A07693, Outlines General Approach Which Util Intends to Utilize in Responding to Generic Ltr 88-20, Individual Plant Exams for Severe Accident Vulnerabilities & Summarizes Key Points of Past Involvement in Field of PRA & Accident Mgt
| ML20247R451 | |
| Person / Time | |
|---|---|
| Site: | Millstone, Haddam Neck, 05000000 |
| Issue date: | 07/27/1989 |
| From: | Mroczka E CONNECTICUT YANKEE ATOMIC POWER CO., NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| A07693, GL-88-20, NUDOCS 8908080007 | |
| Download: ML20247R451 (20) | |
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' J ww w m.,nm cw.- (203) 665-5000 July 27, 1989 Docket Nos. 50-213 50-245 50-336 50-423 A07693 Re: Generic Letter 88-20 IPE U.S. Nuclear Regulatory Commission Attention: Document Control Desk Washington, DC 20555 Gentlemen:
Haddam Neck Plant Millstone Nuclear Power Station, Unit Nos.1, 2, and 3 Response to Generic Letter 88-20 Individual Plant Examinations for Severe Accident Vulnerabilities Introduction and Purpose The purpose of this submittal is to outline the general approach which North-east Nuclear Energy Company (NNECO) and Connecticut Yankee Atomic Power Company (CYAPCO), on behalf of Millstone Unit Nos.1, 2, and 3 and the Haddam Neck Plant, respectively, intend to utilize in responding to Generic Let-ter (GL) 88-20, " Individual Plant Examinations for Severe Accident Vulnerabil-ities." It will also summarize the key points of our past involvement in the field of probabilistic risk assessment (PRA). integrated safety. assessment (ISA) and accident management (Ali).
Generic Letter 88-20 establishes requirements for ' individual' plant examina-tions for severe accident vulnerabilities. These requirements include Level 1 and 2 PRA-based evaluations, or the equivalent, to ioentify severe accident vulnerabilities, end actions to reduce risk resulting from identified oux liers. Although not an explicit requirement at this time, GL 83-20 also discusses the development of AM strategies and training to prepare the plant- ;
staff, once an accident has exceeded design basis conditions, to return the '
plant to a controlled state in which decay heat removal is ensured and ~ radio-active materials are confined.
Starting in the late 1970s, Northeast Utilities (NU) implemented a compre- 1 hensive PRA program which presently supports many aspects of plant safety and-operation. This program includes the development of plar.t-specific PRAs for all of our units. As defined in our corporate policy on safety goals, actions are taken to correct PRA-identified plant-specific vulnerabilities. Appropri-a;e actions are developed and implemented.
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U.S. Nuclear Regulatory Commission A07693/Page 2 July 27, 1989 l For Millstone Unit No I and the Haddam Neck Plant, these actions are imple-mented based on a living schedule which is developed via the Integrated Safety ,
Assessment Program (ISAP;. NU has experienced all of the benefits of this l "integread safety assessment" as described in GL 88-20 and presented by :
Mr. Tom dox, of your Staff, at the IPE Werkshop in Ft. Worth. Three of the i more noteworthy benefits that we have seen are: (1) evaluate licensing and generic issues, on a plant-specific basis, against all other pending actions; (2) rank issue importance such that the most important are dealt with first; and (3) provide integrated schedules for licensing, regulatory, and safety issues on a predictable basis. This requires periodic updates of our PRA, and !
continuous involvement of PRA in evaluation of new projects as well as support of ruintenance, surveillance, procedure development, and . safety evaluations.
, Some scenarios beyond the design bases have been extracted from the PRAs and are being utilized for training emergency operations personnel to respond to I accidents beyond the design bases. A portable personal computer-based thermal ,
hydraulic program has been developed to help the emergency response organiza- !
tion predict the onsat of degraded core conditions as they track the evolution j of accident scenarios. i i
It is NU's assessment that our PRA program contains all of the elements of f individual plant examination (IPE) and exceeds the intent of the IPE program.
It is therefore NU's intent to fulfill GL 88-20 requirements by continuing our PRA program. Closure with IPE documentation requirements will be accomplished on a schedule consistent with NRC Staff expectations and with deliverables further described in the following sections. Beyond IPE requirements, we ;
believe that NU already has the framework of an AM program in place, with the l objective to further expand these capabilities, as appropriate. l l
Northeast Utilities' Probabilistic Risk Assessment Development NU's first 'nvolvement with Probabilistic Safety Assessment (PSA) dates back ,
l to August 1976 when Hurricane Belle caused significant salt spray to be blown '
onto the 345-kV insulators in the switchyard of the Millstone Station. This produced arcing that resulted in a complete loss of off-site power. Seaweed and debris began to clog up the cooling water intake structure. For the Millstone Unit No.1 BWR, the emergency diesel successfully started and ran, but the emergency gas turbine failed due to human error. This caused the unavailability of the safety-grade feedwater system, which is the only high-pressure, high-capacity makeup system at Millstone Unit Nn. 1. The plant was successfully cooled using the isolation condenser which had a history of maintenance unavailability due to tube leakage.
