ML19301F086
| ML19301F086 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/03/1986 |
| From: | Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML19301F081 | List: |
| References | |
| NUDOCS 8603120304 | |
| Download: ML19301F086 (64) | |
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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION PELATED TO THE DETAILED SCOPE FOR THE INTEGRATED ASSESSMENT OF ISAP TOPICS FOR MILLSTONE NUCLEAR POWER STATION, UNIT NO.]
DOCKET NO. 50-245 1.0 Introduction In a letter dated May 17, 1985, Northeast Utilities (the licensee) outlined a proposal for the conduct of an Integrated Safety Assessment Program (ISAP) for Millstone Unit No. 1.
The proposal generally described the procedures for the performance 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 program period, and (2) those licensing matters and plant improvement projects which should be evaluated in an ISAP integrated assessment.
The proposal 6150 described criteria that would be used to evaluate each of the topics to be considered in an integrt.ted assessment and to subsequently prioritize the resulting recommended corrective actions.
The staff, in a letter dated July 31, 1985, concluded that the projects identified by the licensee to be performed independent of ISAP for Millstone D Unit No. I were appropriate. Also, the staff concurred in the general g
content of the topics to be evaluated under ISAP.
These topics include pending licensing actions from NUREG-0748, " Operating Reactor Licensing Actions Summary," for which the staff's review is not yet complete, and licensee initiated plant improvement projects.
In addition, the staff identified a number of safety / generic issues, derived from the high-priority issues in NUREG-0933, "A Prioritization of Generic Safety Issues," which the staff felt that, although not yet issued for licensee implementation, could be substantially addressed on a plant-specific basis in concert with the Millstene Unit No. 1 Probabilistic Safety Study (PSS). These additional issues could, therefore, be resolved in whole or in part in the integrated assessment.
N h DOCM0304 860303 05000245 P
2.0 Discussion In the staff's July 31, 1985 letter, the staff requested that safety analyses be submitted for each of the ISAP topics.
Each topic safety analysis was to consist of a scope of review with a concise summary of each topic which identified the fundamental concern being addressed.
Also, for each topic that could be addressed with a specific PSS evaluation, a PSS topic surcary was to be submitted.
By letters dated August 13, 1985, through November 25, 1985, the licensee provided the scope and the safety analyses for the 80 Millstone Unit No. 1 ISAP topics, including PSS summaries for 21 of the topics.
In addition, the licensee provided supplemental information for several of the topics in submittals dated through February 4, 1986.
The staff reviewed each of the topic submittals to determine if the content of the jetailed topic scope was appropriate and to determine if any of the topics were resolved by the information submitted. The staff also reviewed the Millstone Unit O. 1 PSS and the plant operating history to identify any possible additional topics that would warrant consideration in the integrated assessment. The conclusions reached by the staff in their review of the topics scope and summaries, the PSS and the plant operating experience can be categorizeo as one of the following:
(1) The staff concurs that the issues identified with re h t to dhe topic scope are appropriate for consideration in the integrated assessment.
(2) The staff has redefined the scope of the topic to reflect related issues; (3) The staff has develpped a new topic for consideration in the integrated assessment; or (4) The scope of the topic has been resolved and, therefore, does not need to be evaluated in the integrated assessment.
Table 1 is a listing of each ISAP topic and its category as determined in this review.
3.0 Evaluation Presented below are summaries of each of the staff's topic evaluations and conclusions.
The first section contains the staff's evaluations of the 80 topics identified in our July 31, 1985 letter. The second section details three additional tcpics that the staff considers appropriate for consideration in the integrated assessment.
3.1 Evaluation of July 31, 1985 ISAP Topics 3.1.1 Topic 1.01, Gas Turbine Generator Start Logic Modifications As part of the Millstone Unit No. 1 SEP review (Topic VIII-2), the diesel generotor and gas turbine generator (GTG) were reviewed against the criteria of General Design Criterien (GDC) 17, as iDplemented by Standard Review Plan (SRP) Section 8.3.1 and Branch Technical Position (BTP)
ICSB 17, with regard to trips under accident conditiens and maintenance programs. As documer,ted in the Integrated Plant Safety Assessment ?eport (NUREG-0824, IPSAR) Supplement, all of the issues raised in SEP Topic VIII-2 have been resolved, with the exception of bypass under accident conditions of the light-off. speed and generator excitation trips, the high lube oil temperature trip and five GTG output breaker protective trips.
% 2-g In a letter dated August 13, 1985, the licensee proposed to evaluate the safety implications of the outstanding GTG trips, and the need to implement necessary GTG start logic modifications, in the integrated assessment and to prioritize any necessary modifications.
The staff concurs with the scope of this topic.
3.1.2 Topic 1.02, Tornado Missile Protection As part of the Millstone Unit No. 1 SEP review (Topic III-4.A), the plant was reviewed agairit the criteria of GDC 2, as implemented by SRP Sections 3.3.1 and 3.3.2 and Regulatory Guides 1.76 and 1.117, with regard to its ability to withstand the effects of tornado generated missiles. The IPSAR identified those structures and components vulnerable to tornado missiles s
and it was the staff's position that the licensee was to provide sufficient protection to ensure that the plant could be brought to the hot shutdown conditicn.
I r. a letter dated December 2, 1983, the licensee addressed the staff's position.
To resolve this open issue, the licensee proposed to rely on the isolation condenser as the prir,ary means for reaching hot shutccwn.
Also, the licensee cemitted to provide a ccnnection to the underground city water system and a missile protected pump in the reactor building to suppl., additional cooling water to the isolation condenser.
In a safety evaluation dated November 25, 1985 and as documented in the IPSAR Supplement, the staff concluded that the licensee's proposal will provide adequate protection against torrado missiles.
However, because of the importance of the Condensate Storage Tank (CST) and fire-water tanks as cooling water sources, the staff recommended that the capability of the tanks anchor bolts to provide substantial resistance to wird-load induced failure be confirmed.
In a letter dated February 4, 1985, NNECO provided the additional information recuested by the staff.
This information is currently under review by staff and will be addressed in the staff's review of the Millstone Unit No. 1 integrated assessment.
In a letter dated October 16, 1985, the licensee proposed to prioritize the recorrgded modifications in ISAP. The staff concurs with the sccpe of this toph.
q 3.1.3 Topic 1.03, Containment Isolatien - Appendix A Modifications As part of the Millstone Unit No. 1 SEP review (Topic VI-4), the isolation provi:; ions for lines penetrating the prinary containment were reviewed against the criteria of GDC 54 through 57, as implemented by SRP Section 6.2.4 and Regulatory Guides 1.11 and 1.141. As documented in the IPSAR Supplement, all of the issues raised in SEP Topic VI-4 have been resolved, with the exception of the adequacy of the isolation provisions for penetration X-204, the cooling water return lines that branch off between takeoffs to the containment spray pumps.
In a letter dated October 16, 1985, the licensee proposed to evaluate the potential modifications to penetration X-204 in relation to other NRC requirements (Appendix J Modifications: ISAP Topic 1.14) in the integrated assessment. The staff concurs with the scope of this topic.
3.1.4 Taoic 1.04, RWCU System Pressure Interlock As part of the Millstone Unit No. 1 SEP review (Topic V-11.A), the redundancy and diversity of pressure interlocks for motor-operated valves which isolate the high pressure systems from low pressure systems were evaluated against the criteria of 10 CFR 50,44a, as implemented by SRP Section 7.6 and BTP ICSB 3.
As documented in the IPSAR, it was recommer M that the licensee either demonstrate the adequacy of the reactor water cleanup system (RWCU) relief valve or install an independent pressure interlock for actuation of RWCU systen isolation on high pressure.
The licensee proposed to install an independent pressure interlock for the inboard suction isolation valve.
In a letter dated August 13, 1985, the licensee proposed to evaluate the adequacy of the RWCU relief valve and the need for an independent pressure interlock with the priority of the evaluation and rrodification to be determined in the integrated assessment. This type of evaluation was an alternative to system modifications available to Millstone Unit No. I and D-g other plants that particioated in SEP.
The results of the system evaluation in concert with the PSS findings will be used to determine what, if any, plant nodifications are appropriate.
The staff concurs with the scope of this topic.
3.1.5 Tcoic 1.05, Ventilation Systems As part of the Millstone Unit No. 1 SEP review (Topic IX-5), the ventila-tion systems were reviewed against the criteria of GDC 4, 60 and 61, as irrplemented by SRP Sections 9.4.1 through 9.4.5, with regard to the capability to provide a safe environment for plant personnel and safety features.
As documented in the IPSAR Supplement, all of the issues raised in the SEP review were resolved except for providing emergency power to the FWCI spect coolers and to the intake structure ventilation system.
The licensee co citted to make the necessary modifications to the ventilation systems to resolve this issue.
In a letter dated August 13, 1985, the licensee proposed to evaluate the conservatisms in the ventilation system's design limits and use risk insights and best estimate success criteria from the PSS to determine the need to implement the modifications, and, if necessary, to determine a priority for implementation.
The staff concurs with the scope of this topic.
3.1.6 Topic 1.06, Seismic Oualificatio1 of Safety Related Piping IE Bulletin 79-02 requires licensees to verity tne adecuacy of pipe support base plates and expansion anchor bolts utilized in supporting safety-related piping.
IE Bulletin 79-14 requires licensees to verify that seismic analyses perfomed cn safety-related piping systems apply to the actual configuration of the safety-related piping systems.
The review was conducted to the criteria in the bulletins, Regulatory Guide 1.29 and the Millstone Unit No. I FSAR.
IE Inspection Report 50/245/85-04, dated April 18, 1985, reviewed the licensee's completeness of response to IE Bulletins 79-02 and 79-14. The scope of the inspection included a review of engineering design and quality assurance documentation to verify that the licensee had met bulletin requirements. Also, a walkdown inspection of the plant was conducted to verify repairs relating to the bulletins.
No violations were reported.
In a letter dated Octcber 11, 1985, NNECO stated that approximately 1100 modifications were identified.
NNECO planned to implement about 100 pipe support modifications during the November 1985 refueling outage. To date approximately 270 modifications remain.
The scope of work in the 1985 outage reflected the results of NNECO's August 23, 1985 importance enalyses.
In those analyses, the high priority modifications were incorporated into the outage work scope. With respect to the rem 6ining modifications, NNECO's position is that since analysis has shown that the remaining modifications will not provide a significant improvement in safety, a priority for implementation should be determined in ISi.
The staff concurs with the scope of this topic.
3.1.7 Tooic 1.07, control Room Design Review As part of the generic requirements of the TMI Action picn, identified in NUREG-0737 Supplement I and as implemented by Generic Letter 82-33, all licensees and applican.s for operating licenses are to conduct a detailed control room design review. This review is intended to improve the ability of operators to prevent accidents ce cope with accidents by identifying any modifications of control room configurations inat would contribute to a significant risk reduction or improvement in safety.
By letters dated April 15 August 11, November 28 and December 20, 1983 and April 9, 19E4, NNEC0 provided the schedule and plans for completing the requirements detailed in Generic Letter 82-33. The schedule for the control room design review called for a progran plan and a summary report, including a proposed schedule for implementation, to be submitted to the NRC by March 2, 1987. The staff concurred with this schedule and on June 12, 1984 issued an order confirming NNECO's conmitment.
In a letter dated October 16, 1985, NNEC0 proposed that the schedule for the control room design review be evaluated and prioritized in ISAP.
The staff concurs with the sccpe of this top'.
3.1.8 Topic 1.08, Safety Parameter Disolay System As part of the generic requirements of the TMI Action plan, identified in NUREG-0737 Supplement I and as implemented by Generic Letter 82-33, all licensees are to provide a safety parameter display system.
This system is to display to the operators a minimum set of parameters defining the safety status of the plant and be capable of displaying a full range of important plant parameters and data trends on demand, and be capable of indicating when process limits are being approached or exceeded.
By letters dated April 15, August 11, November 28 and December 20, 1983 and April 9, 1984, NNECO provided the schedule and plans for completing the requirements detailed in Generic Letter 82-33.
The schedule for meeting the safety parameter display system requirements calls for the submittal of a safety analysis and pian by April 9, 1987.
The staff concurred with this schedule and on June 12, 1984 issued an order confirming NNECO's commitment.
In a letter dated October 16, 1985, NNEC0 proposed that the schedule for the safety parameter display system be evaluated and prioritized in ISAP.
The staff concurs with the scope of this topic.
3.1.9 Topic 1.09, Regulatory Guide 1.97 Instrumentation As part of the generic requirements of the TMI Action Plan, identified in NUREG-0737 Supplement I and as implemented by Regulatory Guide 1.97, Rev. 2, licensees and applicants for operating licenses are required to provide instrumentation to monitor variables and systems over their anticipated ranges for normal operation, for anticipated operational occurrences and for accident conditions to assure safety. Regulatory Guide 1.97, Rev. 2, identifies parameters to be monitored, ranks the parameters according to the importance of their function, and provides reconmended ranges for the parameters.
In a le' -
cc March 11, 1985, the staff issued an interim report to NNECr.
g.
the results of their review of the NNEC0 submittals on Regula E. !e 1.97 for Millstone Unit No.1.
In that report, the staff i.i
'...ed that the information submitted by NNEC0 was acceptcble with the exception of the following items which require additicnal justification.
3)
Environmental qualification - there are five Category 2 variables for which environmental qualification should be addressed in accordance with 10 CFR 50.49.
2)
Power source - there are four Category 1 variables for which the licensee should provide independent Class IE power sources for each of the redundant channels.
3)
Coolant level in reactor - the licensee should justify the ncn-redundant instrumentation from +60 in. to the centerline of the main steamline.
4)
Drywell sump level - the licensee should provide additional justification for deviating from the recommendations for this variable.
5)
Drywell drains sump level - the licensee should provide additional justification for deviating from the recommendations for this variable.
6)
Containment drywell hydrogen concentration - the licensee should show that the post-accident sampling system is a viable alternative means of obtair" 'g this information.
7)
Containment drywell oxygen concentration - the licensee should show that the post-accident sampling system is a viable alternative means of obtaining this information.
8)
Radiation exposure rate - the licensee should identify the ranges and locations 'hese instruments.
9)
Main feedwater flow - environmental qualification should be addressed in accordance with 10 CFR 50.49.
- 10) Suppression chamber spray flow - the licensee should provide additional justification in support of their deviation.
- 11) Drywell spray flow - The licensee should provide additional justification in support of their deviation.
12)
Standby liquid control system flow - The licensee should provide additional justification for the alternative instrumentation.
13)
Residual heat removal system flow - the licensee should provide additional justification for the alternative instrumentation.
- 14) Cooling water flow to ESF system components - the licensee should show that the instrumentation for pump discharge pressure meets the Category 2 requirements.
