waterford, CoCnecticut 06385

OtRE G&, q UNITED STATES NUCLEAR REGULATORY COMMISSION WASHINGTON. D. C. 20555 JUL 311985 Docket Nos.: 50-213 and 50-245 Mr. John F. Opeka, Senior Vice President Nuclear Engineering and Operations Northeast Utilities Post Office Box 270 Hartford, Connecticut 06141-0270

SUBJECT:

INTEGRATED SAFETY ASSESSMENT PROGRAM In your letter dated May 17, 1985, Northeast Utilities outlined a proposal for the conduct of the Integrated Safety Assessment Program (ISAP) for the Haddam Neck and Millstone Unit I facilities. Your proposal generally described the procedures for the conduct of this effort and specifically identified (1) those projects which would be conducted independent of ISAP, including plant modifications which will be implemented during the next refueling outage for each facility and ongoing ("baseload") engineering studies, and (2) those licensing matters and NNECO/CYAPCO plant improvement projects which should be evaluated in an integrated assessment for each facility.

Your proposal also described the criteria you would use to evaluate each of the issues in the integrated assessment and subsequently prioritize the corrective action resulting from the integrated assessment.

The NRC has allocated resources for the conduct of ISAP for Haddam Neck and Millstone Unit 1. Based on the availability of the probabilistic safety analysis (PSA) results, and in consideration of the refueling outage schedules, we have scheduled completion of the integrated assessment (i.e., draft report) for Millstone Unit 1 in September 1985, and Haddam Neck in September 1986. We believe that these schedules are achievable and will ensure-efficient use of both NRC and NNECO/CYAPCO resources.

The staff has concluded that the projects for each facility listed in Enclosure 1 to this letter should be conducted independent of ISAP. These projects include (1) plant modifications and procedural changes that are ready for implementation during the next scheduled refueling outage or before the scheduled completion of ISAP and/or are significant safety improvements that should not be delayed, and (2) ongoing engineering studies and plant modification design efforts that have a well defined objective and would not likely be enhanced by an integrated assessment.

s0 Mr. John F. Opeka

- 2 to this letter identifies those projects for each facility that we believe should be evaluated in ISAP. These projects have been numbered for ease of reference and accountability. The first group of projects includes all current licensing actions and plant activities, and also includes pending and potential licensing requirements. The second group includes NNECO/CYAPCO initiatives and plant improvements. We have added to the projects you proposed:

(1) the pending licensing actions from NUREG-0748 for which the staff's review is not yet complete; (2) the NNECO/CYAPCO ongoing engineering studies ("baseload")

that we believe should be evaluated, at least in a broad sense, for their potential contribution to plant safety; and (3) those Unresolved Safety Issues and generic issues, derived from the high-priority issues in NUREG-0933, that we believe can be substantially addressed on a plant-specific basis in concert with the PSA for each facility. These projects comprise the "topics" for evaluation under ISAP. The staff considers the scope of review represented by these topics, in conjunction with the PSA evaluations and updated operating experience evaluations, to be sufficiently comprehensive, such that the results of the integrated assessment will provide effective integrated schedules and the basis for future regulatory actions.

The tables presented in Enclosure 2 identify those topics for which you identified specific deterministic or probabilistic evaluations to support the integrated assessment. A number of your proposed topics identify specific plant improve ments for the purpose of prioritization; however, it would appear that more effective or efficient alternatives may evolve from the integrated assessment by evaluating the original motivation for these plant improvements. In order to ensure that the issues to be addressed in the integrated assessment are clearly defined and understood, we request that evaluations be prepared in the following formats:

1. All of the topics for which safety analyses have already been submitted to the NRC and no further NNECO/CYAPCO analysis is considered necessary should be clearly identified. While we have identified some references of this nature in your May 17, 1985 submittal, we believe it would be useful for you to sumarize these topics in tabular form, noting the dates of your submittals and any related staff evaluations. We are currently assembling this material and will use your tabular summary to ensure that all of the pertinent documentation has been identified.

2. All of the topics for which NNECO/CYAPCO safety analysis are or will be performed should be submitted in the format specified in your May 17, 1985 letter. For Haddam Neck, we request that you provide the submittal schedules so that we may plan our review; all of these analyses should be submitted by about June 1986.

