ML20076B629

ECCS Repts (F-47) TMI Action Plan Requirements,Re Ginna Nuclear Power Plant, Technical Evaluation Rept

ML20076B629
Person / Time
Site:	Ginna
Issue date:	09/24/1982
From:	Vosbury F FRANKLIN INSTITUTE
To:	Chow E NRC
Shared Package
ML17255A355	List: ML20076B629
References
CON-NRC-03-81-130, CON-NRC-3-81-130, RTR-NUREG-0737, RTR-NUREG-737 TER-C5506-276, NUDOCS 8209290028
Download: ML20076B629 (14)

Text

-

TECHNICAL EVALUATION REPORT 1

S REPORTS (F-47) ,'

CTION PLAN REQUIREMENTS ITER GAS AND ELECTRIC COMPAf#-

GINNA NUCLEAR R0WER. PLANT

. KET NO. 50-244 FRC PROJECT C5506 FRC ASSIGNMENT 7 mACT NO. NRC-03-81 130 FRC TISK 276

% F. W. Vosbury Research Center -

Author: G. J. Overbeck 9 ace Streets B. W. Ludington hia, PA 19103 FRC Group Leader: G. J. Overbeck for

'egu!atory Commission Lead NRC Engineer: E. Chow on, D.C. 20555 September 24, 1982

was prepared as a:1 account of work sponsored by an agency of the United States it. Neither the United States Government nor any agency thereof, or any of their I makes any warranty, expressed or implied. or assumes any legal liability or hty for any thiro party's use. or the results of such use. of any information, appa-

, luct or process disclosed in this report. or represents that its use by such third

,, ,, J not infringe privately owned rights.

~

e: '

Reviewed by: Approved by: -

• L Luthor As Grodp Leader om MAL,-

Department Dir to

-l) t/. @ Date-

Datet b2

2

., , x:r 2 ~s., m

~N

(^ gggk gh

,/-

I: . .

Franklin Research Center A Division of The Franklin Institute

. . . . , ., The Ben >arran Frankkn Paruway. Phda Pa. 19103 4215J 4481000 A

r. - . ,

>,. \ ,j .. . .-v g  ; XA Copy Has Been Sent to PDR

TER-C5506-276 l

l .

CONTENTS Section . Title '

Page 1 INTRODUCTION . .. . .. . . . . . . . . . 1 1.1 Purpose of Review . . .. . . . . . . . . 1 1.2 Generic Background. . . . . . . . . . . 1 1.3 Plant-Specific Background . . . . . . . . . 2 2 REVIEW CRITERIA. . . . . . . . . . . . . 3 3 TECHNICAL EVALUATION . . . . . . . . . . . 4 3.1 Review of Completeness of the Licensee's Report . . . 4 3.2 ~ Comparison of ECC System Outages with Those of Other Plants. . . . . . . . . 5 3.3 Review of Proposed Changes to Improve the Availability .of ECC Equipment . . . . . . . . 8 4 CO NCLUSIONS. . . . . . . . . . . . . . 9 5 REFERENCES . . . . . . . . . ... . . . .. 10 Me ,

. ..:. Franklin Research Center ,

% ww r,. e %.u.

TER-C5506-276 9

FORZWORD '

This . Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. N'uclear' Regulatory Commission (Office of-Nuclear Reactor Regulation, Division of Operating Reactors) for techn'ical .

assistance in support of NRC' operating reactor licensing actions. The technical evaluation was conducted in accordance with criteria established by the NRC. ..

Mr. G. J. Overbeck, Mr. F. W. Vosbury, and Mr. B. W. Ludington contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.. .

~

~

. F V

Mm *

.... Franklin Research Center -

• • A Dresion of The Frannha ennenme

_. _ ._ m... . , _ . . . - . - . .