15)
High radioactivity liquid tank level - the licensee should justify the use of high and low level alarms in lieu of continuous level indication.
- 16) Reactor building or secundary containment area radiation - the licensee should provide additional justification for not implementing this variable.
By letter dated October 25, 1985, NNECO identified the sccpe of ISAP Topic 1.09 as addressing the conformance of Millstone Unit No. I with the recommendations of Reculatory Guide 1.97, Rev. 2.
Also, in that letter, NNECO addressed the 16 items, listed above, identified in the staff's interim report.
The licensee has provided additional information as requested, identified some items to be specifically evaluated in the ISAP integrated assessment, or stated that evaluations of items are still ongoing. The additional information submitted by NNECO is currently under staff review. Therefore, it is the staff's position that all 16 items should be evaluated in the integrated assessment.
3.1.10 Topic 1.10, Emergency Response Facilities Instrumentation As part of the generic requirements of the TMI Action plan, identified in NUREG-0737 Supplement 1 and as implemented b" Generic Letter 82-33 and Regulatory Guide 1.97, Rev. 2, all licensees and applicants for operating licenses are to provide an Emergency Response Facility (ERF) with appropriate instrumentation. Millstone Unit No. I has an approved ERF.
By letter dated October 24, 1985, NNECO stated tha' the final list of variables to be displayed is currently being developed and will be provided at a lati.r date.
It is the staff's position that this information be determined in the integrated assessment together with the schedules for ISAP Topic 1.09.
3.1.11 Topic 1.11. Post-Accident Hydrogen Moniter As part of the generic requiremunts of the TMI Action Plan, identified in NUREG-0737 Supplement I and as implemented by Generic Letter 83-28 all licensees and applicants for an operating license are required to monitor containment hydrogen concentrations.
In response to the requirements of 10 CFR 50.44, NNECO claimed the benefit of the normally inerted atmosphere during reactor operation for its Mark I BWR containment.
Inerting reduces the potential for combustion or deflagration by reducing the oxygen content in the containment atmosphere to a level where it is not capable of reacting with hydrogen produced by an accident.
In response to the post-accident hydrogen monitor requirements of the TMI Action Plan, NNEC0 installed a single channel hydrogen nonitoring system.
By letter dated July 30, 1984, the staff issued a safety evaluation for the Millstone jnit No. 1 post-accident hydrogen monitoring system and concluded that the monitoring system meets the TMI Action Plan requirerents with the exception of redundancy.
However, it was noted that Millstone Unit No. I has the capability to determine both hydrogen and oxygen concentrations through the use of the Post-Accident Sampling System (PASS).
This system provides only grab samples.
The staff indicated that NNECO should demonstrate that tte sampling and analysis frequency of the PASS is sufficient to respond ta any rapid changes in hydrogen or oxygen concentration before the single channel analyzer can be considered acceptable.
By letter dated November 25, 1985, NNECO reiterated its position on the acceptability of the hydrogen monitoring system as installed at Millstone Unit No. I and stated that further information supporting this conclusion will be submitted in early 1986.
At that time, the staff will complete its review of this topic. The licensee should recognize in its integrated assessment of ISAP topics that this issue is still open and should consider alternate instrumentation in conjunction with Topic 1.09, Items 6 and 7.
3.1.12 Topic 1.12, Control Room Habitability As part of the generic requirements of the TMI Action plan, identified in NUREG-0737 Supplement I and as implemented by Generic Letter 82-33, all licensees were to assure that control room operators are adequately protected against the effects of an accidental release of toxic or radioactive gases and that the nuclear power plant can be safely shutdown under design basis accident conditions.
By letter dated October 24, 1985, NNECO proposed to evaluate alternative methods of protecting control room operators, because of the high cost of control room HVAC design modifications, in ISAP.
The staff concurs with the scope of this topic.
The evaluation of ISAP Topic 2.07, Sodium Hypochlorite System, should be considered in conjunction with this topic.
3.1.13 Topic 1.13, BWR Vessel Water Level Instrumentation As part of the genuric requirements of the TMI Action plan, as identified in NUREG-0737 and as implemented by Generic Letter 84-23, licensees were to modify or sc pleme.,t existing equipmen+ to assure accurate indication r
of vessel water level.
In a letter dated October 11, 1955, the licensee presented a concern (i.e., reference leg flashing) which could produce inaccuracies in vessel water level monitoring.
The licensee proposed to investigate different approaches to resolve the reference leg flashing issue and to evaluate the need for further evaluation of the water level instrumentation in the integrated assessment.
The staff concurs with the scope of this topic.
3.1.14 Topic 1.14, Appendix J Modifications Appendix J to 10 CFR 50, " Primary Reactor Containment Leakage Testing for Water-Cooled Reactors," was issued in 1973, more than two years after the POL was issued for Millstone Unit No. 1.
An NRC concern that plants completed or reaching an advanced stage of design or construction by the time of the issuance of the regulation prompted a request to licensees to determine if they were in full compliance with Appendix J.
Specifically, NNECO was requested to identify any design features that do not permit conformance to the recuirements or existing technical specifications which are in conflict with Appendix J.
By letter dated October 16, 1985, NNECO proposed to evaluate the current status of Millstone Unit No. I compliance with the requirements of Appendix J in ISAP. NNECO intends to identify remaining areas of non-compliance, propose acceptable means of addressing all open issues and develop schedules for resolution of such issues. The staff concurs with the scope of this topic.
3.1.15 Topic 1.15, FSAR Update Section 50.71(ej(3)(ii) of 10 CFR Part 50 requires that those plants initially subject to the NRC's Systematic Evaluation Program (SEP) must file a complete updated final r'ety analysis report within 24 months after receipt of notificaticn that the SEP has been completed.
By letter dated March 16, 1983, the staff informed NNECO that SEP had been completed for Millstone Unit No. I and that, pursuant to 10 CFR 50.71(e)(3), the licensee was required to file an updated Final Safety Analysis Report (FSAR).
In the original NNECO Millstone Unit No.1 ISAP proposal, dated December ?8, 1983, NNECO proposed to establish the schedule for the FSAR update in an integrated assecsment of all issues. Also, by letter dated February 4, 1985, the licensee requested an exemption to defer submittal of the updated FSAR on the bases that ongoing reviews (e.g., the Integrated Safety Assessment Program) would directly affect the content of the updated FSAR and that the required submittal would affect the work of the licensee's engineering perscnnel on issues of higher safety significance.
The NRC staff reviewed the licensee's request and granted the exemption, by letter dated April II, 1985.
However, the staff determined that only a six month exemption from compliar.ce with 10 CFR 50.71 was warranted at that time.
By letter dated September 13, 1985, the l censee requested an additional i
exemption from the schedular requirements of 10 CFR 50.71 for the submittal date of an updated FSAR.
However, the NRC staff, by letter dated October 2, 1985, informed NNECO that this exemption request was not responsive to the terms identified in the exemption granted on April 11, 1985, and therefore, the exemptia-request was denied. The denial was based on the licensee's failure '
iy the required milestones or schedules, but left open the possibiL further consideration by the staff provided the licensee s
took certste actions by October 11, 1985.
By letter dated October 11, 1985, NNECO resubmitted a request for an exemption to the schedular requirements of 10 CFR 50.71 for an updated FSAR.
In that letter, NNECO stated that the FSAR update would be based largely on the format of Regulatory Guide 1.70, Revision 3, which would result in a more conprehensive document than would result from minimal compliance with 10 CFR 50.71.
NNECO also provided milestones for the submittal of component parts of an updated FSAR and a completion schedule for the entire FSAR update.
An exemption until March 31, 1987 was granted by the staff in a letter dated November 22, 1985. Therefore, according to the strict terms of that exemption, the schedule for the submittal of the FSAR update has been established and the work necessary to complete it will be carried on independent of ISAP. Therefore, Topic 1.15 is considered resolved.
However, the licensee should determine the resources required to complete the FSAR update and design changes resulting from ISAP as they affect the integrated schedule.
3.1.16 Topic 1.16, Appendix R Appendix R to 10 CFR 50, " Fire Protection Program for Nuclear Power Facilities Operating Prior to January 1, 1979," requires that licensees establish a fire protection program at their nuclear power plants.
The program is to establish the fire protection capability for structures, systems and components important to safety, and develop the procedures, equipment and personnel required to implement the program.
NNECO, as a result of the review of proposed exemption requests submitted for Millstone Unit No. 1 for compliance with 10 CFR 50.48 and Appendix R to 10 CFR 50, identified four plant modifications to bring the plant into compliance with fire protection requirements. These modifications are:
1)
Provide capability to use a diesel generator from Millstone Unit No. 2 to supply emergency power to Millstone Unit No. I during a control room or turbine building fire and provide backfeed capability trat would allow Unit 1 to supply Unit 2 with shutdown power for a Unit 2 control room fire.
2)
Modify the control rod drive pumps to operate without an external source of cooling water during a control room or turbine building / intake structure fire.
3)
Modify portions of the shutdown cooling system to allow the plant to be cooled to cold shutdown conditions without the use of service water.
4)
Protect the three drywell penetrations in the shutdown cooling pump cubicle that contain power and control cables to certain valves.
By letter dated November 6, 1985, the staff issued an exemption to certain requirements of 10 CFR Part 50, Appendix R, Section III.G.2, in response to the request by NNECO.
The exemption was based, in part, upon the licensee's commitment to implement the four proposed rodifications listed above.
NNECO, in a letter dated November 21, 1985, informed the staff that they had performed a revalidation and verification of the completeness and accuracy of their previous Appendix R submittals since the submittal of their original Appendix R review.
NNECO stated that, as a result of their review, new fire areas were established and the corresponding fire boundaries were identified. The new fire er.alysis resulted in a request by NNEC0, in a letter dated November 21, 1985, for new exemptions from the requirements of 10 CFR part 50, Appendix R.
NNECO stated that approval of the previous exemptions do not affect their new submittal insofar as the previously proposed modifications and exemptions are no longer required since they have been replaced by the proposed modifications and exemptions in the November 21, 1985, submittal.
NNEC0 has stated that the four items to be evaluated in ISAP are still valid given the results of their Appendix R re-review.
The November 21, 1985, exemption request mainly addresses the adequacy of existing fire barriers. The staff concurs with the scope of this topic.
3.1.17 Topic 1.17, Replacement of Motor Operated Valves 10 CFR 50.49, " Environmental Qualification of Electric Equipment Important to Safety for Nuclear Power Plants," requires that the environmental qualification program at Millstone Unit No. 1 be completed at the end of the second refueling outage following March 31, 1983.
The staff granted an extension of the completion date to November 30, 1985 for 28 valve motor operators. These operators were the only components yet to be qualified to achieve compliance with 10 CFR 50.49.
By letters dated September 30, 1985 and October 16, 1985, NNEC0 stated that, upon further evaluation, eleven valve motor operators would be replaced by the end of the 1985 Millstone Unit No. I refueling outage (before November 30, 1985), exemption requests for six valves have been submitted and that the remaining eleven valves would require schedule extension requests so that they may be evaluated in ISAP.
By letter dated December 18, 1985, the staff issued an evaluation that found the six valve motor operators for which a permanent exemption was requested to be outside the scope of equipment to be environmentally qualified. Therefore, an exemption for these six motor operators was not necessary.
On November 20, 1985, the Commission issued a Memorandum and Order to NNEC0 regarding the deadline for environmental qualification of electric equipment at Millstone Unit No. 1.
The subject of the Memorandum and Order was the Commission's decision regarding the eleven valve motor operators for which NNEC0 requested a schedular exemption.
The Commission approved an extension of the schedule to the next outage of sufficient duration after the staff has made a determination on whether an exemption to 10 CFR 50.49 can be granted, or to the next refueling outage, but in no case later than August 30, 1987.
NNEC0's proosed scope for this topic is to evaluate the safety and risk aspects associated with the environmental qualification of the remaining eleven valve motor operators.
By letter dated January 17, 1986, NNECO submitted a request for exemption for the eleven valve motor operators based on the criteria of 10 CFR 50.12.
Four of the eleven valves (valves 1-LP-15A and B, 1-LP-16A and B) are the drywell spray valves which are located in the containment spray line and provide isolation to prevent inadvertent spraying of the drywell.
Two valves (1-IC-2, 1-IC-4) (.re the isolation condenser steam inlet isolation and condensate return valves.
They are two of the four isolation condenser primary side isolation valves (the other two isolation condenser isolation valves are already qualified). Two valves (1-00-2, 1-CU-3) are in the reactor water cleanup system (RWCU) and serve as the inboard and outboard isolation valves on the RWCU suction line. Two valves (1-RR-2A and B) are used to isolate the low pressure coolant injection system from a break in the recirculation pump piping.
The last of the eleven valves (1-MW-96A) is an isolation valve between the main condenser hotwells and the condensate storage tank. The staff is presently evaluating the exemption request for these eleven valves and will address the safety benefits of the proposed exemptions in the integrated assessment.
3.1.18 Topic 1.18, ATWS GDC 20 requires that nuclear power plants be provided with a protection system which will automatically initiate the operation of appropriate systems to ensure that specified fuel design limits are not exceeded as a result of anticipated operational occurrences and serse accident conditions and initiate the operation of systems and components important to safety.
Section 50.62 of 10 CFR 50 was promulgated to require additional protection against an anticipated transient without scram (ATWS), an anticipated operational occurrence followed by a failure of the protection system to shutdown the reactor.
As stated by NNECO in letters dated August 13 and September 9, 1985, Millstone Unit No. 1 is c';rrently in compliance with the requirementr of 10 CFR 50.62 regarding alt =rnate rod insert'on and recirculation pump trip systems. However, as documented by NNECO, Millstone Unit No. 1 is equipped with a Standby Liquid Control System (SLCS), but, its capacity does not meet the requirements of 10 CFR 50.62.
NNEC0 plans to evaluate the significance of this difference and perform an assessment of the need to increase the SLCS capacity 'n the integrated assessment. The staff will address the licensee's atement of compliance for the other two items required by 10 CFR 50.62 in the integrated assessment. The staff concurs with the scope of this topic.
3.1.19 Topic 1.19, Integrated Structural Analysis As part of the Millstone Unit No. 1 SEP review (Topics II-3.B. II-4.F, III-2, III-3, III-4, III-6 and III-7.B), components and structures were evaluated with regard to structural adequacy under various loading conditions. As documented in the IPSAR Supplement, many issues still remain to be resolved. These issues include the following areas:
1)
II-4.F pile / soil interaction, turbine building evaluation 2)
III-2 tornado loads 3)
III-3.A groundwater, ficoding 4)
III-3.B flooding, masonry walls 5)
III-4.B tornado missile protection 6)
III-6 seismic considerations TII-7.B load combinations and code changes In a letter dated October 7, 1985, the staff issued a safety evaluation in which the method to achieve safe shutdcwn in the event of flooding is resolved if it can be shown that the blockwalls of the fire pump house are adequate to resist flood loads. Also, the staff concluded that additional analyses should be conducted to confirm the margins of safety for all Category I structures for loading conditions in combination with groundwater.