3. For all of the other topics, which consist primarily of the NNECO/CYAPCO plant improvement projects, we request that you submit a concise sumairy of each topic which identifies the fundamental concern being addressed, advantages and disadvantages of any proposed corrective action relative to the attributes described in your May 17, 7985 letter, and plant design features pertinent to the issue.

K.

s ubnm

_-d todi vidually or collectively.

John F. Opeka

- 3

4. For each of the topics that will be addressed with a specific PSA evaluation, a summary of the PSA finding should be submitted which includes:

(1) the issues being addressed, (2) any corrective actions considered, (3) the affected systems, (4) a description of the analysis performed, (5) the associated fault tree(s), (6) the analysis assumptions, and (7) the conclusion. These summaries may also be submitted individually or collectively.

The staff will review these submittals and issue safety evaluation reports which will identify the specific issues to be addressed in the integrated assessment.

The topics identified in Enclosure 2 include all pending licensing actions and ongoing engineering studies, some of which are nearly complete. Consequently, should you identify topics during the course of the review that are resolved or change in scope, please notify us.

Similarly, because 6f the ISAP review schedule for Haddam Neck relative to the plant's refueling outage schedules, it is apparent that a substantial amount of engineering and procurement for the Spring 1987 outage will have to be accomplished before the integrated assessment is complete. We encourage you to schedule the Haddam Neck topic analyses so that significant safety improve ments can be developed for implementation in the Spring 1987 outage. As you identify such implementation plans, you should notify us so the scope of the related topic reviews may be adjusted accordingly.

Because of the schedule and resource constraints, the staff's review of the PSA for each facility will be limited to an audit of your analyses. The staff and consultants assigned to this effort are currently reviewing the Millstone Unit 1 IREP results and the material presented in a meeting held on June 3, 1985, to prepare for this audit. The objectives of the staff's review will be to judge whether (1) the appropriate contributors to risk have been identified for consideration in the integrated assessment, and (2) the conclusions presented for each topic analysis are appropriate. In view of the schedule constraints for Millstone Unit 1, we request that the topic-specific analyses be submitted as soon as practical. We will organize the Haddam Neck review later in the year when the analyses for that facility are nearing completion.

With respect to the attributes you have proposed as prioritization criteria, we do not intend to review all of these criteria and the ranking procedures in detail. Rather, we will evaluate your assessment of the impacts on public health and safety for each topic and subsequently judge the overall integrated implementation schedules proposed based on those conclusions.

As a result of the staff's screening review of topics to be addressed in ISAP, we note that there are a few issues that may require exemptions or license amendments to defer action until ISAP is complete. In accordance with the required procedural practices, you should separately request the necessary exemptions or license amendments to defer these requirements. We will act on such requests promptly. Your requests for deferrals or exemptions should

John F. Opeka

-4 clearly state the justification for continued plant operation and compensatory plant design features which form the basis by which you have concluded that the issue involved may be deferred until a more effective resolution and implementation schedule can be developed from the integrated assessmeht.

For example, your May 17, 1985 submittal indicates that you intend to complete all of the inadequate core cooling instrumentation (ICI for Haddam Neck durin the 1986 refueling outage; however, you have proposed a concurrent evaluation of the need for a heated junction thermocouple system (Topic 1.13). If you decide to defer completion of ICCI, you should notify the staff promptly and include a detailed explanation of the technical basis by which you have concluded that issue can be deferred.

New issues that arise during the course of this review will be evaluated for inclusion in ISAP in the following manner:

Issues raised by the staff on a plant-specific basis will be reviewed for safety significance. If prompt action is deemed necessary, we will request such action in accordance with 10 CFR 50.54(f). Otherwise, we will forward the issue for your evaluation and note the assigned ISAP topic number. Issues raised generically, including new regulations, generic letters and IE Bulletin follow-up actions should be evaluated by NNECO/CYAPCO and, if you conclude the matter should be addressed in ISAP, a deferral should be formally requested. We will respond to such requests promptly and, if they involve exemptions, the deferral request will be evaluated in accordance with 10 CFR 50.12, as described above.

This letter concludes the staff's screening review of topics to be addressed in ISAP for Haddam Neck and Millstone Unit 1. In order for us to expeditiously proceed with this effort on Millstone 1, we request that you promptly provide the analyses discussed above. For Haddam Neck, we request that you-provide submittal schedules and identify where specific probabilistic analyses will be conducted, based on your preliminary planning. Should you have any questions concerning this matter, please contact Christopher Grimes, (301) 492-8414.