TER-C5506-276

1. INTRODUCTION -

1.1 PURPOSE OF REVIEW <-

• This technical evaluation report (TER) documents.an independent review of the outages of the emergency core cooling (ECC)' systems.at Rochester Gas and Electric Company's (RG&B) R.. E..'Ginna Nuclear Power Plant. Tne purpose

• of this evaluation is to determine if -the Licensee ~ has submitted a repor't that 'is complete and satisfies the requirements of 'IMI Action Item II.K.3.17, " Report on Outages of Dnergency Core-Cooling Systems Licensee Report and Proposed Technical Specification Changes." .

1.2 GENERIC BACKGROUND Ibilowing_ the 'Ihree Mile Island Unit 2 accident,. the Bulletins and Orders Task Force reviewed nuclear steam supply system (NSSS) vendors' small break loss-of-coolant accident (IDCA) analyses to ensure that an adequate basis existed for developing guidelines for small break LOCA emergency procedures.

During these reviews, a concer,n developed about the assumption of the worst single failure. Typically, the small break LOCA analysis for boiling water reactors (BWRs) assumed a loss of the high pressure coolant injection (HPCI) systera as the worst.. single f ailure. However, .the technical specifications permitted plant operation for substantial periods with the HPCI system out of service with no limit on the accumulated outage time. There is concern not only about the HPCI system, but also about all ECC systems for which substantial outages might occur within the limits of the present technical specification. Therefore, to ensure that the small break LOCA analyses are

~

consistent with the act,ual plant response, the Bulletin and Orders Task Force

~

recommended in NUREG-0626 (1) , " Generic Evaluation of Feedwater Transients and Small Break Ioss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Term Operating License Applications," that licensees of General Electric (GE)-designed NSSSs do the following:

1 l

l 4

1 I

&--m '

l

..:.:, Franklin Research Center ao .ew rr. e .

• TER-C5506-276

" Submit a report detailing outage dates and lengths of the outages for all ECC systems. Th'e report should also include the cause of the outage (e.g. , controller failure or spurious isolation) . The outage data for ECC components should include all' outages "for the last five years of operation. The end result should be the quanti"ication of historical unreliability due to test and maintenance outages. This will establ'ish if a.need exists for cumulative outage requirements in technical specifications."

Later, the recommendation was incorporated into NUREG-0660 (2), '"NRC

• Action Plan Developed as a Result of the TMI-2 Accident," for all GE-designed plants as TMI Action Item II.K.3.17. In NUREG-0737 (3], " Clarification of TMI Action Plan Requirements," the NRd staff expanded this action item to include all PWRs and added a requirement that licensees propose changes that will improve and control availability of ECC systems and components. In addition, the contents of tne reports to be submitted by the licensees were further clarified as.fllows:

o "The report should contain (1) outage dates and duration of outages; (2) cause of the outager (3) 2CC systems or components involved in

• the outage; and (4) corrective action taken."

1.3 PLANT-SPECIFIC BACKGROUND On December 31, 1980 [4), RG&E submitted a report 16 response to "

NUREG-073 7, Item II.K.3.17, " Report on Outages of Emergency Core-Cooling Systems Licensee Report and Proposed Technical Specification Changes." The report submitted by RGEE covered the period from June 1976 through October 1980 for the Ginna plant. Af ter reviewing the data, RG&E stated that no changes are presently proposed for improving the availability of specific ECC equipment.

?

I i

r

~~

f;s3 l ,, P u.'. Frankhn Research at w .en.r. ain. w. C. enter

TER-C5506-276

2. REVIEW CRITERIA The Licensee's response to'iiUREG-U737,' Item II.K.3.17, was evaluated against criteria provided by the NRC in a letter.. dated July 21, 1981 [5-]

outlining Tentative Work Assignment F. Provided as review criteria in Reference 5, the NRC stated that' the Licensee's response should contain' the following information:

1. A report detailing outage dates, causes of outages, and lengths of outages for all ECC systems for the last 5 years of operation. This report was to include the' ECC systems or components involved and corrective actions taken. Test and~ maintenance outages were to be included.

2. .A quantification of the historical unavailability of the ECC systems and components due to test and maintenance outages.