The staff recommended that these analyses be considered in ISAP.
In a letter dated October 16, 1985, NNECO proposed to consolidate these structural issues and address them in ISAP.
NNECO has submitted information concerning the open issues.
This information is currently under staff review.
NNEC0 should consider the resources necessary to complete these open items and possible plant modifications in the integrated assessment. The staff concurs with the scope of this topic.
3.1.20 T- 'ic 1.20, MOV Interlocks As part of the Millstone Unit No. 1 SEP review (Topic III-10.A), the thermal overload devices for motor operated valves were reviewed against the criteria of GDC 13, 21, 22, 23 and 29, as implemented by IEEE Standard 279-1971, with " ard to their trip settings and size.
As documented in the IPSAR supplement, all of the issues raised in SEP Topic III-10. A have now been resolved. Therefore, this topic is resolved.
3.1.21 Topic I.21, Fault Transfers As part of the Millstone Unit No. 1 SEP review (Topic VI-7.C.1), the AC and DC power systems were reviewed against the criteria of GDC 2, 4, and 18, as implemented by SRP Sections 8.2 and 8.3 and Regulatory Guide 1.6, to ensure that the onsite electrical power supplies and distribution systems have sufficient independence to perform their safety function assuming a single failure.
As documented in the IPSAR supplement, all of the issues raiseo in the SEP review were resolved except for the design of the automatic bu3 transfer devices and the reed for interlocks on three load centers that are manually transferred between redundant sources.
By letter dated October 16, 19E5, NNECO stated that the 480 VAC motor control center EF-7 has oeen disarmed and cannot automatically transfer loads of one load group to the power source of another load group. DC motor control centers DC-11A-1, 2, and 3 are each provided with manual transfer switches which are administratively controlled.
The feeder breaker from the alternate DC source to the transfer switch is kept open (de-energized).
The remaining six automatic bus transfers were reviewad by NNECO.
It was found that disarming er removal of the devices was not practical due to the safety function of the devices.
The October 16, 1985 NNECO submittal and the September 6, 1985 importance analysis for this topic are currently under review by the staff and will be addressed in the integrated assessment.
3.1.22 Topic 1.22, Electrical Isolation As part cf the Millstone Unit No. 1 SEP review (Topic VII-1.A), the isolation of the reactor protection system was reviewed against the criteria of 10 CFR 50.55a(h), as implemented by IEEE Std. 279-1971, to ensure that safety signals be isolated from non-safety signals and that no credible failure at the output of an isolation device can prevent the associated protection system channel from meeting the mininum performance requirements.
As documented in the IPSAR supplement, the staff found that there were no isolation devices between the nuclear flux monitoring systems and the process recorders and indicating instruments, nor were there any isolation devices between the average power range monitor and the process computer.
By letters dated January 31, 1984 and October 16, 1985, NNECO submitted additional information regarding these two open items.
This information is currently under review and will be addressed in the integrated assessment.
3.1.23 Topic 1.23, Grid Separation Procedures Refer to Section 3.1.25, Topic 1.25, Degraded Grid Vcitage Procedures.
3.1.24 Topic 1.24, Emergency Power As part of the Millstone Unit No. 1 SEP review (Topic VIII-2), the emergency power sources were reviewed against the criter,a of GDC 17, as implemented by SRP Section 8.3.1, with regard to preventive maintenance programs for the gas turbine generator.
As documented in the IPSAR supplement, NNECO was to review its preventive maintenance program for the gas turbine generator to identify areas for improvement because operating experience at Millstone Unit No. I and PSS perspectives indicated that one-quarter of the dominant accident sequences involved gas turbine' failure.
The licensee responded by letters dated May 10, 1983 and June 20, 1983.
In these responses, the licensee prnposed changes in the preventive maintenance program regarding the engine mounted feel pump, fuel shutoff valve, air start regulating valve, air receiver and control circuit timers.
Subsequently, by letter dated June 12, 1984, NNECO stated that both the engine-mounted fuel pump and fuel shutoff valve had been replaced in 1976 and upon advice of the engine manufacturer, NNEC0 will inspect these components to the frequency recommended by the manufacturer.
In addition to the changes in the preventive maintenance program recommended by the staff, NNECO has made hardware modifications to increase reliability of the gas turbine generator.
By letter dated August 16, 1984, the staff concluded that the revised maintenance program acceptably resolved this issue.
Therefore, this issue is resolved.
3.1.25 Topic 1.25, Degraded Grid Voltace Procedures As part of the Millstone Unit No. 1 SEP review (Topics VII-3 and VIII-1.A),
the power supply system was ev'luated against the criteria of GDC 17, with regard to the ability of the plant to ccpe with degraded grid voltage.
If the grid voltage were to degrade and become low enough so that the voltage at the Class 1E equipment is less than its qualified operating voltage.
quipment damage could occur.
In a letter dated October 9,1985, NNECO submitted interim degraded grid voltage procedures that would be implemented pending completion of the Millstone Unit No. I design modifications.
The design modifications, identified as Millstone Unit No. 1 Item 3 in the staff's July 31, 1985 letter on ISAP issues, was scheduled to be completed during the 1985 refueling outage.
Final degradeo grid voltage precedures were to be developed following the implementation of the modifications.
The original scope of tne ISAP issue was to submit interim procedures to the staff and then develop a priority for '.he final procedures in the integrated
~
assessment.
By letter dated November 14, 1985, NNEC0 stated that it had reviewed the proposed design changes using the Millstone Unit NJ. 1 PSS to evaluate their impact on plant safety.
NNEC0 found that the chance of a station blackout following a loss of the switchyard would be approximately 2.4 times greater with the new design than with the existing undervoltage protection.
The application of PP.A methods, having identified the increased probability of a station blackout, led NNEC0 to the determination that the design change would not be piudent until the design can be refined. As a result, NNEC0 proposed postponing the design modification until the 1987 refueling outage.
By letter dated December 12, 1985, the staff found the licensee's change of schedule to be acceptable, although the staff required additional information to confirm that the adverse impact of completing the degraded grid voltage modificaticns outweighs the intended benefits of the improvements.
By letter dated January 13, 1986, NNECO provided a description of the probabilistic assessment results and the design analysis of the degraded grid protection modifications which wer? to have been implemented.
This inforration is currently under review by the staff and will be addressed in the integrated assessment.
The scope of this topic should include the proposed redesign of the degraded grid voltage protection system and possible procedural changes.
3.1.26 Topic 1.26, Ecuipment Classification / Vendor Interface (GL 83-38, Item 2.1)
NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," as implemented by Generic Letter 83-28, recuired licensees to: (1) confirm that all components whose functioning is required to trip the reactor are identified as safety-related on documents, procedures and information handling systems used in the plant for controlling safety-related activities, including maintenance, work orders anc parts replacement, and (2) establish, implement and maintain a continuing program to ensure that vendor information is complete, current and controlled throughout the life of the plant, and appropriately referenced or incorporated in plant instructions and procedures.
By letter dated September 9, 1985, NNECO stated that at Millstone Unit No. 1, components required for reactor trip are identified as safety-related on documents, procedures and information handling systems used in the plant to control safety-related activities.
These components are currently identified on the Category 1 Material, Equipment and Parts List (MEPL) and will be identified on the Production Maintenance Management System (PMMS) data base, which will replace the document form of the MEPL.
Details on the PMMS and information submitted by NNECO in letters dated November 8, 1983 and May 9, 1985, are currently under review by the staff and will be addressed in the integrated assessment.
3.1.27 Topic 1.27, Post-Maintenance Testing Procedures (GL 83-28, Items 3.1.1 and 3.1.2)
NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," as implemented by Genaric Letter 83-28, required licensees to submit the results of their review of test and maintenance procedures and technical specifications to assure that pcst-maintenance operability testing of safety-related components in the reactar trip s dem is required to be conducted snd that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service. Also, licensees were required to submit the results of their check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures and technical specifications, where required.
By letter dated September 9, 1985, NNECO stated that a review of the Millstene Unit No. I test and maintenance procedures and technical specifications indicatea that post-maintenance operability testing is required in all cases. Also, all known applicable vendor and engineering recommendations regarding testing have been included in test and maintenance procedures. This information and information submitted by letter dated November 8, 1983, was reviewed by the staff and by letter dated February 21, 1986, the staff concluded that the licensee complied with Items 3.1.1 and 3.1.2 of Generic Letter 83-28.
Therefore, this issue is resolved.
3.1.28 Topic 1.28, Post Maintenance Testing TS Changes (GL 83-28, Item 3.1.3)
NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," as implemented by Generit Letter 83-28, required licenses to identify, if applicable, any post-maintenance test requirements in existing technical specifications which can be demonstrated +'
egrade rather than enhance safety. Appropriate changes to those i..
requirements should be submitted for staff review and approval.
By letter dated November 8,1983, NNEC0 provided a response to Item 3.1.3 of Generic Letter 83-28.
In tb. response, no technical specification requirements for post-maintenance testing were identified which will result in degraded safety of the reactor trip system or other safety-related components. The licensee has committed to continuously review incoming vendor information and engineering recommendations with respect to impact on component reliability should the potential for degradation of safety due to post-maintenance test requireents be identified, the licensee will submit appropriate technical specification changes and justification for NRC approval.
In a letter dated September 3,1985, the staff issued a safety evaluation for this topic in which the staff concluded that the licensee's response to Item 3.1.3 was acceptable and that their commitment to review new information with regard to this item provides additional assurance that the Millstone Unit 1 technical specifications will continue to provide a basis for safe plant operation.
On the basis of the September 3, 1985 safety evaluation, this topic is resolved.
3.1.29 Topic 1.29, Response to Generic Letter 81-34 On August 31, 1981, the staff sent Generic Letter 81-34 to BWR licensees wherein pMnt-specific responses conforming to the guidance contained in NUREG-0803, " Generic SER Regarding Integrity of BWR Scram System Piping,"
were requested. The NUREG-08rG guidelines essentially addressed the need for imptovement in procedures, periodic inservice inspection and surveillance for the scram discharge volume, and environmental qualification for essential equipment needed for mitigation of the consequences of postulated pipe failures in the scram system piping.
These guidelines were developed to address the consequences of a postulated crack in the system piping resulting in large leakage downstream of the isolation valves. The scope of Topic 1.29 was to respond to the requirements of Generic Letter 81-34.
By letter dated September 13, 1985, NNECO augmented the response to Generic Letter 81-34.
NNEC0 stated that the scram discharge volume system was modified in 1982 to replace all piping downstream of the hydraulic control unit header and the single scram discharge volume was replaced by two larger volumes. The new piping meets the design, fabrication, installation, testing and quality assurance requirements of ASME Code,Section III Class 2 components.
The Ming and supports were designed to seismic Category I criteria.
Also, NNECO participated in the BWROG efforts to address the concerns of NUREG-0803.
By letter dated January 3, 1986, the staff forwarded Generic Letter 86-01 to all BWR applicants and licensees.
In that generic letter the staff concluded that (1) through wall leakage cracks or breaks do not need to be postulated; (2) the SDV piping system leakage from a flaw will not give rise to a harsh environment so that no additional environmental qualification of equipment is necessary; (3) revised BWROG Emerger.cy Procedure Guidelines, together with normal plant procedures snd the proposed visual verification of piping integrity, are sufficient measures for detecting and mitigating system leakage, and (4) water hammer is not a contributing factor in potential system breaks.
On the basis of the Generic Letter 86-01 findings, NNECO's scram discharge volume systen modifications, NNECO's participation in the BWROG on this issue and the plant specific features of Millstone Unit No. 1, the staff considers this issue resolved.
3.1.30 Topic 1.30, Post-Trip Review Data and Information (GL 83-28, Item 1.2)
NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," as implemented by Generic Letter 83-28, requires licensees to prepare a report which desc-ibes and justifies the adequacy of equipment for diagnosing an unscheduled reactor shutdown. The report, as a minimum, should describe: (i) the capability for assessing sequence of events; (2) the capability to determine the cause of functioning of safety-related equipment; (3) other data and information provided to assess the cause of the shutdown; and (4) the schedule for any planned changes to existing data and information capability.
By letter dated September 9, 1985, NNECO stated that the plant process computer is currently schedule for replacement (see related Topic 2.03).
In a letter dated November 6, 1985, the staff provided a draft report representing the staff's initial judgement of NNEC0's respnse to Generic Letter 83-28, Item 1.2.
In response to a request for additiona' information, NNECO submitted information on Item 1.2 on January 10, 1986.
This information is currently under review and will be addressed in the integrated assessment.
3.1.31 Topic 1.31, Ecuipment Classification / Vendor Interface (GL83-28,Ite;2.y NUREG-1000, " Generic Implications of ATWS b ents at the Salem Nuclear Power Plant," as implemented by Gene-it 'A thir 83-28, requires licensees te describe their program for ensuring that all components of safety-related systems necessary for accomplishing required safety functions are identified as safety-related on documents, procedures and information handling systems used in the plant to control safety-related activities.
Also, licensees wer' required to establish, implement, and maintain a continuing program to ensure that vendor information is complete, current, and appropriately referenced or incorporated in plant instructions and procedures.
By letter dated September 9, 1985, NNEC0 stated that thev have estr.blished criteria to identify systems, structures and components which are safety-related. The information handling system used to 14 ntify these items is the Category 1 MEPL.
This informatior. is currentiy in document form and will eventually replace the MEPL.
This information and informatic;.
submtted by NNECO in letters dated November 8,1983 and May 9,1925, is currently under review by the staff and will be addressed in the integrated assessment.
(Also See Section 3.1.26) 3.1.32 Topic 1.32, Post-Maintenance Testing Procedures (GL 83-28, Items 3.2.1 and 3.2.2)
NUREG-1000, " Generic Implications of ATWS Events at Salem Nuclear Power Plant," as implemented by Generic Letter 83-28, requires licensees to submit the results of their review and maintenance procedures and technical specifications to assure that post-maintenance operability testing of safety-related components in the reactor trip system is required to be conducted and that the testing demonstrates that the equipment is capable of performing its safety functions before being returned to service Also, licensees were requircd to submit the results of their check of vendor and engineering recommendations to ensure that any appropriate test guidance is included in the test and maintenance procedures or technical specifications, where required.
By letter dated September 9,1985, NNECO stated that post-maintenance operability testing is required in all cases. All known applicable vendor and engineering recommendations regarding testing have been inclnded in test and maintenance procedures.