This event raised several disturbing questions. What if the isolation con-denser had still been out of service? What if further clogging of the intake structure had eliminated cooling water to the diesel? This event seemed to support the perspective provided by the Reactor Safety Study (WASH-1400),
which had shown that the dominant contributors to damaging the reactor core were not the larps break loss-of-coolant accider,t (LOCA) events for which significant safeguards had been implemented, but events likely to occur during the life of the plant, coupled with plausible failures of decay Mat removal 1
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U.S. Nuclear Regulatory Commission A07693/Page 3 1 July 27, 1989 systems. Further, this implied that a plant conforming to current NRC regula-tions could still have a possible pathway to severe core damage.
These concerns led to an NU corporate decision to. perform a limited-scope decay heat removal system PSA for Millstone Unit No.1. This work was per-formed in-house (completed in 1978) and identified weaknesses in systems believed acceptable because they met licensing requirements. Improvements were committed for implementation, and a similar study was initiated for the ;
Haddam Neck Plant. i The accident at Three Mile Island in 1979 ag'ain highlighted the vulnerability I of nuclear power plants to simple failures in decay heat removal systems, coupled with operator error, and stimulated the use of PSA on a larger scale.
l The accident, and subsequent complications created by the avalanche 'of indi-vidual mandated backfits, raised critical questions within NU with regard to the value of dozens of generically inspired plant backfits.. Of concern was that in the process of implementing all these changes, more important changes could'be neglected; even worse, that the net result.of some of these backfits could be a reduction of safety. A growing sentiment developed that there had i to be a better way to understand, prioritize, and manage safety issues and to j assure that utility-originated . safety projects. would ' be given appropriate J consideration. I In the 1980-1981 time frame, the Interim Reliability Evaluation Program (IREP) was established by the NRC to apply PSA techniques to five. nuclear power plants. One of these five plants chosen was Millstone Unit No.1. A group of two to three individuals dedicated to PSA was formed which laid the foundation !
for NU's later involvement in PSA.
i In the spring of 1981, NU formed a Task Force to propose various options for developing PSA capability. The following recommendations were provided by the !
Task Force
- 1. Rather than contracting our PSA work to outside consulting firms, a' PRA 1 l section would be formed in the Safety Analysis Branch. I 1 l
- 2. The PRA section would be responsible for providing PSA-related support to the engineering, design, and licensing groups. '
- 3. The PRA section was to develop and maintain living PSA models for all NU-operated nuclear power plants.
- 4. A general 5-year plan was developed to staff up and initiate 'PSA models for each of the operating nuclear units, with priority based on the age of the plants; i.e., the Haddam Neck Plant,. followed by Millstone Unit Nos. 1, 2, and 3.
A goal in developing this capability in-house was to develop full plant' models ;
to give NU insights on where it stood in terms of ;aeeting proposed safety .
goal s. NU would then be in a position to judge the merits of further work ~in l
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A07693/Page 4 l July 27, 1989 j 8
various areas. This strategy obviously differs from that of developing !i limited-scope nodels, responding on an individual basis to narrowly defined issues such as' fire protection, additional isolation valves, improved auxil-iary feedwater, etc. NU's general; concern with that approach was that limited-scope PSA models (such as an auxiliary feedwater reliability study)'
might conclude a particular backfit was not a significant improvement over what existed currently, but it could not provide a perspective over whether the existing design was adequate in terms of overali plant safety or whether the proposed design change would be more or less significant than changes in )
other plant. systems in terms of overall plant safety. In summary, theJ 1- l
.imited-scope PSA approach would 'not have provided NU 'with the comprehensive i decision-making tool that was wanted.
The recommendations of- the NU Task Force were fully . endorsed by senior NU management ta July of 1981. In September 1981 the NRC requested NU to provide a Level 3 PRA.to support their review of the operating license application for !
Millstone Unit No. 3. Because a Level 3 PRA of Millstone Unit No. 3 had to be l completed in a short time frame to meet commitments to the NRC, it was con--
cluded that NU did not have the total PRA resources at the time to undertake the effort on its own. Because of this, the Level 3 PRA of Millstone Unit No. 3 was performed with Westinghouse (W) in 1982-1983,- with technical coordi- !
nation from the NU PRA section and was submitted to the NPC on July 27, 1983.
This was desired because Nt' was . going to be ' the eventual user of the PRA model, and it was judged essential to have NU involved, particularly in areas where judgments were be'.ig made on data bases and operator actions. This effort included a full-scope technology transfer from M to NU, which resulted in the installation of the Millstone' Unit No. 3 living PRA model on dedicated computers. The staffing level was increased- to six in 1983 to. provide a minimum number of individuals who could become cognizant in all aspects of risk assessment.
