ML20213E506

Advises That TMI Action Item II.K.3.17, Rept on Outages of ECCS, Closed Out for Listed Plants.Model Ltr & Safety Evaluation to Accomplish Closeout Encl

ML20213E506
Person / Time
Site:	Rancho Seco
Issue date:	07/28/1983
From:	Rooney V Office of Nuclear Reactor Regulation
To:	Stolz J Office of Nuclear Reactor Regulation
References
TASK-2.K.3.17, TASK-TM TAC-45662, NUDOCS 8308100384
Download: ML20213E506 (5)

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MEMORANDUM FOR: John F. Stolz, Chief Operating Reactors Branch #4, DL Robert A. Clark, Chief Operating Reactors Branch #3, DL Steven A. Varga, Chief Operating Reactors Branch #1, DL Dennis M. Crutchfield, Chief Operating Reactors Branch #5, DL W/[h THRU:

Domenic B. Vassallo, Chief Operating Reactors Branch #2, DL FROM:

Vernon L. Rooney, Project Manager Operating Reactors Branch #2, DL

SUBJECT:

CLOSE00T OF TMI ITEM II.K.3.17, REPORT ON OUTAGES OF ECC SYSTEMS lists plants for which Item II.K.3.17 can now be closed out. contains a model letter and safety evaluation which can be used to accomplish closecut by simply making the letter and safety evaluation plant specific. Also enclosed are Franklin Research Center TERs to be attached as enclosures to the SEs for the plants in your branch.

Please ask the PMs in your branch to issue these closeout letters as soon as

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possible.

If there are any questions, please contact Vern Rooney, ext. 28286.

k'l w VernonL.Rooney,'f

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Project Manager Operating Reactors Branch #2 Division of Licensing

Enclosures:

As stated k

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. ENCLOSURE 1 l

'o OperatingPlantsReviewedUnderII.K.3.17

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ANO-l ANO-2 Beaver Vally 1 Big Rock Point

• Browns Ferry 1, 2 & 3 Calvert Cliffs 1 & 2 Cook 1 & 2 Cooper Davis Besse 1 Dresden 2 & 3

• Duane Arnold Farley 1 Fitzpatrick Ft. Calhoun Ginna Haddam Neck

• Hatch.1 & 2 Indian Point 3 Kewaunee t

Lacrosse Maine Yankee Millstone 1 Millstone 2 Monticello Nine Mile Point North Anna 1

• 0yster Creek Palisades

-* Peach Bottom 2 &,3 Pilgrim

• Point Beach 1 & 2 Prairie Island 1 & 2 Quad Cities 1 & 2 Rancho Seco Robinson 2 Salem 1 San Onofre 1 St. Lucie 1 Surry 1 & 2 Trojan

• Turkey Point 3 & 4 Vermont Yankee Yankee Rowe Zion 1 & 2

• Indicates that additional in ' depth review was performed on' the ECC5 outages.

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4 jof UNITED STATES NUCLEAR REGULATORY COMMISSION e

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E WASHINGTON, D. C. 20555

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Docket No.' 50-(Licensee)

SUBJECT:

REVIEW OF TMI ITEM II.K.3.17, REPORT ON OUTAGES OF ECC SYSTEMS (Plant Name)

The TMI Action Item II.K.3.17 required that licensees submit a report detailing outage dates and length of outages for all ECC systems for the last five years of operation. We have completed our review of your submittal and a copy of our Safety Evaluation is enclosed for your information.

We have concluded that the requirements of NUREG-0737, Item II.K.3.17 have been met. Therefore, this completes our review of TMI Action Item II.K.3.17 for your facility.

Sincerely, Branch Chief Operating Reactors Branch #

Division of Licensing

Enclosure:

As stated cc w/ enclosure:

See next page

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UNITED STATES NUCLEAR REGULATORY COMMISSION T

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SAFETY EVALUATION BY THE OFFICE OF NUCLEAR REACTOR REGULATION (PLANT NAME)

TMI-ACTION PLAN ITEM II.K.3.17 REPORT ON OUTAGES OF EERGENCY CORE COOLING SYSTEMS Introduction The TMI Action Item II.K.3.17 states that the licensees (of all light water reactors) should submit a report detailing outage dates and lengths of outages for all ECC systems for the last five years of operation. The report should also include the causes of the outages. The clarification of this requirement states that the infomation provided will be used by the staff to determine if a need exists for cumulative outage ' time requirements in the technical speciff.

cations, and also states that licensees.should propose technical specifications or changes to improve availability of ECCS equipment if needed.