No recommendations were applicable to the technical specifications.
This information and information submitted by letter dated November 8, 1983, has been reviewed by the staff.
3y letter dated February 21, 1986, the staff concluded that this item is resolved.
3.1.33 Topic 1.33, Item 3.2.3 - Post Maintenance Testing TS Changes NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," as implemented by Generic Letter 83-28, required licensees to identify, if applicable, any post-maintenancc est requirements in existing technical specifications which can be demonstrated to degrade rather than enhance safety. Appropriate changes to those test requirements should be submitted for staff review and approval.
By lett r dated November 8, 1983, NNECO provided a response to Item 3.2.3 of Generic Letter 83-28.
In that response, no technical specification requirements for post-maintenance testing were identified which will result in degraded safety of the re ctor trip system or other safety-related components.
The licensee has comitted to continuously review incoming vendor information and engineering recommendations with respect to impact on component reliability.
Should the potential for degradation of safety due to post-maintenance test requirements be identified, the licensee will submit appropriate technical specification changes and justification for NRC approval.
In a letter dated September 3, 1985, the staff issued a safety evaluation for this topic in which the staff concluded that the licensee's response to Item 3.2.3 was acceptable and that their commitment to review new information with regard to this item provide additional assurance that the Millstene Unit 1 technical specifications will continue to provide a basis for safety plant operation.
On the basis of the September 3, 1985 safety evaluation, this topic is resolved.
3.1.34 Topic 1.34, Reactor Trip System Testing (GL 83-28, Items 4.5.2 and 4.5.3)
NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," as ' clemented by Generic Letter 83-28, requires licensees to justify not maki modifications to permit on-line testing ;f the trip system, if the plant is not currently designed to permit such testing.
Also, licensees were required to review existing intervals for on-line functional testing required by technical specifications to determine that the intervals are consistent with achieving high reactor trip system availability.
By letter dated September 9, 1985, NNEC0 stated that since on-line testing is performed at Millstone Unit No. 1, Item 4.5.2 is not applicable.
This information and the BWROG response is currently under review by the staff and will be addressed in the integrated assessment.
3.1.35 Topic 1.35, Reactor Trio System Functional Testina (GL 83-28, Item 4.5.1)
NUREG-1000, " Generic Implications of ATWS Events at the Salem Nuclear Power Plant," as impleme:ted by Generic Letter 83-28, Item 4.5.1, requires licensees to perform on-line functional testing of the reactor trip system including independent testing of diverse trip features.
By letter dated September 9, 1985, NNEC0 stated that on-line functional testing of the Millstone Unit No. 1 reactor trip system is performed at frequencies defined by the technical specifications.
A procedure to fully test the normal and ATWS back-up scram vclves is being implemented in response to this item.
This information ard information submitted by letter dated November 8, 1983 has been reviewed by the staff. By letter dated February 21, 1986, the staff concluded that this item is resolved.
3.1.36 Topic 1.36, Technical Specifications Covered by Generic Letter 83-36 As part of the generic requirements of the TMI Actior. Plan, identified in NUREG-0737 and as implemented by Generic letter 83-36, all licensees of boiling water reactors were to provide technical specifications for assurance that facility operation is maintained within the limits determined acceptable following implementation at each facility.
By letter dated October 15, 1985, NNEC0 provided a status of each of the items in the enclosure to Generic Letter 83-36 for Millstone Unit No. 1.
The staff currently has the information submitted by NNEC0 under review.
NNECO identified several items for which further information will be submitted. The staff concurs with the scope of this topic.
3.1.37 Topic 1.37, Technical Specification Changes to Address 10 CFR 50.72 and 50.73 On January 1,1984, revisions to 10 CFR 50.72 and 50.73 became effective.
Licensees were recuired by Generic Letter 83-43 to propose revisions to their technical specifications to address the revised reporting requirements in the regulations.
By "B11584, Application for Amends to Licenses DPR-61,DPR-21 & DPR-65, Changing Tech Specs Re Reportable Events & Reporting Requirements in Response to Revs to 10CFR50.72 & 73 & [[generic letter" contains a listed "[" character as part of the property label and has therefore been classified as invalid..Proposed Tech Spec Revs Encl|letter dated July 9, 1985]], NNEC0 submitted proposed changes to the Millstone Unit No. 1 Technical Specifications to address 10 CFR 50.72 and 50.73.
This submittal is currently under review by the staff in Region I.
The Region I staff are scheduled to complete their safety evaluation input by March 31, 1986 and the safety evaluation is scheduled to be completed by April 10, 1986.
It is anticipated that little or no further licensee action will be required. Therefore, on the basis of the current status of this topic, this topic need r.ot be considered in the integrated assessment.
3.1.38 Tcpic 1.38, Expand CA List During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of safety / generic issues, derived from the high-priority issues in NUREG-0933, that t e staff felt could be substantially addressed on a plant specific basis in concert with the PSS.
One of these i: sues is Topic 1.38.
(TMI Action Plan Item I.F.1)
NUREG-0660, the TMI Action Plan, identified systems important to safety at TMI Unit 2 that were not designed, fabricated and maintained at a level equivalent to their safety importance. The NRC proposed to develop guidance for licensees to expand their Quality Assbrance (QA) lists to cover equipment important to safety and rank the equipment in order of its importance to safety. The results of the Interim Reliability Evaluation Program ar.d the systems interaction tasks would be used to establish the importance of equipment as it relates to safety.
Experience in use of the revised NRR review procedure for developing QA lists for individual operating lic3nse applicants would also be factored into the generic guidance.
By letter dated October 16, 1985, NNECO presented information regarding the classification of structures, systems and components in the Millstone Unit No. 1 OA program. NNECO stated that structures, systems and components at Millstone Unit No. 1 are categorized into two groups:
Safety related equipment and non-safety related equipment and, as such, their QA program is a two-tier program.
NNECO stated that further expansion of the QA program or redefinition of QA program categories is unnecessary at Millstone Unit No. 1.
Non-safety related structures, systems and components are routinely evaluated for their impact upon safety related equipment. The licensee contends that several recent licensing issues (e.g., environmental qualification and block walls) substantiate this statement.
In those cases where the General Design Criteria in 10 CFR 50, Appendix A, are applicable, NNECO interprets the regulations to require that any non-safety related structure, system or component that could irpact safety related structures, systems, or components fron performing the function necessary to comply with the particular GDC would need to be desicned to preclude that impact.
NNEC0 stated that management routinely applies accepted codes and standards to plant backfit projects involving non-safety related equipment. Design, installation and maintenance activities are subject to measures to ensure the quality of that equipment.
The Millstone Unit No. I maintenance program addresses all plant equipment, not just safety related equipment.
Based on the degree of attention given to non-safety related equipment at Millstone Unit No. 1, NNEC0 does not believe that a measureable increase in safety would be obtained by expansion of the OA safety related list to include any additional non-safety related equipment.
By letter dated August 13, 1985, the NRC forwarded to NNECO the Systematic Assessment of Licensee Performance (SALP) for Millstone Unit No. I for the period from September 1, 1983 to February 28, 1985, which is the most recent rating period. Three rating areas (plant operations, maintenance, and surveillance) are directly applicable to this topic.
In the area of plant operations which include engineering support, design control, training, staffing, and overall conduct of facility operations the staff found that a strong sense of forehandedness was evident during several planned power reductions for corrective maintenance.
0A audits are done by corporate staff and are generally broad in scope.
The audit staff is small and aggressive with good communications to senior management.
Analysis of LERs indicated a high level of performance.
In the area of maintenance, the equipment classification program was reviewed to determine proper equipment classification commensurate with system application.
The program was judged to be effective. The current manual program is planned to be replaced by an automated system.
QA involvement in post-maintenance testing and in modifications was reviewed and found to be aggressive, well supported by management and generally thorough.
In the area of surveilla.nce, a master surveillance control list correlates sur. illance to licenree requirements and receives management oversight.
Individual departmental controls are effectively used to schedule and track completion of surveillance.
The reting for each of these three areas was 1, the highest SALP rating category. This rating together with the fact that, as of August 13, 1985, Millstone Unit No. I was on-line for 374 continuous days, are indicative of good QA maintenance and surveillance programs with management support and involvement for both safety related and non-safety related equipment.
The staff finds that the QA program as implemented by NNEC0 at Millstone Unit No.1 is an appropriate and effective program and would benefit very little from any change in QA list requirements.
The staff concludes that NNEC0 satisfies the d-+ ant of Topic 1.38 and, therefore, this topic is considered adequately resolved.
3.1.39 Topic 1.39, Radiation Protection Plans During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of safety / generic issues, derived from the high-priority issues in NUREG-0933, that the staff felt could be substantially addressed on a.lant specific basis in concert with the PSS. One of these issues is Topic 1.39 (TMI Action Plan Item III-D.3.1).
NUREG-0660, the TMI Action Plan, proposed a project to improve nuclear power plant worker radiation protection programs by better defining the criteria and responsibility for such programs. Detailed appraisals of health physics programs at all operating facilities were performed and found that certain generic deficiencies existed at many plants due in part to lack of specific performance criteria and/or assigned responsibility. The establishment of a radiation protection plan as a guidance document for implementing procedures has been proposed.
By letter dated October 22, 1985, NNEC0 stated that the radiation protection plan at Millstone Unit No. I compares well with the draft guidance document and that, although some differences do exist, all of the banic criteria is addressed.
- 1
During the licensing review of Millstone Unit No. 3, the Mil h tone Radiation Protection Plan, which is applicable to the three Millstone units, was reviewed by the staff.
The staff concluded in its evaluation (NUREG-1031) that the radiation protection prcgram will maintain in-plant radiation exposures within the applicable limits of 10 CFR 20 and will maintain exposures ALARA in accordance with Regulatory Guide 8.8.
By letter dated August 13, 1985, the NRC forwarded to NNECO the SALP for Millstone Unit No. 1 for the period from September 1, 1983 to February 28, 1985.
One rating area, radiological controls, is directly applicable to this topic.
The assessment found that the radiation protection program is defined by good policies and procedures. The licensee had good performance :n the area of contamination control, personnel monitoring, radiological surveillance and job control, instrumentation reliability and effluent control. The licensee has also implemented a formal ALARA program designed to analyze specific tasks and effect dose reduction methods, as well as monitor task performance relative to performance goals. Overall, the licensee's performance during major projects involving high levels of radioactivity demonstrated planning and preparation, good procedure develcpment and the establishment of acceptable radiological controls.
However, the staff found some deficiencies in the licensee's radiological controls. These deficiencies were not in the area of the radiation plan itself, but in the area of implementation of the plan.
The SALP rating was a Category 2, which indicates that licensee management attention is evident and resources are adequate and reasonably effective so that satisfactory safety performance is being achieved.
The SALP rating board recommended an evaluation of training for first level supervisors as a measure for improving adherence to requirements and to upgrade adherence to routine radiation protection requirements by individual workers.
On the basis of the radiation protection plan in effect, the staff's review of that plan and the licensee's implementation of the plan, the staff finds that NNEC0 satisfies the intent of Topic 1.39 and, as such, this issue is resolved.
However, as indicated by the SALP finding, NNEC0 should strive to improve its implementation of the radiation protection plans.
3.1.40 Topic 1.40, 30iting Degradation or Failure During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of safety /ger.eric issues, derived from the high-priority issues in NUREG-0933, that the staff felt could be addressed on a plant specific basis in concert with the PSS.
One of these issues is Topic 1.40 (Generic Issue 29).
There are numerous bolting applications in nuclear power plants.
The most crucial bolting applications are those constituting an integral part of the primary pressure boundary.
Failure of these bolts could result in the loss of reactor coolant and thus jeopardize safety operation.
Other bolting applications, such as component support or embedded anchor bolts, are essential for withstanding transient loads created during abnormal or accident conditions.
In recent years, the number of bolting related inc' dents has increased. Therefore, there is increasing concern regarding the integrity of the primary pressure boundary and the reliability of the component support structures.
NUREG-0933 recognized that there was no simple solution to this issue but stated that improvements in one or all of the five following areas could be recommended:
design, materials, fabrication, installation and inservice inspection.
In evaluating the relative priority of this issue in NUREG-0933, the staff only considered improvements in inservice inspection.
In September 1984, INPO SOER 84-5, " Bolt Degradation or Failure in Nuclear Power Plants," was issued and it provided recommendations to address the concerns of this issue.
By letter dated October 16,19E5, NNECO addressed the recommendations of the INP0 50ER.
The staff recognizes that the NNEC0 program for bolting inspection addresses the INPO SOER reconmendations and NRC requirements. As a result, by letter dated December 23, 1985, NNECO informed the staff of the discovery of fully severed bolting on main steam line support base plates. On January 2,1986, the staff concluded that the corrective actions for this incident are adequate to support the safe operation of the plant.
However, the staff requested additional information to identify the cause of the bolting failure.
The staff concludes that NNEC0 meets the intent of Topic 1.40 since the licensee observes industry codes and standards and NRC regulations in the design of, materials for, fabrication of, installation of and inspection of bolting. Therefore, the staff concludes that the licensees present bolting program is adequate and need not be addressed in the integrated assessment, except for the following two exceptions.
1)
The licen,ee should adcress the concerns in the staff's January 2, 1986 letter, and 2)
The licensee should evaluate the need for more frecuent visual examinations of the plant bolting.
(The failed bolting was found during a routine plant walkdown. Alerting plant personnel of the need to examine bolting conditions during system maintenance and related plant activities may satisfy this need).
3.1.41 Topic 1.41, Floodinc of Compartments by Backflow During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of safety / generic issues, derived from the high-priority issues in NJREG-0933, that the staff felt could be substantially addressed on a p' ant specific basis in concert with the PSS.
One of these issues is Tcpic 1.41 (Generic Issue 77).
On July 1,19CG, the NRC issued I&E Information Notice 83-44,"Potertial Damage to Redundant Safety Ecuipment as a Result of Backflow through the Equipment and Floor Design System," to address the concern that improperly designed drain systems could contribute to equipment damage as a result of backflow.
By letter dated October 16, 1985, NNECO submitted an evaluation of the equipment and floor drain system design for Millstone Unit No. 1.
NNECO stated that the results of a review previously conducted for IE Circular 78-06," Potential Common Mode Flooding of ECCS Equipment Rooms at SWR Fecilities," showed that flooding in one ECCS equipment room would not result in flooding of the redundant equipment room since each ECCS room has a separate floor drain sump, with no cross connections.
NNECO also stated that the emergency diesel generator, standby gas treatment, feedwater coolant injection, isolation condenser and standby liquid control systems were not capable of being flooded because of the elevation of these components. Systems such as the gas turbine generator, emergency service water or service water systems are not affected by the equipment or ficor drain system since they can drain directly to Long Island Sound.