In 1982, NU developed a corporate safety goal identifying public risk and core-melt frequency levels which would be utilized to trigger corrective rctions commensurate with the issues identified via PRA techniques. In ' early 1984, the Millstone Unit No.1 Probabilistic Safety Study (PSS)- (Level 1 PRA) was initiated in-house byJU under the auspices of the ISAP and submitted to the NRC on July 10, 1985. )
In 1986, the Haddam Neck PSS (Level 1 PRA) was submitted to the NRC.(2) This effort, along with the accompanying . Best Estimate LOCA report, was also ;
performed in-house. Additional PRA evaluations have since been performed for a fire flooding for both Haddam Neck. and Millstone Unit No.1.g4)iggl NU personnel are currently working on the Millstone Unit j No. 2 PSS (Level 1 PRA), which is approximately two-thirds complete. ;
4 The PRA staff is currently-involved in the development and utilization cf the PRA models for all four nuclear units operated by NNECO and CYAPCO. An ,
inuhouse, dedicated staff of some 13 engineers and technicians support the NU PRA Program. Collectively, this staff is experienced in all phases of PRA.
inckding plant systems analysis, degraded core and containment analysis, i i
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external events analysis, and source term analysis. A separate Radiological Assessment Branch has lead responsibility in radiological consequences analy-sis. The PRA Program also draws upon. in-house staff with expertise in' tran-sient and LOCA analysis. Many members of. these organizations also serve on the Corporate Emergency Operations Center (EOC) Technical Support team, and thus are cognizant in AM.
It is noteworthy that our PRAs, though utilizing plant-specific data bases, quite closely follow PRA procedure guides.and have been subjected to extensive i
. review. In addition to fulfilling internal quality assurance requirements, ;
our PRAs have been reviewed by the Staff and have received extensive internal review by. training and operations personnel who routinely interact with the PRA engineers and provide feedback on assumptions and conclusions of- the PRA.
This review process is further enhanced by PRA assumption documents that we -
are currently developing and maintaining as a configuration management-type program in support of.our PRA. In fact, these documents enhance the scruta-bility of our analyses and make key assumptions visible to engineering and plant staff.
As stated previously, NU's PSS submittals and -updates. have been numerous.
Internal events analyses have provided enlightening insights regarding plant-specific vulnerabilities. These submittals have, undergone comprehensive ar._
ar scrutable reviews by the Ngggyf contractor personne1' as described in the referenced documents These t. comprehensive reviews have compared our process and results to accepted practice. We believe that your reviews have found both our analytical modeling practices and 'res'ults to be generally acceptable. For example, an ex of your PSS Report on Millstone Unit No. 3gt isfrom as folSecticci k s: 1.5, Conclusion ,
Perhaps the most important benefit From the performance and review l of the PSS is the safety improvement gained (unquantifiable) because of the knowledge gained by NU from the detailed, logical analyses ;
required to put the PSS together.
We believe the above excarpt accurately reflects both our corporate intent and an important objective of the IPE. Furthermore, our ' overall approach ' for ;
examination of the other units is comparable to the logical, analytical j examination of Millstone Unit No. 3.
"Livinn" Uodated Probabilistic Risk Assessment NU maintains a corporate policy on nuclear safety goals. To implement this i' policy, it is NU's intention to maintain and exercise to the fullest .the "Living PRA" concept. The PRA models have been used to varying degrees in support of such activities as conceptual design reviews, safety evaluations, ,
prioritization of quality assurance activities, and prioritization of operator J training on emergency operating procedures (EOPs). Several exercises of the '
emergency plan have been based on important accident sequences identified by the plant-specific PRAs. Since early 1988, all new Project Assignments are )
required by procedure to obtain a PRA-based review. On an as-needed basis, l
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U.S. Nuclear Regulatory Commission A07693/Page 6 July 27, 1989 the PRAs for each nuclear unit are updated to reflect desg)g operational.
changes. Three major updates to the Millstone Unit Nos. I and 3 PSSs have already been performed. Therefore, NU already exceeds the provision of GL 88-20 regarding " certification to reflect current design" and will continue to update the PRA models on a periodic basis.
. Plant Vulnerabilities Identified and Addressed As a direct result of the PSS's finding. NU has implemented a number of l design and procedural modifications at three of our units. For example, the Haddam Neck PSS identified a number of potential . vulnerabilities that have i resulted in design modifications and procedural changes. A critical depen-
' dency ogth emergency diesel generator cooling systems on a common electri-cal bus Motor Control Center 5 (MCC 5) was identified and corrected. ~ The PRA. effort also identified a narrow window of small-break LOCAs for which the emergency core cooli Sjystems would be ' ineffective during the high-pressure recirculation phase As a consequence, procedural changes have been implemented and major design changes have been mads which will become fully operational at the end of the 1989 refueling outage. In addition, design modifications for mitigating the loss of MCC 5 have been implemented as well as the addition of an E0P as a result of PSS insights into the critical dependencies of many systems on this equipment. Attachment 1, Table 1, describes identified vulnerabilities and current values of their contribution to core-melt frequency for the Haddam Neck Plant.
The original Millstone Unit No.1 PSS(16) somittal identified that'two-thirds of the core-melt frequency was due to limitations in long-term cooling sys-tems. A major reanalysis effort was undertaken which resulted in the issuance of the Millstone Unit No.1 Long-Term Suppression- Pool . Cooling Capability Analysis report. This detailed analysis identified new long-term cooling success criteria that refined many of the assumptions in previous evaluations.