3. Proposed changes to improve the availability of ECC systems, if necessary.

The type of information required to satisfy the review criteria was clarified by the NRC on August; 12, 1981 [6). Auxiliary systems such as component cooling water and plant service water systems were not to be considered in determining the unavailability of ECC systems. Only the outages of the diesel generators .were to be included along with the- primary ECC system outagas. Finally, the "last five years of operation" was to be loosely interpreted as a continuous 5-year period of recent' operation.

On Jely 26, 1982 [7), the NRC further clarified that the purpose of the review was to identify those licensees that have experienced higher ECC system outages than other licensees with similar NSSSs. The need for inproved

_. reliability, of diesel generators is under review by the NRC.

A Diesel Generator Interim Reliability Program has been proposed to effect improved performance at operating plants. As a consequence, a comparison of' diesel generator outage information within this review is not required.

.... Frankhn Research Center

• :~ oa e % r...w.a .w.

l TER-C5506-276

3. TECHNICAL EVALUATION 3.1 REVIEW OF COMPLETENESS OF THE LICENSEE *S REPORT The ECC systems at RG&E's R. E. Ginna Nuclear Power Plant consist of the folicwing five separate systems: ,' .

o safety injection system -(SIS) .

o chemical volume and control' system (CV'CS)'

o accumulators o residual heat removal (RRR) system o refueling water storage tank (RWST) . - -

In Reference 4, RGGE also included systems and components that support the ECC . systems in carrying out their design . functions under various accident conditions. The support systems are:

o service water system o standby diesel generators o component cooling system.-

In addition to outage data on five ECC systems and the support systems, RG&E also included data on the auxiliary feedwater (AFW) system, containment cooling system,, , core, spra,y (CS) system, and reactor overp2, essure protection system.

c since the AFW system mitigates the consequences of the loss of normal

~

ieedwater, this system is not required to prevent core damage and, therefore, is not considered an ECC system. The containment cooling and core spray systems are designed to reduce containment pressure in the event of a LOCA or sc eam line break,- a'nd the reactor overpressure protection system is designed ti relieve primary system pressure to, prevent exceeding system design pressure.

Trese systems are not required to prevent core damage and, therefore, are not onsidered ECC systems.

For each ECC system outage event, RG&E provided the outage dates, the duration, and the dause, plus sufficient description to discern the corrective action taken. ' Maintenance and surveillance testing activities were included

..d Franklin Research Center

- A cw w n. %e %. -

TER-C5506-276 l in the ECC system outage data, unless these activities were performed during a shutdown condition in which the affecte'd ECC system was 'not required to be operational. The results of RG6E's review were provided for the period from June 1976 through October 1980 for the Ginna Plant.

Based on the preceding discussion, it has been established that RG&E has submitted a report which fulfill's the. requirements of review criterion 1 without exception.

3.2 COMPARISON OF ECC SYSTDi OUTAGES WITH THOSE OF OTHER PLAN'IS The outages of ECC systems can be categorized as (1) unplanned outages due to equipment failure or (2) planned outages due to surveillance testing or preventive maintenance. Unplanned outages me reportable as Licensee Event Re ports (LERs) .under the technical specifications. Planned outages for periodic maintenance and testing are not reportable as LERs. The technical specifications identify the type and quantity of ECC equipment required as well as the maximum all8vable outage times. If an outage exceeds the maximum allowable time, then the plantioperating mode is altered to a lower status consistent with the available ECC system components still operational. The purpose of the technical specification maximum allowable outage times is to prevent extended plant operation without sufficient ECC system protection.

The maximum allowable outage time, specified per event, tends to limit the unavailability of an ECC system. However, there is'no cumulative outage time limitation to prevent repeated planned and unplanned outages from accumulating extensive ECC system downtime.

i Unavailability, as defined in general terms in WASH-1400 [8], is the

_, probability,of,a system being in a failed state when required. However, for this review, a detailed unavailability analysis was' not required. Instead, a l preliminary estimate of the unavailability of an ECC system was madei by 1

( calculating the ratio of the ECC system downtime to the number of. days that the plant was in operetion during the last 5 years. To simplify the tabula-tion of operatimi time, only the period when the plant was in operational Mode j 1 was considered. 'Ihis simplifying assumption is reasonable given that the l period of time that a plant is starting up, shutting down, and cooling down is

.;/.' Franklin Research Center *

• w #n neawunm.

