ML20199L054

Forwards Summary Rept of plant-specific Probabilistic Safety Study Including Best Estimate LOCA Analysis Prepared by Northeast Util Svc Co.Willingness to Review Rept W/Staff Expressed

ML20199L054
Person / Time
Site:	Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png
Issue date:	03/31/1986
From:	Opeka J CONNECTICUT YANKEE ATOMIC POWER CO.
To:	Charemagne Grimes Office of Nuclear Reactor Regulation
Shared Package
ML20199L056	List: B12020, Summary of Results of Haddam Neck Probabilistic Safety Study ML20199L054 ML20199L073
References
B12020, NUDOCS 8604100283
Download: ML20199L054 (5)

Text

r --a CONNECTICUT YANKEE ATOMIC POWER COMPANY B E R L I N, CONNECTICUT P o Box 270 HARTFORD. CONNECTICUT 06141-0270 TELEPHONE 203-66s-5000 March 31,1986 Docket No. 50-213 B12020 Office of Nuclear Reactor Regulation Attn: Mr. Christopher I. Grimes, Director Integrated Safety Assessment Project Directorate Division of PWR Licensing - B U.S. Nuclear Regulatory Commission Washington, D.C. 20535 Gentlemen:

Haddam Neck Plant r

P_robabi'.istic Safety Study - Summary Report and Results Northeast Utilities Service Company (NUSCO), on behalf of Connecticut Yankee Atomic Power Company (CYAPCO), has recently completed a plant-specific Probabilistic Safety Study (PSS) including a Best Estimate LOCA Analysis for the Haddam Neck Plant. The studies have been and will be utilized for many purposes. Among the near-term applications for the PSS is for assistance in making backfitting decisions in the context of the Integrated Safety Assessment Program (ISAP) currently being implemented at the Haddam Neck Plant.

Background

During the past several years, NUSCO has been implementing a Living PRA Program for the nuclear plants within the Northeast Utilities (NU) system. The major element of this program is the development, maintenance and use of PRA models, of each of the system's nu:lcar power plants, for assistance in evaluating potential plant backfits and operating procedures modifications. The Living PRA Program affords us the flexibility to quickly and accurately analyze the impact 4 on plant safety of changes to the plant's design configuration and significant changes to operating procedures.

NUSCO has completed the development of a computerized PRA model of the Haddam Neck Plant. This model will be periodically updated to incorporate plant design changes, significant operational changes and relevant updated equipment performance data.

The Haddam Neck PSS utilized current state-of-the-art computational and modeling techniques. Included in the study were a comprehensive analysis of best estimate plant specific LOCA and transient response, Haddam Neck specific operating, reliability and maintenance data, the latest plant design 8604100283 860331 PDR N

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1. 7 j changes and draf t symptom-oriented Emergency Operating Procedures. The scope 'of the PSS also included support system failure initiated events, 3

Interfacing system LOCA events and fire initiated events.(l) '

r r As noted in SECY-85-160,(2) the integrated assessment of the Haddam Neck

Plant is scheduled to be completed during 1986. The schedule was based, in part,  ;

on our schedule for completion of and documentation of a summary report of the  !

! - Haddam Neck PSS during March 1986. In keeping with the schedule for j completion of the integrated assessment, our submittal of the Haddam Neck PSS Summary Report and Best Estimate LOCA Analysis will enable the Staff to begin 4

i the probabilistic safety analysis review phase of the integrated assessment of ,

Haddam Neck.  :

] Results i

j it is Northeast Utilities Corporate Policy to provide dependable and economic ,

i power to its customers, without endangering the health and safety of the public.  ;

As one element of achieving this policy, NU has set forth quantitative safety goals for the design and operation of our nuclear power plants. If it is '

4 determined that any one of our nuclear power plants does not' meet the  !

} acceptance criteria as established by our safety goals, our policy is to_ take the  ;

corrective action necessary to meet the criteria, on a schedule commensurate

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with the level of deficiency found.

} The Haddam Neck PSS calculated a mean core melt frequency of 5.5 x 10-4 per j reactor-year for the Haddam Neck Plant which exceeds one of our safety goals.

! As a result, we have committed resources to work to decrease the overall core

! melt frequency at the Haddam Neck Plant. We have begun evaluating some of j the engineering insights obtained from the PSS, in order to determine the

alternatives available to reduce the calculated core melt frequency. A summary .

! of the engineering insights obtained from the PSS and items we are evaluating to l j reduce the calculated core melt risk at the Haddam Neck Plant are contained in  !

! Enclosure 2.

j As outlined in Enclosures 1 and 2, the PSS and Best Estimate 1.OCA Analysis ,

j yielded many insights into the operation of the plant including:

1 l o Containment heat removal is available for most core melt sequences. As t 1 the availability of containment heat removal is considered to be an 4

effective means of maintaining containment integrity and removing fission

products to prevent large scale radioactive releases if core melt occurs, 1 the public risk impact of a majority of the dominant core melt sequences is j reduced by having containment heat removal available.

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(I) The fire initiated events section of the PSS has not been completed at the present time. NUSCO will document and docket the updated results to the '

Staff following completion of the analysis. .,

(2) SECY-85-160, " Integrated Safety Assessment Program - Implementation 1 Plan," dated May 6,1985.