The Millstone Unit ho. PSS also identified failure of the Instrument AC automatic bus transfer (iii) switch as a significant contributor to the loss of control power for both low-pressure coolant injection heat exchanger outlet valves (Containment Cooling System supply) and motor-operated valves in the shutdown cooling system. A special procedure has been implemented to test the ABT switch every refueling outage. Attachment 1, Table 2, describes identi-fied vulnerabilities and curront values of their contribution to core-melt frequency for Millstone Unit No. 1.
The Millstone Unit No. 3 PSS identified partial loss of DC power as an impor-tant contributor to core-melt frequency. Prior to commercial operation, a special test was performed to identify the consequences of such a loss. Also, an abnormal operating procedure has been written to address loss of service water, and an alternate means of providing charging pump cooling has been identified to ensure adequate reactor coolant pump seal cooling for this event. To resolve the concern with a potential incore instrument tube rupture with failure of quench spray (S C sequence in WASH-1400) whereby there would be insufficient water in the s6mp for the containment recirculation pumps, steps were added to tne E0Ps to ensure that the containment recirculation i
U.S. Nuclear Regulatory Commission A07693/Page 7 July 27, 1989 pumps would not be damaged if there were no quench spray pumps in operation.
Millstone Unit No. 3 plant design changes performed, since the PSS submittal, have not. resulted in a significant change in core-melt frequency; therefore, no before/after table attachment is included.
A planned AC power backfeed between Millstone Unit Nos. I and 2, an Appendix R motivated project, was found to yield a marginal benefit in terms of public risk reduction. However, PRA insights were used to identify a small change in ;
project scope which would provide a significant benefit to the entire Millstone site in terms of stdion AC blackout protection. By allowir.g a 1 i
provision for Millstone Unit No. 3 to be connected to the backfeed in the future, all three nuclear un ts woulo be able to share on-site electrical
, supplies. In response to the Station Blackout Rule, NU has stated that the statia No. 3 gkV electrical crosstie vill be extended to encompass Millstone l' nit As a result of numerous design and procedural changes, -the core-melt frequen-cies for the Haddam Neck Plant and Millstone Unit No. I have decreased signif-
- icantly from the initial values. Attachment 2 illustrates comparison of core-melt frequencies for NU's nuclear plants.
NU's commitment to the "living PRA" process, to support the corporate safety goal (and broader nuclear safety ethic), results in a continual screening review for plant-specific vulnerabilities. This commitment and associated efforts were established well before the issuance of GL 88-20 and, more importantly, will continue long after the three-year defined examination period has passed. Upon completion of the "back-end" portion of the IPE, we intend to review the accident sequences from the Level 1 PRAs to confirm that all significant contributors or outliers have been responsibly addressed.
Backaround and Northeast Utilities' "Intearated" Evaluation Experience l The NRC-initiated Systematic Evaluation Program (SEP), which began in 1977, was the precursor to similar comprehensive evaluation programs initiated to address reactor safety issues. The Millstone Unit No.1 SEP final report was issued by the NRC as NUREG-0824 in February 1983. The Haddam Neck SEP final report was issued by the NRC as NUREG-0826 in June 1983. Given the positive experience of the SEP effort, NNECO and CYAPCgegsg(ig1983Onthat the NRC Staff expand the SEP integrated assessment November 9, 1984, the Commission issued a Policy Statement (49 Federal Register 45112) on the implement dion of ISAP, again endorsing the concept. The NRC provided the go-ahead for a pilot ISAP involving plants selected by the Staff from industry volunteers. Subsequently, the pilot ISAP was focused on two plants--the-HaddamNeckPlantandMillstone(tjgtNo.1. The Commission approved the pilot program based upon SECY-85-160, an implementation plan for the Haddam Neck Plant and Millstone Unit No. 1 pilot ISAPs.
CYAPC0 and hNEC0 have devoted exter.sive efforts to the development of the ISAP methodology for the Haddam Neck Plant and Millstone Unit No.1. The PSSs have allowed for the identification of the major contributors to risk at Millstone e __ __- _ _ _ ._
U.S. Nuclear Regulatory Commission A07693/Page 8 July 27,1989 Unit No. I and the liaddam Neck Plant, which in turn led to the ISAP for:the initiation of activities to resolve those dominant safety concerns.
and the Haddam Neck P The wereISAP " final" submitted in reports for Millstone letters dated July 31, Unit 1986,Nf23) and December 12, 1986 g respectively. As described in these ISAP final reports, the program was developed to foster effective corporate assessment and decision making and to facilitate an Integrated Implementation Schedule (IIS) with respect to imple-mentation of the existing backlog of plant improvement projects.
In letters dated August 4, 1987,(25) and November 13, 1987,(26) NNECO and CYAPC0 submitted to the NRC Staff updated reports on the ISAP. The submittals provided comments on NUREGs 1184 and 1185, respectivg,y the NRC's draft ISA -
Reports for Millstone Unit N@8f dated April 2,1987, and the Haddam Neck Plant dated August- 18, 1987. These submittals also provided discussion on:
scope, evaluation, prioritization, recommended resolution, and a schedular status for each ISAP topic. Proposed'IISs, based on the review of each topic, were includen as attachments to the letters. Extensive efforts were' and continue to be invested to make the transition from the pilot ISAPs to a longer term, established pregram.