Hugh. Thom son, rector Di ion of Lice ing Of ce of Nuclear Reactor Regulation

Enclosures:

As stated cc: See next page

Mr. John F. Opeka Haddam Neck Plant Connecticut Yankee Atomic Power Company Millstone Nuclear Power Station, Northeast Nuclear Enerpy Company Unit No. 1 cc Gerald Garfield, Esauire Kevin McCarthy, Director Day, Berry & Heward Radiation Control Unit Counselors at Law

[epartment of EnyironnEntal City Place Pretection Hartford, Connecticut 06103-3499 State Office Buildingq Hartford, Connecticut 06103 Edward J. Mroczka Vice President, Nuclear Operations Board of Selectmen Worthepst Utilities Service Company Town Hall Post Office Box 270 Haddam, Connecticut 06103 Hartfore, Connecticut 06141 Superintendent State of Connecticut Haddam Neck Plant 004ice of Policy and Manaoement RFD #1 ATTN:

Under Secretary Energy Post Office Box 127E Division East Hampton, Connecticut 06424 an Wastington Street Hartford, Connecticut 06106 Resident Inspector Hadeari Neck Plant Regional Administrator co U.S. NPC Nuclear Regulatory Commission, Pepion I East Haddam Post Office 631 Park Pvenue East Haddam, Connecticut 06423 King of Prussia, Pennsylvania 19406 Northeast Nuclear Energy Company ATTN:

Superintendent Millstone Nuclear Power Station P. 0. Pox 128 Waterford..Connecticut 06358 Resident Inspector C/o U.S. WR.C Villstone Nuclear Power Station P. 0. Box 811 iantic, Connecticut 06387 First Selectman of the Towo of Waterford Hall of Records 200 Boston Post Pepe WaterforeK Connecticut D63Pr

CLOSURE 1 COMPLETION OR ONGOING PROJECTS INDEPENDENT OF ISAP Haddam Neck 1

Emergency Diesel Generator Trip and Lockout 2

Process Computer UPS Installation 3

Inadequate Core Cooling Instrumentation 4

EQ of Electrical Equipment 5

RCS Loop RTD Replacements 6

Seismic Qualification of Safety Related Piping 7

Plant Paging System Upgrade 8

MOV Thermal Overload Modifications 9

Appendix R Modifications 10 Refuel Cavity Drain Piping Seismic Upgrade 11 Reactor Cavity Seal Ring Modifications 12 Replace Main Control Board Cat. 1E Relays 13 Feedwater Heater Modification 14 Steam Generator Wet Layup Recirculation 15 Review of Existing Control Air Systems 16 Doc. and Coord. of Protective Relay and Breaker 17 Replace Main Generator Neutral Grounding Trans.

18 Rev. and Upgrade Eval. for Diesel Air Start System 19 Waterbox 'A" Retubing 20 Steam Generator Nanway Cover Handling Device 21 Limitorque Motor Operated Valves-Lubrication 22 Elevated Well Water Temperature 23 RCCA Changeout Tool 24 Emergency Plan Rad. Assessment Equipment 25 Microwave System 26 Conversion to Standard Technical Specifications 27 Plant Design Change Task Group Support 28*

Plant Training Simulator 29*

Probabilistic Safety Study 30*

Unit Availability Model 31*

Computer Analysis for Generation of Heatup/Cooldown Curves

ENC URE 1 (continued)

Millstone I I

Seismic Qualification of Safety Related Piping 2

Primary Containment Leak Rate Monitoring 3

Undervoltage Protection-Emergency Bus 4

Replacement of Motor Operated Valves 5

Floodgate Mods. and Installation of Scuppers 6

Refueling Cavity Seal Evaluation 7

Records Vault Temperature and Humidity Control 8

Process Computer UPS Installation 9

MS Relief Valve Vacuum Breaker Load Qualification 10 Station Battery 'A" Replacement II Rev. and.Upgrade Eval. for Diesel Air Start System 12 Gas Turbine Generator Elec. Equipment Protection 13 Auxiliary Equipment System Oscillograph 14 Supervisory and Events Recorder Systems 15 Replace Main Generator Neutral Grounding Trans.