Evaluation The licensee's report has been reviewed by our consultant, Franklin Research Center (FRC), under a technical assistance contract. FRC has compared the 11-censee's historical unavailability of ECCS eauipment with perfomance throughout l

the industry.

A. copy of FRC's Technical Evaluation Report is enclosed. Based' on the reports from all light water reactors, FRC has developed a characteri-zation of ECC system unavailability for the entire industry. FRC then compared-the ECC system unavailability for individual plants with the average for all plants. FRC has concluded that this licensee has met the requirements of Item II.K.3.17. We agree with this conclusion.

We have considered the results of the FRC review in order to detemine the'need for cumulative outage. time technical specifications. We have not detemined definitely whether there is need for a cumulative outage time requirement in the technical specifications. The detemination of any need for modification of allowed ECCS equipment outage periods should be.most rationally based on the risk reduction produced by a change to allowed ECCS equipment outage periods in the technical specifications, together with the impacts produced by the change.

These considerations are part of a generic technical activity (B-61) and will be pursued separately by the NRC staff.

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f However, we have attempted to determine, on an interim basis, whether there is a need for a cumulative outage time requirement, by" comparing the ECCS unavailability of a particular plant to the average of that of all plants.

If the ECCS unavailability of a particular plant did not significantly' exceed the average, then we considered it acceptable, and did not require modifica-tions to the technical specifications. If, on the other hand, a plant exhibited a cumulative ECCS outage time appreciably in excess of the average, it was looked at more closely.

Conclusion We conclude that for this plant the requirements of NUREG-0737, Item II.K.3.17 have been met. We further conclude that for this plant there is no need for cumulative outage time technical specifications at this time.

Principal Contributor:

E. Chow Dated:

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I TECHNICAL EVALUATION REPORT ECCS REPORTS (F-47)

T: 41 ACTION PLAN REQUIREMENTS SACRAMENTO MUNICIPAL UTILITY DISTRICT RANCHO SECO NUCLEAR GENERATING STATION NRC OOCKET NO. 50-312 FRC PROJECT C5506 FRC ASSIGNMENT 7 NRC CONTRACT NO. NRC-03-81-130 FRC TASK 292 Prepared by y,_y, yogyg7, Franklin Researcts Center Author:

G. J. Overbeck 20th and Race Streets B. W. Ludington Philadelphia.PA 19103 FRC Group Leader:

G. J. Overbeck Prepared for Nuclear Regulatory Commission Lead NRC Engineer:

E. Chow Washington, D.C. 20555 January 27, 1983 This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, or any of their employees, makes any warranty, expressed or implied, or assumes any legal liability or responsibility for any third party's use, or the results of such use, of any information, appa-ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.

Prepared by:

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Department Olrfcted Date: ' /> ~b 7 Date:

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TER-C5506-292 CONTENTS Se ction Title Page

.. 1 1

INTRODUCTION 1.1 Eurpose of Review.

1 1

1.2 Generic Background.

1.3 Plant-Specific Background.

2 2

REVIEW CRITERIA.

3 3

TECHNICAL EVALUATION 4

3.1 noview of Complet=a=== of the I.icensee'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!CLUSIONS.

9 5

REFERENCES.

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TER-C5506-292 FOREWORD This Technical Evaluation Report was prepared by Franklin Research Center under a contract with the U.S. Nuclear Regulatory Commission (Office of Nuclear Reactor Regulation, Division of Operating Reactors) for technical 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.

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3 TER-C5506-292 1

1.

INTRODUCTION 1.1 PURPOSE OF REVIEN This technical evaluation report (TER) documents an independent review of the outages of the emergency core cooling (ECC) systems at Sacramento Etnicipal Utility District's (SMUD) Rancho Seco Nuclear Generating Station.

The purpose of this evaluation is to determine if the Licer;sce has submitted a report that is complete and satisfies the requirements of TML Action Item II.K.3.17, " Report on. Outages of Emergency Core-Cooling Systems Licensee

)

Report and Proposed Technical Specification Changes."

1.2 GENERIC BACKGROUtc Tullowing the Three Mile Taland' Unit 2 accident, the Bulletins and Orders Task Force reviewed nuclear steam supply system (NSSS) vendors' small break i

loss-of-coolant accident (IOCA) analyses to ensure that an adequate basis existed for developing guidelines for small break LOCA emergency procedures.

During these reviews, a concern 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) system as the worst single failure. However, the technical specifications j

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 where substantial outages might occur within the limits of the present technical specification.