I TER-C5506-276 small compared to the time it is operating at power. In addition, an ECC -

system was considered dow'n wnenever an'ECC system component was unavailable due to any cause. -- -

It should be noted that the ratio calculated .in. this manner is not a true measure of the ECC system unavailhbility, sin'ce' outage events are included that appear to compromise system performance when, in fact, partial or ' full function of the system would be expected. Full" furiction of an ECC system

• would be expected if the design capability of the system exceeded the capacity required for the system to fulfill its sa'fety function. For example, if an ECC system consisting of two loops with multiple pumps in each loop is designed so that only one pump in each loop is required to satisfy core cooling requirements, then an outage of a single pump would not prevent the system from performing its safety function. In addit, ion, the actual ECC system unavailability is a function of planned and unplanned outages of essential support systems as well as of planned and unplanned outages of primary ECC system components. In accordance with the clarificaticq discussed in Section 2, only the effects of outages associated with primary ECC system components and emergency diesel generators are considered in this review. The inclusion of all outage events assumed to be true ECC system outages tends to overestimate the unavailability, while the exclusion of s'upport system outages tends to underestimate the unavailability, of ECC systems and components.

Only a detailed analysis of each ECC system for each plant could improve the confidence in the calculated result. Such an analysis is beyond the intended scope of this report.

The planned and unplanned (forced) outage times for the four ECC systems (SIS, CVCS, and RHR) and the standby diesel generators were identified from

~ ^

the outage information in Reference 4, and are shown in numbers of days and as percentages of plant operating time per year in Table 1 for the.Ginna plant.

The accumulators and RWST are not listed because no outages were reported during the sample period [9) . Outages that occurred during nonoperational periods were eliminated, as were those caused by failures or test and mainte-nance of support systems. Data on plant operating conditions were obtained g '

.... Franklin Research Center ~

ao a.aw re .io u.

i

$5i t:

um "3 .

b a .

• DE Table 1. Planned and Unplanned (Forced) Outage Times for the Ginna Plant

• 1 13 7" ,

3%

ff2 Days of, Plant CVCS Outage in Days

SIS Outage in Days RHR Diesel Generator Outage in Days Outage in Days Year Operation Forced Planned Forced Planned Forced Planned Forced Planned ,

4 1976 213.04 0.0 0.0 0.0 0.0 0.0 0.47 'O . 0 0.0 (0.2%) .

1977 312.08 0.16 0.0 0.28 0.0 '

O.0 ' O.0 ,0 .14' O.0 4

(<0.1% ) (0.1%) , (<0.1%)

1978 294.17 0.0 0.0 0.0 0.0 0.63 ' O.0 0.0 0.0 (0.2%). ,

1979 265.75 0.36 0.0 0.08 0.0 0.05 0.17 0.38 5.52 (0.1%) (<0.1%) (<0.1%) ( <0.11,)

(0.1%) :(2.1%)

1980 280.01 0.50 0.0 "

O.0 0.0 0.0 0.0 1.18 0.39 (0.2% ) ,

(0.4%) (0.1%)

Total 1365.05 1.02 0.00 a 0.36 0.0 0.68 0.64 '1.70 5.91 5 '

(<0.1%) (<0.1%) (<0.1%) (<0.11) (0.1%) .(0.4%) , b o

.

• Numbers in parentneses indicate system outage time as a percentage of total plant operating time. I tj 1

TER-C5506-276 from the annual report , ' Nuclear Pcwer Plant Operating Experience" [10-13),

and from monthly reports,* -Licensed Operating Reactors Status Summary Report" (14]. ne remaining outages wece segregated into planned and unplanned outages based on an interpretation of RG&E's description of the causes. The c

outage periods for each category ,were calcula.te.E by summing the individual outage durations.