16 Limitorque Motor Operated Valves:Lubrication 17 460 Volt Motor Soft Start Capability 18 IGSCC Countermeasures 19 Circulating Water Piping Thrust Block Repairs 20 Gas Turbine Generator Battery Replacement 21 Solid Radwaste Building Ventilation and Roof Mods 22 Spent Fuel Pool Cleanup 23 Voltage Regulator, Instrument Trans Replacements 24 House Heating Boiler Stack 25 Procurement of Nuclear Grade Material/Service Sensitive Lines 26*

Plant Training Simulator 27*

Probabilistic Safety Study 28*

Unit Availability Model 29*

Roof Replacement 30 Emergency Gas Turbine Generator Vibration Switches Onqoing activities that represent a level of effort that must be maintained.

ENCLOSURE 2 Haddam Neck ISAP Issues Title - NRC Issues to be Included in ISAP Source Cross Refernace Action Letters Sus PSA Notes III---

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1.01 Switchgear Room Cooling Rodifications SEP 11-5 TIA 83-89 04/05/841 y

1.02 High/Lou Pressure Valve Interlocks SEP V-11.A TIA 83-89 04/05/34D T

1.03 Containment Penetration Evaluations SEP VI-4 50935;0660:I1.E.4.3 Y

1.04 Seismic Qualification of Safety Related Piping SEP 111-6 51937,51939;USI A-40 041051840 3

1.05 Seisaic Structural Rodifications S 111-6 51937,51939;US1 A-40 04105/841 1.06 lind and Tornado Loadings/Torado Rissiles SEP 111-21-4.A 51938,51939 y

1.07 Vital Bus Feed Realignment Rodifications SEP VI-7.C.1 51934;TIA 83-89 1.08 Seismic Hodifications to Reactor Coolant System SEP 111-6 51937,51939;USI A-40 04105/84D 1.09 Design Codes, Design Criteria, Load Combinations SEP 111-7.8 51939 Y

1.10 Torque Seitch Nodifications SEP 111-10.A y

1.11 PAD Ventilation System Kodifications SEP I1-5 T

1.12 Control Room Habitability 0737:11.1.3.4 56320 04/05/B4D Y

(i)c 1.13 Inadequate Core Cooling Instrumentation 0737:11.F.2 45138 V

4 1.14 Appendix R Rodifications 50.41 1.15 FDSA Update 50.71 04/11185E 1.16 Anticipated Transients lithout Scram ATUS USI A-9 1.17 Replacement of Rotor Operated Valves 50.49 42521,52074;USI A-24 04105/84D;03128185E 1,5 1.18 RC Seal Cooling Rodifications -

0737:11.K.3.25 SI-23 1.19 Control Room Design Ryite 0737:Suppl 1 56128,51165 1.20 Safety Parameter Display System 0737:Suppl 1 51245 1.21 RS 1.97 Instruaentation 0737suppl I 51095,45949,etc t 3 1\\L c,

1.22 Emergency Response Facilities Instrumentation 0737:Suppl I 51095 1.23 Post-Accidient Hydrogen Ronitor (RS 1.97) 0737:Suppl I 47745,49352;USI A-48 Y

TS Surveillance for Hydraulic Snubbers HPA 1-17 08465 2

TS Surveillance for Rechanical Aubbers PA 3-22 08469 2

C

~ RV and SV Testing HPA F-4 44585 1.27 Compliance with 50.46 (ECCS) 0737tiI.K.3.31 48169;HPA F-51 1428 RCP Trip HPA -1 49667 1.29 Flooding Evaluation SEP Ill-3.C 51932 1.30 RS Isolation SEP 51933 I.31 Pipe breaks SEP 51936 1.32 Item 2.1-Equipsent Classification/Vendor later.