In letters dated March 24, 1988,(29) and March 31,1988,(30) NNECO and CYAPC0 prepared and subinitted proposed license conditions requiring the implementa-tion 'and maintenance of the IIS program plan. The program plans provided a methodology to be followed for scheduling plant modificationgd and engineering evalugns. In subsequent letters dated November 9,1988, 1989 March 2, NNECO and CYAPC0 submitted revised IISs, developed fully consistent with these program plans submitted to the NRC Staff.
NNECO and CYAPC0 recognize the Millstone Unit No. I and 'Haddam Neck Plant experiences as integral to the further development of the ISAP methodology.
ISAP is regarded as an evolving and dynamic process and, as'such,- has proven
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to be an effective approach to nuclear regulation and licensee decision l making, responding to long-standing NRC support for a systematic review of the.
safety of operating nuclear power plants. Therefore, NNEC0 plans to expand ISAP to include Millstone Unit .Nos. 2 and 3, consistent with the provisions for ISA contained in GL 88-20. This expansion .for Millstone Unit No. 2 involves completion of the plant-specific PSS prior to beginning the ISAP. A Level 1 PRA is currently under development (approximately two-thirds' com-plete) . The total resource expenditure to complete this Level 1. effort, including the best-estimate LOCA analysis, is approximately 13 person years.
Efforts to fully implement the ISAP prccess for Millstone Unit No. 3 are developing and will include utilization of the existing Level 3 PSS, subse-quent update, and utilization of the existing analytical ranking methodology to evaluate and prioritize proposed backfit and betterment projects.
At the recent Ft. Worth IPE workshop, Mr. Tom Cox, NRC Policy Development and Technical Support Branch Section Chief, discussed ISA and its relationship to IPE. He described its genesis beginning with the SEP through and integration with GL 88-20. The ISA components consist of a minimum Level 1 PRA. plant ,
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U.S. Nuclear Regulatory Commission A07693/Page 9 July 27, 1989 operating experience review, issue analysis and prioritization, propued integrated schedule, and a procedure for ISA PRA and schedule update.
The ISA, as proposed by the Staff, has an integral relationship with IPE. An ISA PRA plus containment assessment plus licensee evaluation satisfy IPE requirements. The ISA integrated schedule will contain IPE results and schedule. Lastly, if a PRA is done for IPE, it may also be used for ISA. We believe that our approach is fully consistent with the ISA proposed by the Staff.
Accident Manaaement We view AM activities as those actions taken during the course of an accident by the plant operating staff and the technical staff that supports it, consis-tent with the Nuclear Management and Resources Council's position:
o To prevent the accident from progressing to core damage.
o To terminate damage if it begins.
o To maintair, the integrity of the containment as long as possible if the reactor vestel is breached, o To minimize the effects of off-site releases.
Although the focus of accident management is on actions taken during an accident, a key aspect is preparing in advance to ensure that those actions can be taken most effectively. Those preparations can include:
o Developing procedures or other guidance to aid the operating and support staffs during a severe accident.
o Providing necessary training.
o Ensuring that appropriate instrumentation and other information will be available if needed.
o Ensuring that the decision-making responsibilities are established.
With this perspective in mind, we believe that significant AM capabilities are already in place to respond to an evolving accident at one of NU's plants.
For example, the organizational needs of coping with severe accidents have been recognized for some time. In July 1988, the Corperate Organization for Nuclear Incidents (CONI) procedures were modified to recognize a clear dis-tinction between emergency operations within the scope of E0Ps and management of accident scenarios not covered by E0Ps. The characteristic role differ-ences between operators' response during emergency operations and the supWt team members were enhanced in this CONI procedure revision, based on excerpts from an AM paper written by a member of the NRC Staff. The emergency organ-ization directors are sensitized to the role that they may have to assume in a
U.S. Nuclear Regulatory Commission A07693/Page 10 July 27, 1989 severe accident situation, and organizational lines that exist for handling an accident, should it evolve beyond the scope of our improving E0Ps.
In au_.ition, NU has enhanced the emergency organization's thermal-hydraulic (T-H) analytical response capability. The T-H team has been strengthened with all T-H specialists. A fast, accurate PWR predictive code model, for predict-ing the time of core uncovery, has been developed and loaded on a personal computer (PC) dedicated to the T-H team. This initiative goes beyond NRC expectationsandhasbeennotedbyNRCinsgtorsasnoteworthyinaninspec-tion report letter dated December 2,1988.
Although the emergency procedure framework provides reasonable assurance that the operators are adequately equipped to deal with accidents within E0P applicability, NU believes that our evolving AM response capability will be enhanced by the availability of living PRA models implemented with interactive software on dedicated PCs. Two of our PRAs are substantially implemented or being implemented on PCs, and the remaining two PRAs are being developed /
converted onto two more PCs. When all PRAs are installed on PCs, in the event of an accident, our PRA personnel could be mobilized to extract the PRA model, identify current sequence, alternate paths, vulnerability to failures, and provide valuable inputs to emergency organization directors.
Lastly, our safety analysis technical staff continues to support the Nuclear Training Department in familiarizing the emergency organization directors with the responsibilities which they may be called upon to manage during an emer-gency.