Further, Departmental Instruction No. 1-0P-6.13, " Plugging of Floor Drains," states that a Plant Design Change Pequest is required for plugging of floor drains.
This request is reviewed by the plant operations review committee.
A review of the Millstone Unit No. I plant drainage system design by NNECO concluded that there exists the possibility of backflow through drains which empty into catch basins and oil separator pits located outside the recombiner, turbine and gas turbine buildings.
Drains in these buildings for which backflooding could be postulated were identified. Millstone Unit No. 1 Off-Normal Procedure 514A, " Natural Occurrences," has been revised to provide guidance to operators as to which plant drair s are to be temporarily plugged in the event of site flooding.
Based on the design of the Millstone Unit No. I structures and systems, the staff concludes that the possibility of damage to ecuipment from flooding through equipment and floor drain systems is remote.
The possibility exists for equipment flooding caused by backflow from natural site flooding. However, given the existence of a procedure to preclude the backflow and time to implement the procedure (possible maximum flood levels, necessary procedures and equipment to mitigate flood effects, and time to implement procedures were reviewed in SEP and found to be acceptable) makes the possibility of equipment flooding remote.
Therefore, the staff finds that Millstone Unit No. 1 meets the intent of Topic 1.41 and, therefore, this topic is considered adequately resolved.
4 8
3.1.42 Tcpic 1.42, MSL Leakage Centrol Systems During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of safety ar.d generic issues, derived from the high-priority issues in NUREG-0933, that the staff felt cculd be substantially addressed on a plant specific basis in concert with the PSS.
One of these issues is Topic 1.4.2 (Task Action Plan Item C-8).
General Design Criterion 54 of 10 CFR 50, Appendix A, requires, in part, that piping systems penetrating primary 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.
Regulatory Guide 1.96, " Design of Main Steam Isolation Valve Leakage Control Systems for Boiling Water Reactor Nuclear Power Plants," was issued to describe an acceptable method of meeting the requirements of GDC 54.
The NRC conducted a strvey in 1982 which indicated a high frequency of measured MSIV leakage at some plants, which may be significantly in excess of the Standard Technical Specification limit of 11.5 SCFH.
The NRC proposed the need for improved MSIV maintenance, more frequent testing or installation of a MSIV leakage control system. Task Action Plan Item C-8 was initiated to investigate whether: (1) the leakage control system recommended in Regulatory Guide 1.96 was acceptable, and (2) leakage control systems should be backfit to all BWRs.
By letter dated N tober 16, 1985, NNEC0 stated that Millstone Unit No. I does not employ a MSIV leakage control system. The plant technical specification 4.7.F.2.C limits MSIV leakage to 11.5 SCFH.
A review of plant performance from 1970 to 1985 indicated that the MSIVs failed to meet the leakage limit several times, with the maximum leakage rate less than 40 SCFH. Other BWRS have experienced excess leakage rates on the order of thousands of SCFH.
NNECO concluded that since Millstone Unit No. I does not employ a MSIV leakage control system, concerns regarding adverse impact of the system operation on dose consequences cre not applicable.
Further, since leakage rates at Millstone Unit No. I have exceeded limits but are far below those experienced at other BWRs, NNECO concludes that existing maintenanea practices are adequate to ensure valve operability and limit leakage (MSIV maintenance involves preventive refurbishment of the valve seat and poppet; surveillance is conducted monthly).
The staff has reviewed NNEC0's submittal and concludes that the addition of a MSIV leakage control system at Millstone Unit No. I would not provide a substantial improvement in safety. The current surveillance and maintenance procedures together with the operating history of the plant form the basis for the determination that NNECO satisfies the intent of this topic and, as such, this topic is considered adequately resolved.
3.1.43 Topic 1.43, Water Hammer During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of safety / generic issues, derived from the high-priority issues in NUREG-0933, that the staff felt could be substantially addressed on a plant specific basis in concert with the PSS. One of these issues is Topic 1.43 (USI A-1).
%2-By letter dated February 26, 1981, NNECO informed the staff that the analytical efforts to include the effects of water hammer on the isolation condenser supply and return lines were complete. As a result of the analyses, several additional piping restraints were installed. On August 10, 1981, a reactor scram at Millstone Unit No. I resulted in a water hammer event in the isolation condenser.
An engineering evaluation by NNEC0 demonstrated the structural adequacy of the isolation condenser to withstand the water hammer loading.
On November 6, 1981, the staff safety evaluation was issued accepting NNEC0's isolation condenser water hammer analyses, but recommended that penetration X-10A be reevaluated to assure a safety factor of at least four.
By letter dated August 30, 1982, the staff issued a safety evaluation that concurred with NNEC0's short-term and long-term comitments to address possible future water hamer events.
NNECO committed to: (1) implement a feedwater pump trip on high eactor level; (2) lower the RWCU system isolation setpoint to provide system availability for reactor vessel fill and address associated post-accident radiological concerns; (3) evaluate lowering the programmed reactor water level setpoint following reactor trip to avoid vessel overfill, and (4) implement low flow feedwater controller improvements to allow remote system actuation from the control room and maintain constant reactor vessel water level automatically.
By letter dated October 23, 1985, NNECO provided information on the status of USI A-1 for Millstone Unit No. 1.
NNEC0 stated that Item (1) above has been implemented.
Item (2) is being evaluated under ISAP Topic 2.20.
NNECO stated that, in a telephone conference call, Item (3) was evaluated and it was found that the response of the system was such that a cost-bcneficial modification could not be developed to resolve this item.
Item (4) was implemented in 1984. Also, the issue of water hammer events in the Control Rod System were resolved in ISAP Topic 1.22.
Additionally, the reevaluation of Containment Penetration X-10A is being evaluated under ISAP Topic 1.48.
In order to adequately resolve this topic, NNEC0 should present the arguments for Item 3 in the integrated assessment.
The other items in this topic are resolved.
3.1.44 Mic 1.44, Asymetric Blowdown Loads on Reactor Systems During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of safety / generic issues, derived from the high-priority issues in NUREG-0933, that the staff felt could be substantially addressed on a plant specific basis in concert with the PSS.
One of these issues is Topic 1.44 (USI A-2). o
In the event of a postulated reactor coolant system piping rupture at the vessel nozzle, asymmetric LOCA loading could result from forces induced on the reactor vessel internals by transient differential pressures across the core barrel and by forces in the vessel due to transient differential pressures in the reactor <<vity.
This issue was designated as USI A-2 and was resolved upon issuance of NUREG-0609,
" Asymmetric Blowdown Loads on PWR Primary Systems," in January 1981.
PWRs were evaluated by the staff for the effects of asymmetric blowdown levels and were found, on a generic basis, based on fracture mechanics analysis, to be adequately designed for this event.
BWRs are being generically evaluated under the GESSAR review using the same analysis methods as PWRs.
By letter dated July 7, 1982, the NRC requested NNECO to provide the status of the USIs at Millstone Unit No. 1.
Although USI A-2 was not directly applicable to BWRs, the NRC requested all licensees to confirm the applicability of the staff positions on A-2 to their facilities.
By letters dated October 13, 1982 and October 23, 1985, NNEC0 reported on the status of USI A-2.
NNECO stated that although similar types of loads associated with a rupture of piping in primary systems in BWRs are expected to occur, the overall safety significance is considered to be much less because of the lower operating pressures in primary systems of BWRs.
Based on this information, the staff considers this topic to be adequately resolved pending any future requirements resulting from the generic BWR asymmetric blowdown load review.
3.1.45 Topic 1.45, Systems Interactions During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of safety / generic issues. derived from the high-priority issues in NUREG-0933, that the staff felt could be substantially addressed on a plant specific basis in concert with the PSS.
One of these issues is Topic 1.45 (USI A-17).
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 present design, analysis and review procedures may not acceptably accour.t for potentially adverse systems interactions and to recommend the regulatory action that should be taken to rectify deficiencias. On February 16, 1984, the scope of this issue was redefined to limit it to functionally coupled, spatially coupled and induced human coupled interactions which could degrade a safety function.
Millstone Unit No. I was a participant in the Systematic Evaluation Program (SEP). Although SEP was not designed to explicity address the concerns of USI A-17, the review criteria for some SEP topics included evaluations of hazards created by intersystem deficiencies.
- Also, interactions of spatially coupled components (i.e., pipe break effects and seismic considerations) and operating experience were reviewed.
The corrective actions taken as a result of the SEP review will help to preclude adverse systems interactions.
Generic Letter 82-33 (requirements from Supplement I to NUREG-0737),
dated December 17, 1982, detailed requirements for emergency response capability, including a safety parameter display system (ISAP Topic 1.08) and post-accident monitoring capabilities (ISAP Topics 1.09 and 1.10). These requirements will enhance the ability of the operator to perform mitigating actions in response to events including those due to adverse systems interactions.
At the time of this evaluation the resolution of USI A-17 is imminent.
Based on the foregoing discussion, the staff concludes that there is a reasonable assurance that Millstone Unit No. I can be safely operated until the ultimate resolution of this generic issue.
Potential means to resolve this topic will be considered in the integrated assessment.
3.1.46 Topic 1.46, Determination of SRV Dynamic Loads During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of :afety/ generic issues, derived from the A
high priority issues in NUREG-0933, that the staff felt could be substantially addressed on a,lant specific basis in concert with the PSS.
One of these issues is Topic
.46 (USI A-39).
Operation of BWR primary system pressure relief valves can result in hydrodynamic loads on the suppression pool retaining structures or those structures located within the pool.
This dynamic loading issue was addressed in USI A-39 and was resolved with the issuance of NUREG-0793,
" Guidelines for Confirmatory Inplant Tests of Safety Relief Valve Discharges for BWR Plants," for BWR Mark I containments in 1981.
By letters dated October 13, 1982 and October 16, 1985, NNEC0 presented the status of USI A-39 for Millstone Unit No. 1.
NNECO stated that it had submitted a report on pool dynamic loads for Millstona Unit No. 1, which provided a description of the application of the generic Mark I pool dynamic loads and methods, as well as an evaluation of the contairunent and components to acconmodate pool dynamic loads.
This report was reviewed and the staff issued a safety evaluation on September 12, 1984, that concluded that the licensees analysis and modifications are sufficient to satisfy the original design safety margins for the lifetime of Millstone Unit No.1.
On this basis, the staff concludes that Topic 1.46 is resolved.
3.1.47 Topic 1.47, Containment Emergency Sump Performance During the development of the ISAP program for Millstone Unit No. 1, the staff identified a number of safety / generic issues, derived from the high-priority issues in NUREG-0933, that the staff felt could be substantially adcressed on a plant specific basis in concert with the PSS.
One of these issues is Topic 1.47 (USI A-43).
USI A-43 safety concerns deal with post-LOCA conditions that can degrade long-term recirculation capability. These safety concerns deal with the potential loss of pump NPSH margin due to air ingestion to the pump or suction strainer blockage resulting from LOCA generated debris.
Full scale experiments on SWR type suction strainers have demonstrated that for typical submergences and flow rates, the strainers act as effective vortex suppressors such that air ingestion is very small.
Thus, the likelihood of loss of NPSH margin due to air ingestion is remoto.
In Millstone Unit No. 1, the insulation is of reflective metallic and blanket types.
The blowdown and transport of insulation debris to the torus will be impeded by the physical layout of Mil btone Unit No. 1.
For metallic insulation that is transported to the torus, the low bulk fluid flow and the elevation of the intakes relative to the torus bottom makes the probability that this debris will be drawn to the strair,ers low.
Blanket type insulation could be drawn to the strainers and over a period of time lower the RHR flow rate.
It is possible, because the flow rate is monitored and because ficw rates can be reduced substantially later in the accident sequence, that operator actions could mitigate the effects of any blockage.
It is the staff's opinion that Millstone Unit No. I meets the intent of the safety concerns in Topic 1.47 with the exception of the possibility of blanket type insulation causing reduced flow and possible procedural improvements.
The staff reconmends that NNEC0 evalucte these issues in the integrated assessment.
3.1.48 Topic 1.48, Safety Factor for Penetration X-10A As part of the review of the Millstone U9it No. 1 isolation condenser for water hammer loads (refer to ISAP Topic 1.43), the staff found that the design of the isolation condenser was acceptable except that NNECO should demonstrate that the factor of safety for containment penetration X-10A is at least four.
By letter dated October 23, 1985, NNECO stated that the required analysis is currently in progress. The licensee should recognize in its integrated assessment that this issue is still open so that the required resources may be factored into the integrated schedule.
3.1.49 Tocic 1.49, Reactor ' Vessel Surveillance Program Subsequent to the issuance of the staff's screening review, by letter of July 31, 1985, it was determined that the scope of Topic 1.49 did not meet the criteria for inclusion in the Millstone Unit No. 1 ISAP.
Specifically, the staff concludes that Topic 1.49 deals with a routine licensing matter and should, therefore, be evaluated outside of ISAP.
Recognition of the amount of resources needed to complete this issue and all other routine licensing issues in a timely manner should be factored into the development of the integrated schedule by NNECO.
Therefore, Topic 1.49 will no longer be considered in the ISAP.
3.1.50 Topic 2.01, LPCI Remotely Operated Valves 1-LP-50A+B Millstone Unit No. 1 Technical Specification Section 4.7 A.1 requires that the suppression chamber water level be checked once every shift to assure that the water level in the torus is within acceptable limits.
If the water level exceeds the maximum allowable level, water is drained to the radwaste system by opening two normally closed 3-inch manual gate valves (1-LP-50A and B). During certain accident conditions, these valves will become inaccessible.
By letter dated August 13, 1985, NNEC0 proposed to provide the capability of remote operation for these valves. A feasibility study is proposed to determine the best option for providing this capability.
3.1.51 Topic 2.02, Drywell Humidity Instrumentation Presently, neither of the two methods for primary coolant leak detection employed at Millstone Unit No. 1 provides for continuou, leak detection monitoring. Accordingly, there exists a possibility that pipe cracks might propagate to a greater degree before detection than would be the case if the leak detection system operated continuously.
By letter dated September 26, 1985, NNEC0 proposed to perform an engineering evaluation to detemine the best method for continuous monitoring of the humidity, and airborne gaseous and particulate contamination in the drywell. The staff concurs with the scope of this topic.
3.1.52 Topic 2.03, Process Computer Replacerent The process computer at Millstone Unit No. 1 is presently nearing the end of its expected useful life.
The process computer is necessary during plant startup and for monitoring and performing NSSS calculations during power ramp-up and ramp-down conditions. During normal operations, the process computer is used for daily fuel surveillance.
The failure of the process computer during normal operation does not affect power generation and is not limited by technical specification.
As part of the TMI Action plan, all licensees and applicants for operating licenses were required to install a Safety Parameter Display System (ISAP Topic 1.08).