Sales ATUS 52B43 1.33 Items 3.1.11.2-Post Raintenance Testing Sales ATES 52924 3.34 Item 3.1.3-Post Raintenance Testing TS Changes Sales ATUS 53005 1.35 Reactor Trip System Reliability-Vendor Sales ATIS 53073 1.36 Item 2.2-Equip. Classification/Vendor Interface Sales ATIS 53677 1.37 Items 3.2.1.2-Post Rainton. Testing Procedures Sales ATUS 53760 1.38 Item 3.2.3-Post Raintenance Testing Changes to TS Sales ATs 53843 1.39 Ithe 4.2.3&.4-Prey. Rainten. Proc. for Rs Trip Salem ATiS 53913 1.40 Items 4.5.21.3-Ri Trip System Factional Testing Sales ATS 53987 1.41 Item 4.5.1-Reactor Systes Function Testing Salem ATiS 54070 1.42 ICS Vents TS 0737:

54393 1.43 TS from SL 83-36 & 13-37 KPA 8-83 54538 1.44 Diesel Senerator Reliability HPA 1-19 55964 1.45 ISI Update to 1980 Code 56902 3.46 IST for Diesel Benerator Auxiliaries 56917 1.47 Reliability Engineering URES-0933 0660:1I.C.2 1.48 Seismic Qualification of Equipsent NULRE6-0933 USI A-46 A

19 Steas enerator Tube Integrity NURE-0933 USI A-3 fracture Toughness of Supports NURE6-0933 USI A-12

ENCLOSURE 2 Naddas Neck ISAP Issues Title -

RC Issues to be Included in ISAP Source Cross Refernece Action Letters Sum PSA Notes I

4 I

4

.4

+-+

1.51 Systeer Interactions MRE6-0933 USI A-17 1.52 Pressure Transient Protection NURES-0933 US! A-26 1.53 Containment Emergency Sump Performance NURE6-0933 US A-43 1.54 Safety Implications of Control Systems NURES-0933 USI A-47 1.55 Radiation Protection Plans RES-0933 06O:II..3.1 1.56 Bolting Degradation NUREB-0933 61-29 1.57 Flooding of Safety Equipment by backflow NURE6-0933 61-77 1.58 Steas linding of Ausiliary Feed Pumps HUREB-0933 61-93 ISAP I Title N iU Issues to be Included in ISAP Source Cross Refernece Action Letters Sum PSA Notes 4---

1I 1-I I +-+ 4-4 4-4 2.01 Secondary Side Chemistry Ronitoring N.

U HPA A-3 L 85-02 y

2.02 MNST Dxygen Reduction NUP PA A-3 GL 85-02 Y

2.03 Additional Atospheric Steas bump lU 2.04 Rodernize Reactor Protection I Control Systems NU 2.05 Process Computer Replacement NU ISAP l.20;51245 2.06 Eval. of RCS Loop Iso. Valves to Ritigate SSTR KU USI A-3,ISAP 2.01,2,3 2.07 Auxiliary Pressurizer Spray Nozzle 1U

,A -

Loss of C Power Nil USI A-44;TAP A-30,3-56 VCP Vibration Bonitoring System Upgrade NU L Administration Building Upgrade NU 311 Rain Steam System Evaluation NU 2.12 Turbine-Senrator Trip Logic MU US! A-44 AWtes:

3. An seeption or license amendment say be required.

2. Nay be resolved by implementation of the Standard Technical Specifications.

3.

To date 180 of the approximately 450 pipe racks and supports have been bodified.The licensee intends to modify 8 pipe racks and 90 pipe supports inside containment and 31 pipe supports in the primary auxiliary building during then 1986 refueling outage. The remaining work is to be included in ISAP.

4. The licensee intends to implement all of the ICCI modifications during then 1986 refueling outage unless a more effective alternative for the beated-junction thermocouple can be developed from the PSA.

5. The licensee plans to complete all of the envoronmental qualification of electrical equipment according to the 10FR50.49 schedule requirements, except for the motor operators for the 14 safety related ROVs, Before projects, such as the seismic analyses, need to be evaluated.

Hillone I ISAP Issues Title - NRC Issues to be included in ISAP Source Cross Refernece Action Letters Sue PSA Motes 1.0l Sas Turbine Senerator Start Logic Nodifications SEP VIII-2 US! A-44;KPA D-19 04/05/84D Y A 1.02 Tornado Rissile Protection SEP 111-4.A 04/05184D Y A 1.03 Containment Isolation-Appendix A Modifications SEP VI-4 1.04 RMCU Systes Pressure Interlock SEP V-11.A 49380 04/05/B40 Y A 1.05 Ventilation System hodifications SEP 11-5 Y A 1.06 Seismic Rualification of Safety Related Piping