Therefore, to ensure that the small break IDCA analyses are consistent with the actual plant response, the Bulletin and Orders Task Force recommended in NUREG-0626 (1], " Generic Evaluation of Feedwater Transients and Small Break j

Loss-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:

" Submit a report detailing outage dates and lengths of the outages for all ECC systems. The report should also include the cause of the outage (e.g., controller failure or spurious isolation). The outage data for i

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TER-C5506-292 i

ECC components should include all outages for the last five years of operation. The end result should be the quantification of historical unreliability due to test and maintenance outages. This will establish if a need exists for cumulative outage requirements in technical specifications."

Later the recommendation was incorporated into NUREG-C460 [2], "NBC Action Plan Developed as a Result of the TMI-2 Accident," for GE-designed NSSSs as TMI Action Item II.K.3.17.

In NUREG-0737 [3], " Clarification of TMI Action j

Plan Requirements," the NBC staff expanded the Action Item to include all light water reactor plants and added a requirement that licen.;ees propose chang'es that will improve and control availability of ECC systems and components. In addition, the contents of the reports to be submitted by the licensees were further clarified as follows:

. $a.raport.nhem1d.contain 41). outage dates _and. duration 4 outagess l

(2) cause of the outager (5) ECC systems or components involved in the outage; and (4) corrective action taken."

1.3 PLANT-SPECIFIC BACKGROUE On January 16, 1981 [4], SED submitted a report in response to NUREG-0737, Item II.K.3.17, " Report on Outages of Baergency Core-Cooling l

Systems Licensee Report and Proposed Technical Specification Changes." The j

report submitted by SED covered the period from January 1,1976 to December 31, 1980 forf Rancho Seco Nuclear Generating Station. On November 28, 1982

[5], SED submitted a second report in response to an NBC request for addi-tional information on ECCS outages. S e second repor* covered the same period 1

described in the first report. SMUD did not provide any recommendations to improve and control availability of ECC systems.

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1 TER-C5506-292 2.

REVIEN CRITERIA The Licensee's response to NUREG-0737, Item II.K.3.17, was evaluated against criteria provided by the NRC in a letter dated July 21, 1981 [6]

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 [7]. Auxiliary systems such as component cooling water and plant service water systems were not to be considered in determining the una'vailarility of ECC systems. Only the outages of the diesel generators were to be included along with the primary ECC system outages. Finally, the "last five years of operation" was to be loosely interpreted as a continuous 5-year period of recent operation.

On July 26, 1982 [8], 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 improved 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.

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TER-C550 6-292 3.

TECHNICAL EVALUATION 3.1 REVIEW OF COMPLETENESS OF THE LICENSEE'S REPORT The ECC systems at SED's Rancho Seco Nuclear Generating Station consist of the following four separate systems:

o core flood tanks o high pressure injection (HPI) o low pressure injection (LPI) o borated water storage tanks (BWST).

In References 4 and 5, SED also included the emergency diesel generators, reactor building spray system, auxiliary feedwater system, emergency cooling units of the reactor building emergency cooling system, the nuclear service cooling system, and the nuclear service raw water system. ~The reactor building spray system sprays borated water into the reactor building atmosphere. The spray cools the atmosphere and reduces the post-accident temperatures and pressure within the building. It also removes fission products from the building atmosphere. The auxiliary feedwafer system provides feedwater to the steam generators upon loss of normal feedwater. The reactor building, emergency cooling system circulates the reactor building air and provides for the removal of heat released during an accident. The nuclear service cooling water system is a closed loop system, serving as a medium to transfer heat from the decay heat removal coolers and the reactor building emergency cooling units to the nuclear service raw water system. The nuclear service raw water systems provides a heat sink for the decay heat removal system, reactor building emergency cooling system, diesel generator cooling system, and safety system pump oil coolers, and pump room coolers in the event of a loss-of-coolant accident (LOCA). Because none of these systems are Primary BCC systems, they were not considered in this review.

l For each BCC system outage, SMUD provided the date, the duration, a brief description, and the cause, with sufficient details to indicate the corrective 1

action taken. Information on routine preventive maintenance and surveillance 1

testing was also included.

nklin Research Center A Dewesen of The Fransen Insseuse

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i TER-C550 6-292 SMUD's review encompassed the period from January 1,1976 to December 31, 1980 for Rancho Seco Nuclear Generating Station.

Based on the preceding discussion, it has been established that SMUD has submitted a report which fulfills the requirements of review criterion 1 without exception.

3.2 COMPARISON OF ECC SYSTEM OUTAGES WITH THOSE OF OTHER PLANTS I

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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 l

preventive maintenance. Unplanned outages are reportable as Licensee Event Reports (LERs) under the technical specifications. Planned outages for f

periodic maintenance and testing are not reportable as LERs. The technical specifications identify the type and quantity of ECC equipment required as j

well as the maximum allowable outage times.