Observed outage times of various ECC systems at R. E. Ginna N'ucl' ear Power Plant were compared with those of other PWRs. Based on this comparison, it was concluded that the historical unavail'a bility of the accumulators, RWST, SIS, CVCS, and RHR systems has beEn consistent with the performance of those systems throughout the industry. n e observed unavailability was less than the industrial mean for all five ECC systems, assuming that the underlying unavailabilityis distributed legnormally. W e outage times were also consistent witis' existing technicai specifications.

3.3 REVIEW OF PROPOSED CHANGES TO IMPROVE THE AVAILABILITY OF ECC EQUIIMENT In Reference 4, RG&E did not propose any changes to improve the availability of ECC systems and components.

0 E O e

• e

• • e4 m G

S f

.... Franklin Research Center a % .en.r,wawmu. -

TER-C5506-276

4. CONCLUSIONS Rochester Gas and Electric Company (RG&E) has submitted a report for the Ginna plant that contains (1) outage dates and dur,ations of outages, (2,)

causes .:f the outages, (3) ECC systems or components involved in the outages,

~

and (4) corrective actions taken'. It 'is ' concluded that RGEE has fulfilled the requirements of NUREG-0737, Item II..t.3.17. .- .- .

In addition, the historical unavailability of the ECC systems has been consistent with the performance of those systems throughout the industry. The observed unavailability was less than the . industrial mean for all ECC systems.

The outage times were also consistent with existing technical specifications.

<3

~

...,: Franklin Research Center m anat m r, m va. .

t

r TER-C5506-276

5. REF ERENCES

. .s ., .

1. NUREG-0626

" Generic Evaluation of Feedwater Transiente and Small Break Inss-of-Coolant Accidents in GE-Designed Operating Plants and-

.Near-Term Operating License Applications" NRC, January 1980 - -

2. NUREG-0660 - -

"NRC Action Plan Developed as a Result of the MI-2 Accident" NRC, March 1980

3. NUREG-0737 ..

" Clarification of TMI Action Plan Requiremerits" NRC, October 1980

4. J. E. Maier (RG&E)

Letter to D. M. Crutchfield (NRR, ORB 5)

Subject:

Clarification-of MI Action Plant Requirements (NUREG-0737)

December 30, 1980

5. J. N. Donohew, Jr. (NRC)

Le tter to Dr. S. P. Carf agno (FRC) .

Subject:

Contract No.

NRC-0 3-81-13 0, Tentative Assignment F July 21, 1981

6. NRC Meeting between NRC and FRC.

Subject:

C5506 Tentative. Work Assignment F, Operating Reactor PORV and ECCS Outage Reports August 12, 1981

7. NRC Meeting between NRC and FRC.

Subject:

Resolution of Revicw Criteria and Scope of Work July 26, 1982 8 WASH-1400

" Reactor . Safety Studv"

_. N.RC ,, tober 1975

9. NRC Telephone Conversation Between B. Ludington (FRC) and'J. Lyons (NRC) .

Subject:

Licensee Gubmittal for II.K.3.17

- July 17, 1982

10. 'NUREG-0366-

" Nuclear Power Plant Operating Experience 1976" NRC, December 1977 B~

$s Franklin Research Center

..i

.- 4om aw wr,.nw w - -

j

[ TER-C5506-276 l

11. NUREG-0483

" Nuclear Power Plant Operating Experience 1977" ,

NRC, February 1979 ,

12. NUREG-0618

"!Alclear Power Plant Operating Experience 1978" .

NRC, December 1979 . .

13. NUREG/CR-1496 .

" Nuclear Power Plant Operating Experience.1979" NRC, May 1981 ,

14. NUREG-0020

" Licensed Operating Reactors Status Summary Report" Volume 4, Nos.1 through 12, and.Vo.lume 5, No.1 NRC, December 1980 through January 1981 l

9

.... Franklin Research Center A W of N Frenen arwomme

_ _ _ . _ . _ _ _ . . _ _ _ . . _ . . . - _ _ . _