]ED 79-02&-14 49379;IR 05-04 04/05/84D Y 3 2

1.01 Control Room Design Review 0737:Suppl 1 51176,56130 1.08 Safety Paraseter Display System 0737:Suppl I 51256;ISAP 2.03 1.09 RS 1.97 Instrusentation 0737:Suppl I 51106;UST A-9 Y

1.10 Emergency Response Facilities Instrumentation 0737:Suppl'1 51106;USI A-9 1.11 Post Accident Hydrogen Monitor 0737:Suppl I 47754;USI A-48 Y

1.12 Control Room Habitability 0737:111.D.3.4 56319 04/05/543 1.13 SWR Vessel Mater Level Instrumentation 0M:1I.F.2 03/26/B5 1.14 Appendix J Hoifications IOCFR50 1.15 FSAR Update 50.71 04111/85E;03/2S/BSE 1.16 Appendix R 50.49 4864 I

.1 IPiH2 lackfend A

.2 odify CRD Puaps

.3 Alternative Cooling for Shutdown Cooling A

.4 Power Cold Shutdown Equipment 3

2.17 Replacement of Motor Operated Valves 50.49 42523;USI A-24 04/05184D Y a 1,3 3.18 ATIS 50.62 56840 A

1.19 Integrated Structural Ananysis SEP 49376 1.70 N0V Interlocks SEP II1-I0.A 49390 Fault Transfers SEP VI-7.C.1 49384 A

Electrical Isolation SEP VII-1.A 49385 1.23 Srid Separation Procedures SEP VIII-;A 49396 1.24 Emergency Power SEP YIII-2 4937 1.25 Degradie Srid Voltage Procedures 51702 L26 Ites 2.-Equioaent Classification/Vendor Inter.

Sale AIS 52854 1..7 Itess 3.1.11.2-Pcst maintenance Testing Salem ATMS 52935 1.28 Item 3..3-Post Maintenance Testing TS Changes Salem ATMS 53016 1.29 Response-to 6. 81-34 NPA 3-65 53445;ISAP 2.23

.3.

Item 1."-Post Trip Review Data and Information Sale AIS 53606 1.31 Ites 2.2-EcdipsEnt Classification/Vendor later.

Salem ATMS 53689 1.32 Items 3.2.IL2-Past Maint. Testing Procedures Salem AIMS 5 2 1.33 Ites 3.2.3-Pcst Maintenance Testing TS Changes Salem ATMS 538 1.34 Items 4.5.21.!-Reactor Trip System Testing Sala ATMS 53999 1.15 Item 4.5.1-Reactor System Functional Testing Salem ATIS 5408?

1.36 TS Covered by SL 83-36 HPA 8-83 54545 1.37 TS Affected by 50.72 and.73 (L 83-43)

NPA A-18 55722 1.38 Expand CA List UI

-0933 0660:1.F.1 1.39 Radiation PrctEtion Plans HURE6-0933 0660:111.1.3.1 1.40 Saltin; Degradation or Failure NURES-0933 61-29 1.41 Floodir; of Ecapartaents by Backflov NIRE6-0933 61-77 1.42 MSI Leakage t:ntrol Systems NUREG-0933 TAP C-OS 1.43 ater Haser NURE-933 US! A-1 1.44 AsysEtric Bloudown Loads on Reactor Systems NURES-0933 US] A-2 1.4! Systes Interat:on MIRE-09n US! A-!7 DEttr9:%A~t? of SSV PooE Dynamic Loads lURES-0933 US] A-9 Ezri4zin Sare F&:tc Sup Performance KJRB-0933 U51 A-4C S.4938

Hillstone I ISAP Issues lAt I Title - NU Issues to be included in ISAP Source Cross Refernece Action Letters Sul PSA Notes 1p01 Remotely Operated Valve 1-LP-50AL NU

.02 Drywell Humidity Instrumentation NU 2.03 Process Computer Replacement KU ISAP 1.08 2.04 High Steas Flo Setpoint Increase NU A

2.05 Hydrogen Mater Chemistry Study NU 2.06 Condenser Retube KU 2.07 Sodius Hypochlorite Systes KU A

2.3 Extraction Steam Piping NU 2.09 Upgrading of Piping and Instrumentation Diagrams KU 2.10 Drywell Ventilation System MU 2.11 Stud Tensioners KU 2.12 Reactor Vessel Head Stand Relocation NU 2.13 Turbine later induction Rodifications Nu 2.14 Evaluation and laplementation of KURES-0577 NU 2.15 Torque Switch Evaluation for ROVs KU 2.16 Reactor Protection Trip System NU 2.17 4.169V, 480V & 125Vdc Plant Distribution Prot.