By letter dated September 6, 1985, NNEC0 stated that it was replacing its process computer.
The new process computer will have a larger storage capacity and be capable of meeting the interaction requirements of the SPDS.
The process computer replacement is currently ongoing.
The staff concurs with the scope of this topic.
%2-3.1.53 Topic 2.04, High Stream Flow Setpoint Increase The current high steam flow setpoint for automatic isolation is 120 percent of rated flow.
In order to perform turbine stop valve testing, reactor power must be reduced to approximately 90 percent for the duration of the test (several hours) in order not to reach the high steam flow setpoint and generate an MSIV isolation signal and a reactor trip.
By letter dated August 13, 1985, NNECO proposed to evaluate the feasibility of an increase in the flow setpoint to 140 percent of rated flow.
This increase will allow testing of the turbine stop valves at 100 percent power without reaching the trip setpoint. As presently defined, the increase in setpoint can be accomplished by a calibration change in existing hardware.
The staff concurs with the scope of this topic.
3.1.54 Tooic 2.05, Hydrogen Water Chemistry Study BWR nuclear power plants have exhibited cracking in austentic stainless steel due to intergrarular stress corrosien cracking (IGSCC), a phenomenon wWh can be attributed to susceptible, stressed materials exposed to a carrodent.
By letter dated August 13, 1985, NNECO proposed to study the feasibility of altering the primary water chemistry at Millstone Unit No. I by injecting hydrogen into the feedwater.
The hydrogen injection would suppress oxygen formation (an IGSCC corrodent) in the primary coolant.
.Nowever, the hydrogen will also increase the ve'atility of nitrogen and thereby, increase the concentrati0n of N-16 in the steam, causing a possible increase in personnel exposure.
Also, the hydrogen will change the chemical character of the reactor coolant from oxidyzing to reducing leading to the possibility of different corrosior, problems.
The staff concurs with the scope of this topic.
3.1.55 Topic 2.06, Condenser Retube The main condenser at Millstone Unit No.1 has tubing made of a 70/30 copper-nickel alloy that is susceptible to erosion and corrosion problems caused by sea water.
By letter dated August 13, 1985, NNEC0 proposed to evaluate the replacement of the copper-nickel tubing with titanium tubing. Titanium tubing is not susceptible to the erosion /
corrosion mechanisms. However, it is substantially more expensive and has a lower heat transfer coefficient requiring approximately 10% more tubing to achieve equal condensing capability.
The scope of this topic for the ISAP integrated assessment is the development of a priority for completing +"' condenser retube. The staff concurs with the scope of this topic. NNEC0 has currently scheduled to retube the condenser during the 1987 refueling outage.
3.1.56 Topic 2.07, Sodium Hypochlorite System In order to prevent marine biofouling in cooling water systems using sea water as the cooling medium, chlorine is injected into cooling water systems at Millstone Unit No. 1.
Liquid chlorine is stored on-site in 55-ton rail tank cars.
The possibility exists that if a catastrophic failure of a tank car occurred, a toxic concentration of chlorine gas would be released both on-site and off-site.
By letter dated August 13, 1985, NNECO proposed to replace the gaseous chlorination system by an on-site bulk storage and distribution system of sodium hypochlorite which represents a smaller safety hazard.
The staff concurs with the scope of this topic.
3.1.57 Topic 2.08, Extraction Steam Piping As a result of a generic problem of steam piping erosion at other plants, NNECO has performed inspections of the extraction steam piping and found that the eighth, ninth and eleventh stages of extraction steam piping from the low pressure turbine and the high pressure heaters extraction steam piping need to be replaced.
By letter dated August 13, 1985, NNEC0 stated that the replacement of piping has already been initiated.
The scope of this topic for the ISAP integrated assessment is the development of a priority for completing the piping replacement. The staff concurs with the scope of this topic.
3.1.58 Topic 2.09, Upgrading of P& ids By letter dated August 26, 1985, NNECO informed the staff of a plan to upgrade +.he Millstone Unit No. 1 Piping and Instrumentation Diagrams (P& ids) because of a developing problem of overcrowding, inconsistencies and illegibility of the existing P& ids. As stated in the licensee's letter, imple::,enation of this topic has already begun. Therefore, the resources necessary to complete this project should be factored into the integrated schedule. The staff concurs with the scope of this topic.
In a related topic, Topic 1.15, it was proposed by NNECO to prioritize the implemenation of the Millstone Unit No.1 FSAR update. As stated in the SER section for Topic 1.15, an exemption until March 31, 1987 was granted by the staff for completion of the FSAR update. As part of that exemption, the licensee was to submit certain component parts of the FSAR on a schedule established in the exemption.
The updating of system descriptions :nd pip ha and instrumentation diagrams were considered component parts and are required to be submitted by July 1, 1986. Also, in ISAP Topic 2.29, NNECO proposed to update the FWCI design drawings.
Although the level of detail and scope of the projects is not the same in these three topics, NNEC0 should consider integrating the overlapping parts of the topics in the development of the integrated schedules.
3.1.59 Topic 2.10, Drywell Ventilation System Original design of the drywell cooling system required 4 of the 5 coolers to be in continuous operation. Actual drywell heat loads required all 5 coolers to operate.
For this and other reasons, an additional three coolers were installed, however testing of the coolers identified that a majority of the coolers were operating below their design capacity.
The resultant elevated drywell temperatures could result in premature equipment degradation.
By letter dated August 13, 1985, NNEC0 proposed to perform an engineering evaluation to establish the most cost-effective method for reducing the drywell bulk air temperature. An initial phase of the evaluation has been completed and plant changes recomended.
The staff concurs with the scope of this topic.
3.1.60 Topic 2.11, Stud Tensioners Following development of an automated stud tensioner system, NNEC0 performed a cost-benefit analysis of the new system and found it beneficial because of decreased outage time for any outage requiring reactor vessel head removal.
By letter dated August 13, 1985, NNEC0 proposed replacement of the existing stud tensioners with the new automated system modifying associated equipment, revising procedures and drawings and providing personnel training. The scope of the ISAP topic is the development of a priority for completing the replacement project. The staff concurs with the scope of this topic.
3.1.61 Topic 2.12, Reactor Vessel Head Stand Relocation In 1975, the reactor building crane was modified to provide redundant lift capability. This modification limited the travel of the crane with respect to the reactor building floor resulting in the inability to lineup the centerline of the reactor vessel head stand with the centerline of the crane, causing difficulty in placing the reactor vessel head on its stand.
By letter dated August 13,1985, NNEC0 proposed evaluating the feasibility of relocating the centerline of the reactor vessel head stand.
NNEC0 stated that any modifications will address the provisions of NUREG-0612
" Control of Heavy Loads." The staff concurs with the scope of this topic, 3.1.62 Topic 2.13, Turbine Water Induction Modifications By letter dated October 9, 1985, NNECO informed the staff that Topic 2.13, a licensee initiated plant improvement, has been cancelled.
Therefore, since the licensee no longer has any plans with respect to this topic, Topic 2.13 will no longer be considered in the ISAP.
3.1.63 Topic 2.14, Evaluation and Implementation of NUREG-0577 NUREG-0577, " Potential for Low Fracture Toughness and Lamellar Tearing in PWR Steam Generators and Reactor Coolant Pump Supports", was issued for cormient in November 1979.
In that report, the action plan was expanded to include BWR component supports.
In October 1983, NUREG-0577 Revision I was published.
In that report version, the scope of the action plan was no longer applicable to BWR component supports and the recomendations of the report were only applicable to pressurized water reactors in the construction permit or preliminary design application phase of licensing.
These conclusions were based on a value-impact analysis that demonstrated that modifications to existing support structures could not be justified.
Also, the staff could find no evidence of low fracture toughness or lamellar tearing in BWRs.
Therefore, operating BWRs like Millstone Unit No.1 are not subject to the requirements of NUREG-0577.
However, at the time of the start of ISAP fcr Millstone Unit No. 1, this project was still active in the licensee's project tracking and was therefore included in the proposed scope for ISAP.
Therefore, based upon the scope of NUREG-0577 Revision 1, ISAP Topic 2.14 will no longer be considered in ISAP.
3.1.64 Topic 2.15, Toroue Switch Evaluation for MOVs I&E Information Notice 84-10, " Motor-0perated Valve Torque Switches Set Below the Manufacturers Recommended Value," was issued to inform licensees of recent experience where a licensee determined that a number of torque switches were improperly set.
By letter dated October 2, 1985, NNECO proposed to initiate an investiga-tion to determine if all safety-relatad MOV torque switch setpoints are within the manufacturer's recommended setpoint range and develop an inspection procedure for use in monitoring MOV torque switch setpoints during refueling outages. The staff concurs with the scope of this topic.
3.1.65 Topic 2.16, Reactor Protection Trip System Historically, the Reactor Protection System (RPS) has demonstrated setpoint drift problems which have led to difficulties in maintaining setpoint calibration and accuracy.
By letter dated September 26, 1985, NNECO proposed c project which would: (1) investigate the possible replacement of the 120 second ADS timer with a state-of-the-art timer circuit to alleviate setpoint drift, and (2) remove two pressure switches from the ECCS pump start logic to allow ECCS pump start on either high drywell pressure or low-low water level, and automatic ADS actuation.
The staff concurs with the scope of this topic.
3.1.66 Topic 2.17, 4.16kv, 480V and 125VDC Plant Distribution Protection As part of NNECO's Appendix R re-review 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 a fire would be available despite the failure of circuits not required for safe shutdown. The study was performed for both AC and DC distribution systems.
By letter dated October 1, 1985, NNECO proposed to perform a more extensive review (i.e., exceeding the scope of the Breaker Coordination Study) of the 4.16Kv, 480V and 125VDC plant distribution systems.
This review would include:
a) verification of all protective relay and circuit breaker settings on the 4.1.6 KV and 480 V systems and recoordinttion, if required; b) establishment of criteria for setting selection for all protective devices, if recoordination is required:
c) performance of a fault study to be used in the setting calculations; d) verification of the interrupting and momentary ratings for station service switchgear; e) documentation of protective relaying calculations and settings in accordance with procedures applicable to safety-related equipment; f) establishment of a document control system compatible with the requirements for nuclear plant records; g) verification of containment electrical penetration back-up protection coordination; and h) evaluation of the Appendix R review coordination study recommendations and conclusions.
The staff concurs with the scope of this topic.
3.1.67 Topic 2.18, Scent Fuel Pool Storage Racks / Transportation Cash At Millstone Unit No.1, if the spent fuel storage capacity is not increased, full-core off-load capability will not be possible in 1987 and reload discharge capability will not be possible by 1991.
Loss of full-core reserve could replace the option available for fuel storage and possibly impact maintenance operations during a shutdown.
Loss of spent fuel discharge capability would prevent replacement of depleted fuel assemblies with new fuel, thereby terminating power generation capability at Millstone Unit No. 1 in 1991.
By letter dated October 9, 1985, NNEC0 proposed to evaluate the options available for increasing the spent fuel storage capacity at Millstone Unit No. 1.
The staff concurs with the scope of this topic.
3.1.68 Topic 2.19, DC System Review By letter dated October 9, 1985, NNECO informed the staff that Topic 2.19, a licensee initiated plant improvement study, has been cancelled.
Therefore, since the licensee no longer has any plans with respect to this topic, Topic 2.19 will no longer be considered in ISAP.
3.1.69 Topic 2.20, RWCU System Isolation Setpoint Reduc. tion The Reactor Water Clean-Up System (RWCU) is important in minimizing the amount of radioactivity released to the envi onment in the unlikely event of an a::ident and in keeping occupational doses low by purifying the reac'.or 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. A low reactor vessel water level will also generate a reactor scram.
Plant operations personnel at Millstone Unit No. I have expressed the desire 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 clean-up following a reactor scram.
By letter dated September 6, 1985, NNEC0 proposed to evaluate the feasibility of lowering the isolation setpoint together with the potential radiological consequences, adverse equipment qualification impacts following a break in the RWCU piping with the reduced isolation setpoint and alternative ledt detection methods.
Tne staff concurs with the scope of this ISAP topic.
3.1.70 Topic 2.21, 480V Load Center Replacement of Oil-Filled Breakers The overcurrent trip devices, in the 480V power circuit breaks, are elec'ro-mechanical devices that have worn with age and have decreased in reliau'lity. The devices are oil-filled and can, if a leak develops,
.cause false tripping of the breakers.
This could result in a safety-related piece of equipment being unavailable when needed.
By letter dated October 1, 1985, NNECO proposed to replace the existing Type EC overcurrent trip devices with a solid state design.
The scope of proposed project will include the engineering, design and procurement of material including qualification of the solid state devices to Category 1 standards. As an interim measure, the existing devices were rebuilt during the 1984 refueling outage at Millstone Unit No. 1.
The staff concurs with the scope of this topic.
3.1.71 Topic 2.22, Control Rod Drive System (GRD) Water. lamer Reacting to a report from a CRD system vendor, the BWR Owners Group formed a committee to investigate the potential for water hammers occuring in the scram inlet lines.
By letter dated August 13, 1985, NNEC0 stated that, based on the recommendations of the BWR Owners Group committee, they were currently evaluating water hammer loads in the CRD piping due to scram events.
By letter dated January 3, 19E6 (Generic Letter 86-01), the staff identified concerns associated with pipe breaks in BWR scram systems.
In the letter staff stated that:
"The design basis of SDV piping system has considered transient forces resulting from the worst case control rod drive (CRD) system actuation.
Although water hammer has been analytically postulated and hydraulic instabilities have been experienced in the CRD system, no events have been experienced of a severity significant enough to constitute a water hanner.
Therefore, water hammer is not considered a contributing factor in potential SDV pipe breaks."
Therefore, based upon the staff's cenclusions presented in Generic Letter 86-01, the staff concludes that this proposed study need no longer be considered in ISAP.
3.1.72 Topic 2.23, Instrument, Service and Breathing Air System Improvement The station air systems provide clean dry air for pneumatic controls and utility air for pneumatic tools, filter system, tanks and emergency breathing air. The instrument and service air systems at Millstone Unit No. I have historically required substantial maintenance.
Contaminants including rust / scale, dryer dessicant and " loctite" have been found on valve actuators which have hampered valve actuation.
By letter dated August 13,1985, NNEC0 proposed a plant initiative project to perform an engineering review of the Millstone Unit No. 1 instrument, service and breathing air systems to improve reliability and integrity during both normal and emergency operating conditions. The review is to include, but is not limited'to:
1.
Reliability cf air drying and filtration systems.
2.
Improvement of instrumentation on the systems to monitor demand and detect system problems.
3.
Corrosion problems associated with the piping system.
4.
Adequacy of compressors' capacities to meet imposed demands.
5.
Reliability of air supply at interfaces with vital and safety-related equipment.