We believe that NU currently has the rudiments of an AM program in place. We will continue to develop and expand on these capabilities as appropriate.
Individual Plant Probabilistic Risk Assessment Status A full Level 3 PRA including external events has been developed for Millstone Unit No. 3 and is currently exercised. Internal to NU, Millstone Unit No. 3 participates in an ISA-type program and it is NU's intent that Millstone Unit No. 3 will formally participate in this option as outlined in GL 88-20.
l Independent of ISAP at Millstone Unit No. 3, NU considers the " front-end" and "back-end" portions of the IPE for Millstone Unit No. 3 to be complete. For Millstone Unit No.1, a Level 1 PRA for internally initiated events, as well as fire and internal flooding, is complete and is actively exercised. There-fore, we believe the front end is complete. The back-end analysis effort has been under way at a low level of effort for about a year, and with the issu-a'.cc of guidance from GL 88-20 and NUREG-1335, a full effort will be under way by the end of 1989. Our intent is to complete this work in a manner which will provide insight into the merits of implementing the Staff's recommends-tions on Mark I containments, including hardened venting capability.
Likewise, the Haddam Neck Plant Level 1 PRA, including fire and internal flooding, has been very actively exercised for over 3 years now. Accordingly, we believe its front end is also complete. Major upaates to the PSS to
U.S. Nuclear Regulatory Commission A07693/Page 11 July 27, 1989 reflect major modifications to the plant, particularly the emergency core cooling system and new switchgear building, are scheduled to commence in 1990. The back-end portion will commence following completion of the back-end portion for Millstone Unit No. 1.
As noted earlier, the Hillstone Unit No. 2 Level 1 (internal events) PSS is approximately two-thirds complete and currently scheduled for completion in 1990. The back-end portion of the IPE will commence after the completion of the Haddam Neck Plant back-end analysis. Thus, NU intends to make maximum productive use of personnel and experience by performing the back-end analysis for the remaining plants (Millstone Unit Nos. I and 2, and Haddam Neck) in series. The efforts will be performed essentially in-house. These back-end efforts for Millstone Unit Nos. I and 2 and Haddam Neck, unlike Millstone Unit No. 3, do not currently include full Level 2 PRAs. While NU does not preclude the possibility of performing detailed Level 2 (and ultimately Level 3) PRAs l for these remaining units, we will meet or exceed the provisions, for back-end analysis, set forth in GL 88-20, utilizing an acceptable methodology.
IPE Schedule As noted earlier, a full Level 3 PRA including external events has been completed for Millstone Unit No. 3. NU considers the requirements of IPE to be essentially fulfilled by the existing Millstone Unit No. 3 PSS. To docu-ment compliance with the IPE requirements, a summary report is planned for submittal by mid-1990.
i The Millstone Unit No.1 PSS includes consideration of internal events and internal flooding. Thus, the " front-end" analysis is completed. The "back-end" analysis is scheduled for completion in late 1990. The IPE summary report is planned for submittal in the mid-1991 time frame.
The Haddam Neck Plant PSS also includes consideration of internal events and internal flooding. A major update to the PSS to include the emergency core cooling system modifications and new switchgear building is scheduled for completion in mid-1991. The "back-end" analysis is scheduled for completion l in late 1991. The IPE summary report is planned for submittal in the first l half of 1992.
The Millstone Unit No. 2 PSS (internal events) is currently scheduled for completion in mid-1990, with internal flooding scheduled for completion in late 1990. The "back-end" analysis is scheduled for completion in late 1992.
Since our in-house PRA developmental process is more highly developed for some of our units, the work remaining to be completed varies accordingly. There-fore, completion of the remaining analysis will dictate the corresponding data and information submittals. For example, since the Millstone Unit No. 3 analysis is complete (Level 3 PRA) and based upon previous submittals dis-cussed above, we plan to reference many of these submittals and, as such, the IPE summary report for Millstone Unit No. 3 will contain significantly less information than the corresponding Millstone Unit No. 2 IPE summary report.
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U.S. Nuclear Regulatory Commission A07693/Page 12 July 27, 1989 Similarly, the Haddam Neck Plant and the Millstone Unit No. I summary reports will refer to previously docketed submittals.
Conclusion As a direct result of this company's long-standing management recognition, commitment, and dedication of significant resources to the " nuclear safety ethic," we believe NU has benefited significantly as a result of PRA and
" integrated assessment." This speaks to our inherent commitment to the intent of GL 88-20. Further, we intend to utilize our experienced staff to vigor-ously complete the Individual Plant Examination of our units using accepted and approved techniques to meet the requirements of GL 88-20. NU has already completed significant portions of the requirements of GL 88-20 for our four plants. The remaining portions and the associated documented information submittals will be submitted within the time frame specified in the GL.
Following the actual IPE, we intend to evaluate any contemplated potential design modifications through the ISAP for all four units, consistent with accepted past practice and the pending license condition for two of our four units. Subsequent to the ISAP ranking, we will continue the process by allocatinn resources and scheduling any modifications utilizing tne IIS. We believe that our past experience and future expectations of ISAP are totally congrcent with the ISA/IPE program summarized by Mr. Cox.