KU US! A-44 2.18 Spent Fel Pool Storage Racks/Transporation Cask WU 2.19 DC System Review U

US! A-44;7AP A-30 2.20 RECU System Isolation Setpoint Reduction KU 2.21 490V Load Center Repl. of Dil Filled Breaker NU 2.2 Control Rnd Drive System later Hasser Analysis KU Y

2.23 Instrument, Service and Ireathing Air laprva KU Y

2.24 Offsite Power Systems KU Y

Drywell Tesperature Ronitoring System Upgrade KU iReliability Equipment KU Spare Recirculation Pump Rotor KU 2.28 Long Tere Cooling Study KU US! A-45;PSS 2.29 FlCI Assessment Study KU US! A-I5 2.30 MSIV Closure Test Frequency KU A

A 2.31 LPCI Lobe Oil Cooler Test Frequency KUA

.An uieeption or license amendment may be required.

. o date 700 of the approximately 1074 identified modifications have been made. The licensee intends to modify 109 pipe supports before and during the 1985 refueling outage. The remaining work is to be included in ISAP.

.The licensee plans to complete all of the envoronmiental qualification of ele:trical eouipsent according to the IOCFR5O.49 schedule requirements, except for the mtor operators for the 29 safety related ROVs. before these motor operators can be installed, their effect on other ongoing projects, such as the seismic analyses, need to be evaluated.

• thntitaive risk calculation.

r i

ublit h,,raioloiraloub~r nA

John F. Opeka

- 4 clearly state the Justification for continued plant operation and compensatory plant design features which form the basis by which you have concluded that the issue involved my be deferred until a more effective resolution and implementation schedule can be developed from the integrated assessment.

• or example, your Nay 17, 1985 submittal indicates that you intend to complete all of the inadequate core cooling instrumentation (ICCs) for Haddam Neck during the 1986 refueling outage; however, you have proposed a concurrent evaluation of the need for a heated Junction thermocouple system (Topic 1.13). If you decide to defer completion of ICC, you should notify the staff promptly and include a detailed explanation of the technical basis by which you have concluded that issue can be deferred.

New issues that arise during the course of this review will be evaluated for inclusion in ISAP in the following manner: Issues raised by the staff on a plant-specific basis will be reviewed for safety sign'ificance. If prompt action is deemed necessaryD we will request such action in accordance with 10 CFR 50.54(f). Otherwise, we will forward the issue for your evaluation and note the assigned ISAP topic number. Issues raised generically, including new regulations, generic letters and IE Bulletin follow-up actions should be evaluated by NNECO/CYAPCO and, if you conclude the matter should be addressed in ISAP, a deferral should be formally requested. We will respond to such requests promptly and, if they involve exemptions, the deferral request will be evaluated in accordance with 10 CFR 50.12, as described above.

(3~,

This letter concludes the staff's screening review of topics to be addressed in ISAP for Haddam Neck and Millstone Unit 1. In order for us to expeditiously proceed with this effort on Millstone 1, we request that you promptly provide the analyses discussed above. For Haddam Neck, we request that you provide submittal schedules and identify where specific probabilistic analyses will be conducted, based on your preliminary planning. Should you have any questions concerning this matter, please contact Christopher Grimes, (301) 492-8414.

Hugh L. Thompson, Jr., Director Division of Licensing Office of Nuclear Reactor Regulation

Enclosures:

DISTRIBUTION TSpeis As stated Gentra1 tile" JShea MBoyle RThompson/FMiraglia JZwol inski FRowsoie cc: See next page CGriines AThadani FAks tul ewi cz

• PREVIOUS CONCURRENCE SEE DATE SEPB:DL*

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MBoyle:mn CGrimes FAkstulewicz JShea JZwolinski 7/01/85 7/01/85 7/08/85 7/03/85 7/03/85 AD:SA:DL*

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DCrutchfield AThadani FRnwsome TSpeis EMcCabe 7/19/85 7/19/857/1/85 D: DL Who~b