If an outage exceeds the maximum i

allowable time, then the plant operating mode is altered to a lower status consistent with the available BCC system components still operational. The

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purpose of the technical specification maximum allowable outage times is to l

i prevent extended plant operation without sufficient ECC system protection.

The maximum allowable outage time, specified per event, tends to limit the i

unavailability of an ECC system. However, there is no cumulative outage time limitation to prevent repeated planned and unplanned outages from accumulating j

extensive ECC system downtime.

Unavailability, as defined in general terms in NASH-1400 [9], 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

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preliminary estimate of the unavailability of an ECC system was made by l

calculating the ratio of the ECC system downtime to the number of days that the plant was in operation during the last 5 years. To simplify the tabulation of operating time, only the period when the plant was in operational Mode 1 was considered. This simplifying assumption is reasonable given that the period of time that a plant is starting up, shutting down, and 4

cooling down is small compared to the time it is operating at power. In i

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TER-C550 6-292 addition, an ECC system was considered down 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 unavailability, since outage events are included that appear to compromise system performance when, in fact, partial or full function of the system would be expected.

Full function of an ECC system would be expected if the design capability of the system exceeded the capacity required for the system to fulfill its safety function. For example, if an ECC system consisting of two loops with multiple pumps in each loop is l

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 addition, the actual ECC system unavailability is t function of planned and unplanned outages of j

essential support systems as well as of planned and unplanned outages of primary BCC system components.

In accordance with the clarification discussed in Section 2, only the effects of outages associated with primary ECC system components and emergency diesel generators are considered in this review. 'Ihe inclusion of all outage events assumed to be true ECC system outages tends to overestimate the unavailability, while the exclusion of support 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 ECC systems (core flood tanks, HPI, LPI, and BWST) and the emergency diesel generators were I

identified from the outage information in References 4 and 5 and are shown in i

number of days and as percentage of plant operating time per year in Table 1 I

for Rancho Seco Station. Outages that occurred during non-operational periods were eliminated, as were those caused by failures or test and maintenance of support systems. Data on plant operating conditions were obtained from the annual reports, " Nuclear Power Plant Operating Experit.nce" [10-13], and from monthly reports, " Licensed Operating Reactors Status Summary Reports" [14].

l The remaining outages were segregated into planned and unplanned outages on U00 ranklin Research Center A Omman of The Fransen inseawee

Table 1.

Planned and tmplanned (Forced) Outage Times for Rancho Seco Nuclear Generating Station

• Core Flood Tanks HPI LPI BWST Diesel Generator Days of Plant Outage in Days Outage in Days Outage in Days Outage in Days Outage in Days Year Operation Forced Planned Forced Planned Forced Planned Forced Planned Forced Planned 1976 110.9 0.0 0.0 0.0 0.3 0.0 0.8 0.0 0.0 2.1 0.8 (0.3%)

(1.9%)

(0.7%)

1977 281.4 0.0 0.0 0.8 0.7 0.0 0.8 0.0 0.04 0.2 3.3 (0.3%)

(0. 2 % )

(0.3%)

(<0.1%)

(0.1% )

(1.2%)

b 1978 29 3.6 0.0 0.0 0.0 0.6 0.0 1.0 0.0 0.0 0.3 20.3 8

(0.2%)

(0.3%)

(0.1%)

(6.9%)

1979 282.4 0.0 0.0 24.3 1.0 1.0 0.2 0.0-0.0 1.2 6.6 (8.68)

(0.4 %)

(0.4 %)

(0.1%)

(0.4 %)

(2.3%)

1980 226.0 0.0 0.0 2.0 0.5 0.2 0.5 0.0 0.0 3.8 0.7 (0.9%)

(0.2 %)

(0.14 )

(0.2 %)

(1.7%)

(0. 3 % )

Total 1194.3 0.0 0.0 27.1 3.1 1.2 2.5 0.0 0.04 7.6 31.7 g

(2.3%)

(0.3 % )

(0.1%)

(0.2%)

(<0.1%)

(0.6 % )

(2.7%)

=

o

• Numbers in parentheses indicate system outage time as a percentage of total plant operating time.

g

.I f

I

)

TER-C5506-292 i

1 1

the basis of SMUD's description of the cause. The outage periods for each category were calculated by summing the individual outage durations.