6.
Possible introduction of contaminants, degrading the air below OSHA Grade D breathing air specifications.
The staff concurs with the scope of this topic.
3.1.73 Topic 2.24, Offsite Power Systems By letter dated October 1, 1985, NNEC0 described a plant improvement project currently being implemented at Millstone Unit Nos. I and 2.
The project i's intended to identify areas of potential weakness and the means for improving offsite power system reliability, capacity and availability. -The project includes, but is not limited to:
1) installation of a 345kv circuit breaker on one of the two Millstone Unit No.1 offsite power lines.
2) installation of slow speed bus transfer schemes on selected Millstone Unit No. 1 electrical buses.
3) installation of a main generator disconnect device (circuit breaker).
4) installation of a full capacity station RSS transformer.
5) modification of the Flanders 23kv circuit to make the line available to all Millstone units.
Project Item (1) has been completed.
This circuit breaker will increase the reliability of the switchyard and will allow operations personnci to remotely reclose the bus following a fault in the line. The scope of this topic for the ISAP integrated assessment is the oevelopment of a priority for completing the rest of the project. The staff concurs with the scope of this topic.
The resources already being expended and the implementation priority developed in the integrated assessment should be used in the development of the integrated schedule.
3.1.74 Topic 2.25, Drywell Temperature Monitoring System Upgrade Drywell bulk air temperature is an input to various containment response analyses and must be verified on a periodic basis.
By letter dated September 6,1985 NNEC0 proposed to upgrade the drywell temperature monitoring system to monitor specific components located in the drywell.
Implementation of the project is expected to facilitate verification of compliance with 10 CFR 50.49.
The staff concurs with the scope of this topic. The licensee should evaluate the potential for leakage detection provided by this instrumentation in conjunction with Topic 2.02 in the integrated assessment.
3.1.75 Topic 2.26, Reliability Ecuipment By 'etter date August 26, 1985, NNECO proposed a project to aid in the c.ffort of performing less corrective maintenance and more preventive maintenance to improve plant safety and reliability.
The project consists of the procurement of computerized UT instrumentation and vibration monitoring equipment.
The equipment is to be utilized for performing localized nondestructive field tests on com;;onents and equipment. The staff concurs with the scope of this topic.
3.1.76 Topic 2.27, Spare Recirculation Pump Motor By letter dated September 6, 1985, NNEC0 stated that they are presently rebuilding a recirculation pump rotor to provide a spare rotor for Millstone Unit No. I to expedite, if necessary, any future recirculation pump replacement.
The staff concurs in the scope of this topic.
3.1.77 Topic 2.28, Lona-Term Cooling Study In the Millstone Unit No. 1 Probabilistic Safety Study, it was shown that approximately 64% of the total calculated core melt frequency is due to a failure to maintain adequate long-term decay heat removal capability, with failures of the Shutdown Cooling System and Alternate Shutdown Cooling System being the major contributors.
By letter dated August 26, 1985, FNEC0 proposed to perform an study of the long-term cooling capability of Millstone Unit No. 1.
Further, by letter dated December 23, 1985, NNEC0 elaborated on the short-term actions and modeling conservatisms from which NNECO concluded that this issue does not present an immediate safety hazard.
However, the staff believes that this topic ccntinues to offer potentially substantial safety benefits that may be related to other topics; for example, but not limited to:
tornado shutdown (Topic 1.02), station backfeed (Topic 1.16), ATWS (Topic 1.18), isolation condenser reliability (Topic 1.50), main condenser restoration (Topic 1.51) and loss of power event sequences. Therefore, the staff concludes that the objectives and evaluation criteria for this study should be developed based on a comprehensive assessment of the potential relationship to related issues.
3.1.78 Tooic 2.29, FWCI Assessment Study In the past fe' years, several changes have been made to the feedwater system at Mi '
ane Unit No. 1.
During the implemenation of the changes, it was found that the logic to transfer from level control to flow control was di'ferent from the original system description.
By letter dated September 30, 1985, NNECO proposed to initiate a study that would compare the feedwater system drawings with the as-wired system; revise drawings as necessary; compare system descriptions with the verified wiring drawing; identify the limits of operation of the control system; and review the assumptions and impact of the feedwater control system on the design basis transients.
3.1.79 Topic 2.30, MSIV Closure Test Fr<cuency The Millstone Unit No. 1 reactor protection system design incorporates an MSIV closure trip function that initiates a reactor trip when it senses MSIV closure greater than 10%. This is an anticipatory trip and is backed up by the average power range monitor and high reactor pressure trip functions. The Millstone Unit No. 1 Probabilistic Safety Study evaluated MSIV closure events and found that they contributed to 93% of the reactor transient events with main condenser unavailabile, which accounts for 2.44% of the predicted core melt frequency. Also, two out of a total of five MSIV closure events at Millstone Unit No. I have been caused by over travelling of the valve during testing.
By letter dated September 30, 1985, NNEC0 proposed to reduce the frequency of the 10% MSIV closure testing (thereby reducing the frequency of MSIV closure events) from monthly testing to quarterly testing in conjunction with the MSIV closure stroke testing.
In the integrated assessment, NNECO will evaluate the benefit of the reduction in MSIV closure event frequency against the reduction in RPS signal reliability generated by MSIV position.
The staff concurs with the scope of this topic.
3.1.80 Topic 2.31, LPCI Lube Oil Cooler Test Frequency The Millstone Unit No.1 Probabilistic Safety Study identified that one of the major contributors to low pressure coolant injection system unavailability is the failure of the solenoid valve controlling the low pressure coolant injection pump motor bearing lube oil cooling. Operation of the solenoid valve is presently indirectly confirmed during refueling when the low pressure coolant injection pumps are run.
Due to the infrequent confirmed testing of the solenoid valves, a high probability of the valve failing to open was calculated.
By letter dated September 30, 1985, NNEC0 stated that a change in the surveillance testing procedure of the low pressure coolant injection system to confirm opening of the solenoid valve during monthly testing of the low pressure coolant injection pump was completed.
Based upon this information, this topic is resolved.
3.2 Additional Topics to be Considered in the Integrated Assessment The staff's review of the Millstone Unit No. 1 PSS identified four areas where individual contributors to plant risk are significant and may warrant further evaluation. These areas are: (1) Isolation Condenser failure to start (98 x 10-5/yr CMF); (2) Isolation Condenser makeup failure
( ~ 2.8 x 10-#/yr CMF); (3) failure to restore Main Condenser (*1 x 10-4/yr CMF); and (4) Safety / Relief Valves failure to close (~1 x 10-4/yr CMF).
The staff evaluation of the Millstone Unit No. 1 operating experieice through 1984 concluded that there are no continuing major challenges to plant safety inherent in the operations of the plant.
The operating data shows that the majority of the operating experience problems were confined to random events. However, significant recurring problems were noted in five areas: (1) loss of emergency power supply capabilities; (2) pipe cracks; (3) isolation condenser valve failures; (4) MSIV failures; and (5) safety / relief valve setpoint drift.
Item (1) from the operating experience review concerns the availability of the gas turbine generator (GTG).
It was found that no single component failure or operational problem dominated the unavailability of the GTG.
Most of the problems were identified during the monthly tests and not during
~
actual demands. The consequences of the failure of the GTG to operate on demand is a degradation of the ECCS because onsite power for the FWCI is provided by the GTG only.
Item (1) is already being addressed by ISAP Topics 1.01 and 1.24.
Also, during the 1985 refueling outage NNECO implemented GTG elettrical equipment protection modifications.
Therefore, the staff concludes that Item (1) from the operating experience review is adequately addressed in ISAP and no additional topic is necessary to resolve this item.
Item (2) from the operational experience review concerns the continuing problem of pipe cracks due to Intergranular Stress Corrosion Cracking (IGSCC) at BWRs.
Item 18 in the staff's July 31, 1985 letter for actions that are to be conducted independent of ISAP, directly addresses IGSCC at Millstone Unit No. 1.
Also, by letter dated July 1, 1985, the licensee presented the inspection and IGSCC mitigation history at Millstone Unit No. 1, as well as the basis for the planned inservice inspection during the 1985 refueling outage.
On the basis of the information presented by the licensee, the staff concluded in its letter of December 6, 1985 that the IGSCC inspection plan and inspection results adequately supported the return of Millstone Unit No. I to full power for its next 18-month fuel-cycle.
Therefore, the staff concludes that Item (2) from the operating experience review is a routine licensing activity and is being adequately addressed outside of ISAP such that no additional topic is necessary to resolve this item.
Item (3) from the operating experience review concerns isolation condenser failures from improper functioning of the isolation condenser valves.
Millstone Unit No.1 experienced 12 failures of these valves from 1970 through 1984. These failures had a small affect on the operation of the isolation condenser, but did affect the ability to isolate the containment. The staff concludes th6t, given the importance of the function of the isolation condenser and the containment, the licensee should evaluate this issue in the the integrated assessment.
Item (4) from the operating experience review is concerned with the failure of the main steam isolation valves.
The problems with MSIVs that occurred during the early years of plant cperation involved sticking of the air-slide valves. Modifications were implemented by the licensee that improved the quality of the air supply to the valves. Also, a program of testing and maintenance was implemented.
These problems appear to have been resolved.
Problems with the MSIVs in recent years primarily involved failures of the limit switches. All MSIV limit switches were recently environmentally qualified and this may resolve this problem. Also, ISAP Topics 2.23 and 2.30 address service air quality and MSIV closure test frequency. The staff concludes that the concerns in Item (4) of the operating experience review have either been adequately resolved or will be resolved by existing ISAP topics so that no additional topic is necessary.
Item (5) from the operating experience review is concerned with setpoint draft in safety / relief valves. Three events involving safety / relief valves were identified.
In 1977 a safety / relief valve inadvertently opened for no known reason. The other two events were due to setooint drift. This is a generic problem for BWRs and the BWR Owners Group is currently evaluating this problem. The staff concludes that, given the :aportance of the function of the safety / relief valves, the licensee should evaluate short term actions, in the integrated assessment, that could be taken until the generic resolution to this problem is available.
Therefcre, the staff ccncludes that NNECO should evaluate the following additional topics in the integrated assessment by refined PRA methods or deterministic means to see if cost-beneficial solutions to reduced plant risk can be achieved.
Topic 1.50 - Isolation Condenser startup/ makeup failures 1.51 - Failure to restore Main Condenser 1.52 - Safety / Relief valve failure /setpoint drift A short description of the scope of the integrated assessment for each of these additional topics follows.
3.2.1 Topic 1.50, Isolation Condenser Startup/ Makeup Failures The Millstone Unit No. I PSS determined that isolation condenser rt.akeup failures and initiation /startup failures appear in 21 of the 18 dominant sequences. The PSS further indicates that the makeup failures are dominated by a MOV which fails to open and by failure of the two diverse fire water supply pumps. Th initiaion/ restoration failures are dominated by contact pair failure and failure of motor operated valve IC-3 to open.
The licensee should evaluate, by specific probabilistic analyses or by mechanistic analyses, potential improvements to the isolation condenser to improve its availability / reliability.
3.2.2 Tcpic 1.51, Failure to Restore Main Condenser The Millstone Unit No. I pSS determined that failure to restore the main condenser contributes to sequences estimated to have a core melt frequency of about 1 x 10' / year.
The licensee should evaluate methods whereby the capability to restore the main condenser could be improved.
3.2.3 Topic 1.52, Safety / Relief Valve Failure-Setpoint Drift The Millstone Unit No. I PSS identified that failure of the safety / relief valves to close contributes to sequences estimated to have a core melt frequency of about 1 x 10~4/ year.
In addition, the Millstone Unit No. 1 operating experience review identified a general problem of EWR safety /
relief valve setpoint drift. The licensee should evaluate methods whereby the safety / relief valves reliability can be improved.
4.0 Conclusion The staff concludes that, except for those topics already satisfactorily resolved, the 80 topics identified in our July 31, 1985, letter that are to be evaluated in the Millstone Unit No.1 integrated assessment have been appropriately defined or have been redefined in light of new information.
Also, from our review of the Millstone Unit No. 1 PSS and the plant operating experience, we have identified three additional topics that should be evaluated in the integrated assessment.
The staff corsiders the scope of review represented by the Millstone Unit No. 1 ISAP topics, in conjunction with the PSS evaluations and the updated plant operating experience evaluation,
- o be sufficiently comprehensive such that the results of the integrated assessment will provide effective integrated schedules and the basis for future regulatory and licensee actions.
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February 28, 1986 Docket No. 50-336 DISTRIBUTION:
DOCAET FILE PMKreutzer NRC PDR ACRS (IC) local PDR Mr. John F. Opeka, Senior Vice President PD#8, Rdg Nuclear Engineering and Operations ACThadani Northeast Nuclear Energy Company OELD P. O. Box 270 EJordan Hartford, Connecticut 06141-0270 BGrimes JPartlow
Dear Mr. Opeka:
SUBJECT:
CONTAINMENT ISOLATION DEPENDABILITY, NUREG-0737, ITEM II.E.4.2.7 FOR MILLSTONE UNIT 2 In our letter of April 25, 1984, we requested additional information from you regarding (1) the isolation of the 6-inch hydrogen purge valves, (2) the operability / qualifications of the hydrogen purge valves, ar.d (3) the surveillance frequency of the seal leakage testing for the containment purge and the hydrogen purge valves.
This information wcs needed to complete our review of the remaining issues under NUREG-0737, Item II.E.4.2.7, Containment Isolation Dependability for Millstone Unit 2.
In your December 31, 1955 letter, you committed to provide automatic isolation of the 6-inch hydrogen purge valves on high radiation in the containment.
You also provided a summary of the design information concerning the operability and the qualification of these valves, and ir,cluded the basis for the current surve'l-lance frequency of the containment purge valve and hydrogen purge valve seal leakage.
We find your December 31, 1985 response acceptable and consider the above issues resolved.
The enclosed Safety Evaluation contains the details of our evaluation.
Sincerely,
/s/
D. B. Osborne, Project Manager Division of PWR Licensing-B
Enclosure:
As stated cc:
See ne>t page
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PD#e:
PKKreutzer D0sLorne:dd ACTbadani 2/./86 2/11/E6 2 4 /86
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B603120300 860228 PDR ADOCK 050 6
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Mr. John F. Opeka Millstone Nuclear Power Station Northeast Nuclear Energy Company Unit No. 2 cc:
Gerald Garfield, Esq.