We trust that this letter represents both our intention and proactive work on the subject of IPE. Furthermore, we believe that our plant-specific analyses and acknowledged accomplishment with respect to PRA-type analyses are totally consistent with the provisions of GL 88-20. Additionally, for those plants for which PRA/PSS analyses have been submitted, we consider those provisions of GL 88-20 to be satisfied. Beyond GL 88-20, we intend to continue assess-ments and analytical updating, as appropriate, as we have done in the past.
Perhaps a meeting with the Staff during August would be useful in confirming the adequacy and responsiveness of our plans to respond to GL 88-20. We plan to discuss this option with you in the very near future.
Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY j CONNECTICUT YANKEE ATOMIC POWER COMPANY t ?
E. J Broczka Senior Vice Pre rident//
cc: W. T. Russell, Region I Administrator M. L. Boyle, NRC Project Manager, Millstone Unit No. 1 G. S. Vissing, NRC Project Manager, Millstone Unit No. 2 D. H. Jaffe, NRC Project Manager, Millstone Unit No. 3 A. B. Wang, NRC Project Manager, Haddam Neck Plant W. J. Raymond, Senior Resident Inspector, Millstone Unit Nos. 1, 2, and 3 J. T. Shedlosky, Senior Resident Inspector, Haddam Neck Plant
U.S. Nuclear Regulatory Commissior.
A07693/Page 13
. July 27, 1989 References (1) J. F. Opeka letter to J. A. Zwolinski, " Millstone Unit No.1 Probabilis-tic Safety Study--Results and Summary Report," dated July 10, 1985.
(2) J. F. Opeka letter to C. I. Grimes, "Haddam Neck Plant Probabilistic Safety Study--Summary Roport and Results," dated March 31, 1986.
(3) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, "Haddam Neck Plant Probabilistic Safety Study--Fire Analysis," dated June 1,1987.
(4) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, "Haddam Neck Plant Probabilistic Safety Study--Internal Flooding Analysis," dated December 8, 1987.
(5) J. F. Opeka letter to C. I. Grimes, " Millstone Nuclear Power Station, Unit No.1 Probabilistic Safety Study--Fire Analysis," dated March 26, 1986.
(6) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No.1 Probabilistic Safety Study--Internal Flooding Analysis," dated February 13, 1987.
(7) T. M. Novak letter to J. F. Opeka, "Probabilistic Safety Study for Millstone Nuclear Power Station, Unit 3," dated October 17, 1985.
(8) C. I. Grimes letter to R. M. Kacich, " Millstone 1--Final Report on PRA Review of ISAP Issues," dated January 3, 1986.
(9) C. O. Thomas letter to E. J. Mroczka, "Haddam Neck Integrated Safety )
Assessment Program," dated May 27, 1987. ;
(10) Alan B. Hang to E. J. Mroczka, "Haddam Neck Plant Probabilistic Safety Study--Internal Flooding Analysis (TA 67773)," dated December 20, 1988.
(11) T. M. Novak letter to J. F. Opeka, "Probabilistic' Safety ~ Study for Millstone Nuclear Power Station, Unit 3," dated October 17, 1985. ,;
(12) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, "Probabilis-tic. Safety Stdy Update," dated February 11, 1987.
l (13) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Millstone Nuclear Power Station, Unit No. 1 Probabilistic Safety Study Update !
(Revision 2)," dated February 10, 1989. ;
(14) R. H. Graves letter to U.S. Nuclear Regulatory Commission, " Facility.
Operating ~ License No. DPR-61, Docket No. 50-213, Reportable Occurrence LER 50-213/85-029-00," dated December 2, 1985. 1 L-.---__---------_--__-----_ - - _ -
i
'U'.S. Nuclear Regulatory Commission A07693/Page.14 July 27, 1989 (15) D. B. Miller, Jr., letter to U.S. Nuclear Regulatory Commission, "Facil-ity Operating License No. DPR-61, Docket No. 50-213, Reportable Occur-rence LER 50-213/86-013-02," dated July 14, 1988.
(16) J. F. Opeka letter to J. A. Zwolinski, " Millstone Unit No.1 Probabilis-tic Safety Study--Results and Summary Report," dated July 10, 1985.
(17) E. J. Mroczka letter to Dr. T. E. Murley, " Millstone Nuclear Power Station, Unit Nos.1, 2, and 3, Response to Station Blackout Rule," dated April 17, 1989.
(18) W. G. Counsil letter to W. J. Dircks, !' Integrated Assessment of Regula-tory Requirements'" B10804, dated June 13, 1983.
(19) H. R. Denton letter to W. G. Counsil, " Integrated. Assessment of Regula-tory Requirements," dated July 5,1983.
(20) W. G. Counsil letter to D. G. Eisenhut, " Integrated Safety Assessment Program," B10869, dated September 14, 1983.
(21) W. G. Counsil letter to D. G. Eisenhut, " Integrated. Assessment of Regula-tory Requirements," B10986, dated December 28, 1983.
(22) W. J. Dircks letter to U.S. ~ Nuclear Regulatory Commission Commissioners,
" Integrated Safety Assessment Program Implementation Plan," SECY-85-160, dated May 6, 1985.