)

Observed outage times of various ECC systems at Rancho Seco Station were compared with those of other PWRs. Based on this comparison, it was concluded 2

that the historical unavailability of the core flood tanks, LPI, and BWST has been consistent with the performance of those systems throughout the industry and consistent with existing technical specifications. The observed unavail-ability was less than the industrial mean for the core flood tanks, LPI, and BWST. The HPI system, however, has an observed unavailability significantly higher than that observed in other plants and has exceeded the industrial mean by greater than about one standard deviation, assuming that the underlying l

unavailability is distributed lognormally. The outages of the diesel generators were not included in this comparison.

A review of HPI system outages ' revealed that one outage of 24.3 days comprised 81% of the total outage time. This outage, documented in LER 79-24

[4], started on December 17, 1979 and lasted until January 9,1980. The outage was due to operator error and resulted in improper HPI valve line up.

l This was the only event of this type reported. The unavailability due to the remaining outage events is* less than the industrial mean for systems of this type.

4 3.3 REVIEW OF PROPOSED CHANGES TO IMPROVE THE AVAILABILITY OF ECC EQUIPMENT l

In References 4 and 5, SMUD did not propose any changes to improve the availability of ECC systems and components. This review revealed extensive HPI unavailability; however, as 81% of the unavailability resulted from a I

single isolated event, no changes to equipment or procedures are recommended.

A dhranklin Research Center A Ohemen of The Fransen insense

_,,-._._.....--=-,,,.- _ -,.

=-__ _.~._,-.-,. _,-,_ _,- _ _,, _-- _ -,

i TER-C5506-292 4.

CONCLUSIONS Sacramento Municipal Utility District (SED) has submitted a report for Rancho Seco Nuclear Generating Station that contains (1) outage dates and durations (2) causes of the outages, (3) ECC systems or components involved in the outages, and (4) corrective actions taken.

It is concluded that SMD has i

fulfilled the requirements of NUREG-0737, Item II.K.3.17.

In addition, the historical unavailability of the core flood tanks, low pressure injection and borated water storage tank systems has been consistent.with the performance of those systems throughout the industry and consistent with existing technical specifications. The observed unavailability was less than the industrial mean for the core flood tanks, low pressure injection, and borated water storage tank. However, the observed unavailability of the high pressure injection system was significantly higher than the industrial mean. The high unavail-ability was due to a single isolated event and no changes to equipment or procedures are recommended.

l

}

.5 s

. UOOu Frenidin Research Center A cnw a w n. r, men

TER-C5506-292 i

5.

REFERENCES 4

1.

NUREG-0626

" Generic Evaluation of Feedwater Transients and Small Break Loss-of-Coolant Accidents in GE-Designed Operating Plants and Near-Tern Operating License Applications" NBC, January 1980 2.

NUREG-0660 "NBC Action Plan Developed as a Result of the TMI-2 Accident" NRC, March 1980 l

3.

NUREG-0737 i

" Clarification of TMI Action Plan Requirements" NRC, October 1980 i

4.

J. J. Mattimoe (SED) i

. lestar. to A _W.. Esid..(Chief, Operating Jteactnes, Branch 4, NRC) l

Subject:

Submittal of Information Required by NUREG-0737 SED, January 16, 1981 5.

J. J. Mattimoe (SNUD)

Letter to J. F. Stolz (Chief, Operating Reactors, Branctn,4, NRC)

Subject:

Item II.K.3.17, Request for Additional Information SED, November 28, 1982 6.

J. N. Danohew, Jr. (NBC) i Letter to Dr. S. P. Carfagno (FRC).

Subject:

Contract No.

NBC-03-81-13 0, Tentative Assignment F NRC, July 21, 1981 l

7.

NBC l

Meeting between NRC and FRC.

Subject:

C5506 Tentative Work Assignment F, Operating Reactor PORV and ECCS Outage Reports

}

August 12, 1981 i

8.

NRC Meeting between NRC and FRC.

Subject:

Resolution of Review Criteria l

and Scope of Work July 26, 1982 9.

WASH-1400

" Reactor Safety Study" e

NRC, October 1975 10.

NUREG-0366 "Itaclear Power Plant Operating Experience 1976" l

NBC, December 1977 nklin Research Center j

A Ohemen of The Frerden Inustuse I

TER-C5506-292 11.

NUPEG-0483 "melear Power Plant Operating Experience 1977" NBC, February 1979 4

12.

NUREG-0618

" Nuclear Power Plant Operating Experience 1978" NRC, December 1979 13.

NUREG/CR-1496 "W elear Power Plant Operating Experience 1979" NRC, May 1981 14.

NUREG-0020

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

i nklin Research Center A DMaan of The Frenedm keen no

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