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, e o An AC independent means of providing containment spray during a station blackout is provided by the diesel-driven fire pump. This feature is important in minimizing the offsite public consequences resulting from a core melt accident due to station blackout.

o Approximately 40% of the calculated core melt frequency at Haddam Neck is attributable to small and medium break LOCA scenarios. The PSS utilized LOCA frequencies which were derived from generic PWR data as presented in WASH-1400. As the Haddam Neck Plant utilizes austenitic stainless steel which has superior ductility toughness and resistance to brittle fracture compared to ferritic carbon steel used in many PWR

, plants, the actual Haddam Neck Plant LOCA frequency is expected to be lower.

}' o The Haddam Neck plant-specific control room simulator has been completed and is currently being used for operator training. This affords

us the opportunity to improve the training of operators to respond to plant transient situations.

In light of the above, and our continuing efforts to identify, evaluate and when necessary implement procedural or hardware modifications to upgrade the i '

operation of the plant, CYAPCO has concluded that continued operation of the Haddam Neck Plant does not pose any undue risk to the public.

P i

Summary Report NUSCO has completed summary reports of the Haddam Neck PSS and Best Estimate LOCA) Analysis which we are providing to the Staff as an attachment to this letter.13 The PSS summary report is a comprehensive summary of the

core melt frequencies for various sequences calculated in the study and the calculational methodologies and analysis techniques utilized in the development of the PRA model. The PSS summary report contains the items outilned belows i

i o Determination of Initiating Events System investigations

- Initiator frequency calculations o Accident Sequence Analysis

- Classification of event sequence outcomes

- Plant systems event tree models Plant support system event tree models I o Plant Systems Reliability Analysis ,

Plant component reliability data collection and analysis 1 Plant systems reliability modeling

! (3) The summary reports consist of the Haddam Neck Probabilistic Safety Study (Volumes 1-4) and the Best Estimate LOCA Analysis (one volume) for the Haddam Neck Plant.

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o Human Reliability Analysis

- Introduction and methodology

- Screening analysis

- Detailed representation of operator action

- Summary of results o Accident Sequence Quantification Matrix quantification

- Core melt accident sequence quantification results

- Containment heat removal reliability consideration In support of the PSS, a Best Estimate LOCA Analysis was performed concurrent with the development of the PSS models. The Best Estimate LOCA Analysis summary report describes the results of comprehensive plant specific accident analyses used to determine best estimate system success criteria, best estimate plant response to failed equipment and components, and time frames for operator recovery of failed systems. These calculations were performed using the NULAP 5 Code with best estimate input values. The Best Estimate LOCA Analysis contains analyses of the scenarios described belows o Station Blackout o incore Instrument Tube Rupture o Steam Generator Tube Rupture o Large Break LOCA o Medium Break LOCA o Small Break LOCA o Total Loss of Main and Auxiliary Feedwater with Feed and Bleed Cooling o Anticipated Transient Without Scram o Total Loss of DC Power In accordance with your request we are providing 26 copies of the PSS and Best Estimate LOCA Analysis to the ISAP Project Directorate for distribution within the NRC (including NRR, Region 1, ACRS, etc.).

In addition to the summary reports, we are providing the Staff with additional information on the scope and results of the PSS as outlined below.

o Enclosure 1 A summary of the results of the PSS is provided herein.

o Enclosure 2 Discussions of the major engineering insights NUSCO has obtained from the PSS Including the following two plant design changes already implemented at the plant are provided herein.

o Plant Design change to climinate the common dependence of emergency diesel generator cooling on Motor Control Center-5 (MCC-5). This dependence, found while doing the analysis, could j be significant during a loss of offsite power event as the diesel ,

generators are dependent on MCC-5 to maintain diesel generator cooling. However, the dies-1 generators are not dependent upon l I

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l MCC-5 to start up followir>g a losrof offsite power event. If one or both diesel generators start-up, accept their emergency loads and properly reenergize - MCC-5 as designed, the original configuration would result'h, long-term dhsel generator cooling availability.

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. o Plant design change to mitigate the effects of the loss of MCC-5 i N,, (as an initiator) by tripping thgcharging pumps.

., s These design changes were. credited in our quantification of the PSS as a

, result of our preliminary evaluations of the PSS models which highlighted these design conditions as being potential significant contributors to the l total core melt frequency at the pt2nt.: ,

' Upcoming Activities 4 7' j

i I, .

i % in order to facilltate Staff review and understanding of the Haddam Neck PSS and to ensure that maximum utilization of the P55 is achieved,in ISAP and other applications, we are willing ts treet witin'yo1r Staff to review the PSS. Our .  ;

efforts to date have focused on' completing a:A documenting this study, and on a limited number of high priord of the PSS. In the near terir.', we plan y issues whichand to complete were documentidentified during the our analysis of conduct fire initiated events and to evaluate the report in more detall to identify additional issues which warrant further study. Consistent with the overall ISAP framework, these issues will be identified, new ISAP topics will be proposed and they will be evaluated in a manner snsistent with the process in place for all other ISAP topics.

Very truly yours,  ;

} s  ; ..  ;

Ti CONNF.CTICUT YANKEE ATOMIC POWER COMPANY ,

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Senior Vice President 3

cc: T. E. (Aurley, Region I (with Enclosures 1 and 2 only) -

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