Day, Berry & Howard Mr. Wayne D. Romberg Counselors at Law Superintendent City Place Millstone Nuclear Power Station P. O. Box 128 Hartford, Connecticut 06103-3499 Waterford, Connecticut 06385 Regional Administrator, Region I Mr. Edward J. Mroczka U.S. Nuclear Regulatory Commission Office of Executive Director for Vice President, Nuclear Operations Operations Northeast Nuclear Energy Company P. O. Box 270 631 Park Avenue Hartford, Connecticut 06141-0270 King of Prussia, Pennsylvania 19406 Mr. Charles Brinkman, Manager Washington Nuclear Operations C-E Power Systems Cotbustion Engineering, Inc.
7910 Woodmont Avenue Bethesda, Maryland 20814 1
Mr. Lawrence Bettencourt, First Selectman Town of Waterford Hall of Records - 200 Boston Post Road Waterford, Connecticut 06385 Northeast Utilities Service Company ATTN: Mr. Richard R. Laudenat, Manager Generation Facilities Licensing Post Office Box 270 Hartford, Connecticut 06141-0270 Kevin McCarthy, Director Radiation Control Unit Department of Environmentcl Protection State Office Building Hartford, Connecticut 06106 Mr. John Shediosky Resident Inspector / Millstone Box 811 Niantic, Connecticut 06357 Office of Policy & Management ATTN:
Under Secretary Energy Division 80 Washington Street Hartford, Connecticut 06106
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. c, UNITED STATES e
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NUCLEAR REGULATORY COMMISSION f
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W ASHINGTON, D C. 20555
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SAFETY EVALUATION BY THE OrFICE OF NUCLEAR REACTOR REGL'LATION RELATED TO CONTAIhMh I ISOLATION DEPENDABILITY NUREG-0737, ITEM II.E.4.2.7 NORihEA51 NUCLEAR ENERGi COMPANt' MILL 510NE NUCLEAR P0wER STATION, UNIT 2 DOCui NO. 50-336 3
- 1. 0 Introduction In order to complete the review of NUREG-0737, Item II.E.4.2.7, Contain-ment Isolation Dependability for Millstone, Unit 2, by letters dated i
February 9,1983, September 2,1983, and April 25, 1984, the staff requested additional information regarding (1) the isolation of the 6-inch hydrogen pur:
valves, (2) the operability /qualificatiors of the hydrogen purge valvt and (3) the surveillance frequency for seal leak-age testing of the cc *airment purge and hydrogen purge valves.
The following is our evals tion of each of these issues.
2.0 Discussion and Eva'.uation 2.1 Isolation Of The 6-inch Hydrogen Purge Valve In response to the staffs' request to modify the isolation logic for the 6-inch hydrogen purge valves such that they receive an isolation signal from the radiation monitors that sense primary containment atmosphere, the licensee, in its December 31, 1985 letter, committed to isolate the 6-inch hydrogen purge valves (HV-8377, 8378, 8379 and 8380) on a high radiation signal as derived from the existing containment atmosphere high range radiation monitors (RE 8240 and RE 8241).
The setpoint for automatic act :9 tion of the isolation feature will be 10 R/hr.
The modifications to provide this automatic isolation will be implemented during the end of-cycle 8 refueling outage which is currently scheduled for January 1988.
At dose rates below 10 R/hr., the containment venting would be terminated by administrative controls as follows.
The stack monitor will alarm in the control room when the Technical Specification dose limit is reached.
The existing plant procedures require that the operators terminate the release by manually isolating containment within 15 minutes.
Based on the licensee's commitment and the above response, the staff con-siders this issue resolved.
2.2 Operability Oualifications Of The Hydrogen Purge Vaives In response to the staffs' request for information concerning the operability / qualification of the 6-inch hydrogen purge valves, the licensee, in its December 31, 1985 letter, provided a summary of design inf ormation related to the operability and qualification of these valves.
There are two trains of the hydrogen purge system with containment isolation valves both inside and outside the containment for each train.
Each train is provided with independent valving, instrumentation and power sources.
The valves are built to AMSE Section III, Class 2 standards.
The valves close on receipt of a signal from two diverse para-meters (pressurizer pressure-low and containment pressure-high).
They are required to close within 5 seconds.
This closing time is verified at least once every 92 days.
The operators and valves are designed to close against a 60 psi differential pressure at 289 F.
This is well above the 25 psig differential pressure predicted for the containment following a LOCA, 238 seconds after event initiation.
The valves and operator components are also seismically and environmentally qualified.
Based on the above information, the staff considers this issue resolved.
2.3 Surveillance Frecuency In response to the staffs' request that the licensee propose Technical Specifications to require testing of leakage through the 6-inch hydrogen purge and vent vahes at least once every 3 months a' the leakage through the 42-inch containnent purge valves at least once e, cry 6 months, the licensee, in their Letember 31, 1985 letter, stated that no changes will be proposed to the e. isting Technical Specification covering these valves (T5 3/4 6.1.2) which aequires periodic leakage testing in accordance with 10 CFR Part Appendix.
These valves have been tested at every refueling outage since plant opeiation began in December 1975.
Test results to date evidence no failures due to degradation of resilient seals and no degrada-tion more rapid than stated by the manufacturer.
The staff's requirement for more frequent testing was derived from a con-cern over the effect of seasonal weather conditior.s on the valve seals.
At Millstone Unit 2, the containment is completely surrounded by the enclosure building which houses the containment purge and hydrogen purge valves, and the valve seals are not exposed to seatonal weather extremes.
Based on the above information, the staff considers this issue resolved.
3.0 Conclusion In conclusion, the licensee has provided acceptable responses to the con-cerns raised by the staff with regard to the dependability of containment isolation for the containment and hydrogen purge and vent valves.
We, therefore, consider NUREG-0737, Item II.E.4.2.7 to be resolved for Millstone 2.
Principal Contributor:
R. Ferguson
February 2B, 1986 Docket No. 50-336 DISTRIBUTION:
DOCNET FILE PMKreutzer NRC PDR ACRS (10)
Local PDR Mr. John F. Opeka, Senior Vice President PD#8, Rdg Nuclear Engineering and Operations ACThadani Northeast Nuclear Energy Company OELD F. O. Box 270 EJordan Hartford, Connecticut 06141-0270 BGrimes JPartlow
Dear Mr. Opeka:
SUBJECT:
CONTAINMENT ISOLATION DEPENDABILITY, NUREG-0737, ITEM II.E.4.2.7 FOR MILLSTONE UNIT 2 In our letter of April 25, 1984, we requested additional information frov you regarding (1) the isolation of the 6-inch hydrogen purge valves, (2) the operability / qualifications of the hydrogen purge valves, and (3) the surveillance frequency of the seal leakage testing for the containment purge and the hydrogen purge valves.
This infc.wtion was needed to complete our review of the remaining issues under NUREC-u?37, Item II.E.4.2.7, Containment Isolation Dependability for Millstone Unit 2.
In your December 31, 1985 letter, you committed to provide automatic isolation of the 6-inch hydrogen purge valves on high radiation in the containment.
You also provided a summary of the design information concerning t n operability and the qualification of these valves, and included the basis for the current surveil-lance frequency of the containment purge valve and hydrogen purge valve seal leakage.
We find your December 31, 1985 response acceptable and consider the above issues resolved.
The enclosed Safety Evaluation contains the details of our evaluation.
Sincerely,
/s/
D. B. Osborne, Project Manager Division of PWR Licensing-B
Enclosure:
As stated cc:
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- 1
DTH FEB 111386 MEMORANDUM FOR: Thomas Rehm, Assistant for Operations Office of the ED0 FROM:
R. B. Minogue, Director
/
Office of Nuclear Regulatory Research
SUBJECT:
COMMISSION PAPER ON SENATE BILL S.1765 You asked why we sent you a Commission paper on the subject bill, as opposed to a remo to OGC.
In a recent case in which the NRC was asked to comment on a proposed bill, RES comments were sent to OGC rather than to the Conmission.
Those were the RES comments on a bill which would transfer epidemiological research from DOE to NIH. OGC responded to the inquiry on the basis of those RES comments which in turn reflected the position that the Conmission had previously taken on this question.
Subsequently the Commission stated that due to changing circumstances, it wanted to reconsider this cuestion and any future responses to Congress in which the NRC had a special interest or the Conmission
" tales a position."
Although we are not aware -
any request to the NRC to comment on the subject legislation, we noted it ecognized that it could have significant impact on the NRC, particularly
'e light of our plans to withdrew funding for continued operation of thc regional seismolog cal networks. The position recommended in the Commission paper is consistent with my May 23, 1984 testimony before the Subcommittee on Science, Research and Technology and the Subconmittee on Investigations and Oversight of the House Conmittee on Science and Technology, which had been approved by the Commission. This fact has been added to the enclosed copy of the Commissicn Paper.
These considerations are why we prepared the subject paper in the form of a Commission paper rather than a memo to OGC.
Cr5cinnl airned by:
EETRI 3. KI:D2 R. B. Minogue, Director 8603120299 860211 Office of Nuclear Regulatory Pesearch PDR RES B603120298 PDR Distribution /R-2811:
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FOR:
The Commissioners FROM:
Victor Stello, Jr.
Acting Executive Director for Operations
SUBJECT:
SENATE BILL S.1765 WHICH WOULD ESTABLISH NATIONAL SEISMIC DATA CENTERS PURPOSE:
To inform the Commissioners of Senate Bill 5.1765 which provides for the establishment of national seismic data centers and to recommend that the Commission suppoet this bill.
DISCUSSION:
Senate Bill S.1765 proposes to arend the Earthquake Hazard Reduction Act of 1977 to authorize the National Earthquake Hazard Reduction Program to establish two national seismic data centers. One center would be located east of the Continental Divide and one west of it.
As stated in the bill, the two seismic data centers would collect and analyze data from regional seismographic networks, establish a repository for these data, disseminate them to the public and promote activities to enhance public awareness of seismic hazards.
The Nuclear Regulatory Commission has supported regional seismographic networks to collect and disseminate seismic data from the Central and Eastern United States. The seismic data recorded by these networks have been invaluable in numerous licensing reviews and regulatory actions. This type of data is required to update NRC's regulations and regulatory guidance. The data have also been widely used by other Federal, State, private and commercial entities. The proposed seismic data centers would be excellent vehicles to draw all seismic data together in a readily accessible place and format for future use by all interested parties.
This would be an important undertaking in any case; but it becomes even more 50 in light of the NPC's decision to withdraw its support for the regional seismographic networks in approximately the same time frame that these centers could be coming on-line (FY 1987-1988).
These centers cculd be imoortant nuclei from which the current network operators might seek alternate sources of support for the operation of CONTACT:
A. Murphy, PES 427-4078
4 The Commissioners 2
regional seismographic networks in the Central and Eastern United States. Thus, this seismological data y continue to be available to the NRC in the future.
It is the staff's recommendation that the Commission support this proposed amendment, Senate Bill 5.1765, to the Earthquake Hazard Reduction Act of 1977. This recommendation is consistent with the May 23, load. Commission approved testimony of Robert B. Minogue before the Subconmittee on Science, Research and Technology and the Subcommittee on Investigations and Oversight of the House Committee on Science and Technology.
Victor Stello, Jr.
Acting Executive Director for Operations
Enclosure:
Senate Bill 5.1765
FEB 111986 MEMORANDUM FOR: Thomas Rehm, Assistant for Operations Office of the EDO FROM-R. B. Minogue, Director Dfilce of Nuclear Regulatory Research
SUBJECT:
COMMISSION PAPER ON SENATE BILL S.1765 You asked why we sent you a Commission paper on the subject bill, as opposed to a nemo to OGC.
In a recent case in which the NRC was asked to comment on a proposed bill, RES comments were sent to OGC rather than to the Commission.
Those were the RES comments on a bill which would transfer epidemiological research from DOE to NIH.
OGC responded to the inquiry on the basis of those RES comments which in turn reflected the position that the Conmission had previously taken on this question.
Subsequently the Commission stated that due to changing circumstances, it wanted to reconsider this question and any future responses to Congress in which the NRC had a special interest or the Comission
" takes a position."
Although we are not aware of any request to the NRC to comment on the subject legislation, we noted it and recognized that it could have significant impact on the NRC, particularly in the light of our plans to withdraw funding for continued operation of the regional seismological networks. The position recorrended in the Corrission paper is consistent with my May 23, 1984 testimony before the Subcomittee on Science, Research and Technology and the Subconmittee on Investigations and Oversight of the House Committee on Science and Technology, which had been approved by the Conmission.
This fact has been added to the enclosed copy of the Commission Paper.
These considerations are why we prepared the subject paper in the form of a Comission paper rather than a remo to OGC.
Original sigma 1y; R:EERI 3. YIrr;.;
R. B. Minogue, Director Office of Nuclear Reguletory Research Distribution /R-2811:
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FOR:
The Commissioners FROM:
Victor Stello, Jr.
Acting Executive Director for Operations
SUBJECT:
SENATE BILL S.1765 WHICH WOULD ESTABLISH NATIONAL SEISMIC DATA CENTERS PURPOSE:
To inform the Commissioners of Senate Bill 5.1765 which provides for the establishment of national seismic data centers and to recommend that the Commission support this bill.
DISCUSSION:
Sarate Bill S.1765 proposes to amend the Earthquake Hazard Reduction Act of 1977 to authorize the National Earthquake Hazard Reduction Program to establish two national seismic data centers. One center would be located east of the Continental Divide and one west of it.
As stated in the bill, the two seismic data centers would collect and analyze data from regional seismographic networks, establish a repository for tuese data, disseminate them to the public and promote acti';ities to enhance public awareness of seismic hazards.
The Nuclear Regulatory Comission has supported regional seismographic networks to collect and disseminate seismic data from the Central and Eastern United States.
The seismic data recorded by these networks have been invaluable in numerous licensing reviews and regulatory actions. This type of data is required to update NRC's regulations and regulatory guidance. Tk data have also been widely used by other Federal, State, private and commercial entities. The proposed seismic data centers would be excellent vehicles to draw all seismic data together in a readily accessible place and format for future use by all interested parties.
This would be an important undertaking in any case; but it becomes even more so in light of the NPC's decision to withdraw its support for the regional seismographic networks in approximately the same time frame that these centers could be coming on-line (FY 1987-1988).
These centers cculd be impcrtant nuclei from which the current network operators might seek alternate sources of support for the operation of CONTACT:
A. Murphy, RES 427-4078
The Commissioners 2
regional seismographic networks in the Central and Eastern United States. Thus, this seismological data may continue to be available to the NRC in the future.
It is the staff's recoramendation that the Commission support this proposed amendment, Senate Bill 5.1765, to the Earthquake Hazard Reduction Act of 1977. This recommendation is consistent with the May 23, 1984, Commission approved testimony of Robert 3. Minogue before the Subconmittee on Science, Research and Technology and the Subconmittee on Investigations and Oversight of the House Committee on Science and Technology.
Victor Stello, Jr.
Acting Executive Director for Operations
Enclosure:
Senate Eill S.1765