(23) J. F. Opeka letter to . C. I. Grimes, " Integrated Safety Assessment Program--Final Report for Millstone Unit No.1," dated July 31, 1986.
(24) E. J. Mroczka letter to C. I. Grimes, " Integrated Safety Assessment Program--Final Report for Haddam Neck Plant," dated December 12, 1986.
(25) E. J. Mroczka letter to U.S. Nuclear Regulatory ' Commission, " Integrated Safety Assessment Program," dated August 4, 1987.
(26) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Integrated Safety Assessment. Program," dated November 13, 1987.
(27) C. 0.. Thomas letter to E. J. Mroczka, " Millstone Unit No. 1--Draft l Integrated Safety Assessment Program (NUREG-1184)," dated April 2,1987.
l l (28) C. O. Thomas letter to E. J. Mroczka, "Haddam Neck Plant--Draft Inte-grated Safety Assessment Program (NUREG-1185)," dated August 18, 1987.
(29) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Integrated Implementation Schedule--License Condition," dated March 24, 1988.
(30) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Integrated Implementation Schedule--License Condition," dated March 31, 1988.
U.S. Nuclear Regulatory Commission A07693/Page 15 July 27, 1989 (31) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Integrated Safety Assessment Program," dated November 9, 1988.
(32) E. J. Mroczka letter to U.S. Nuclear Regulatory Commission, " Integrated Safety Assessment Program," dated March 2, 1989. !
(33) L. H. Bettenhausen letter to E. J. Mroczka, " Resident Inspec-tion 50-213/88-19 (9/28/88-11/15/88)," dated December 2, 1988.
i l
l
e Docket Nos. 5,0-213 50-245 A07693 l
Attachment 1 Haddam Neck Plant Millstone Nuclear Power Station, Unit No.1 Contribution of Unique Plant Features to Core-Melt Frequency l
July 1989
p ' N ' 'l U.S. Nuclear Regulatory Commission 1 A07693/ Attachment 1/Page l' July 27, 1989 Table 1 Contribution of Unique Plant Features to-Core-Melt Freauency at Haddam Neck (Internal' Events)~
Before Correct Actions {yy- ' [.yrrent -
Loss of Motor Control Center 5 60% 8%
Single Injection Path for High-Pressure ECC Recirculation 8% 2% '. -!
Loss of DC Power- 3% 9%(2)
Others -J% J%(2)
Total 74% 28%. j Core-Melt Frequency ~10 3/yr 3.4 x 10 4/yr.
.l l
j l
,i I
(1) Actions implemented prior to publication of the PSS.:
(2) Absolute contribution to core-melt frequency has not changed.
l l
4 %
U.S. Nuclear Regulatory Commission A07693/ Attachment 1/Page 2
- July 27, 1989 Table 2 ..
Contribution of Unique Plant Features to Core-Melt Frecuency at Millstone Unit No.1 (Internal Events) .
Before Correct l
Actions {yy Current Containment Cooling System Not Fully dedundant 64% 8%
{
LPCI Lube Oil Cooling Infrequently Tested 13%(2) Small
.. k Service Water Cooling for Diesel Generator Electrically Dependent on Gas Turbine Generator 2%(2) Small Drywell Cocling System Locks Out on Ler;-Uw '
Reactor Water Level Small(2) Small LPCI Loop Select Logic Prevents Opening of Nonselected Iniection Valve Small(2) Small Total 80% 8%
Core-Melt Frequency 9 x 10 4/yr 6.3 x 10 5 'yr (1) Original PSS Analysis (1985)
(2) Changes implemented and reduction reflected in PSS results.
4 4 Docket N;;ul. 50-213 ,
50-245 j 50-336. !
50-423 A07693 l i
i 1
i Attachment 2 Haddam Neck Plant I Millstone Nuclear Power Station, Unit Nos.1, 2, and 3 PRA Core-Melt Frequency Status Report j l
l
{
l l'
July 1989 l
1 E_.______..._._
]
j -. ,0 U.S. Nuclear Regulatory Commission A07693/ Attachment 2/Page 1 July 27,,1989 PRA Core-Melt Freauency Status Report' Connecticut 2 Millstone Millstone -Millstone' Yankee Unit No. 1- Unit No. 2 ~ Unit No. 3' Internal Events - 3.4 E-4/yr 6.31 E-5/yr- (I) 6.3?E-5/yr Fires :5.0 E-4/yr(2) 2.58 E-5/yr (I) 4.8-E-6/yr:
Internal Floods 1.7 E-5/yr 3 2.5'E-7/yr- -(I)- .'Negli gibl e' External Flood III (I)- (I)' '.Negl.igible l
Seismic 'II) -(I)' '(I) '
9.1 E-6/yr.-'
l(
Miscellaneous (3) _
-(1) '(1). f.h Nealiaible Tot'al 3.57 E-4/yr I4) 8.92'E-5/yr II) 7.7E-5/yr; l
l (1) Information not available.
(2) Estimate based on crediting rercoval of PCB oil-filled transformers.during'-
( the last refueling outage.
(3) Examples include: tornados, missiles, and high winds.
(4) Modifications planned for the 1989 refueling ' outage will significantly
~
l improve-this value.
,