ML20023A503
| ML20023A503 | |
| Person / Time | |
|---|---|
| Site: | Big Rock Point File:Consumers Energy icon.png |
| Issue date: | 09/28/1982 |
| From: | Ludington B, Overbeck G, Vosbury F Franklin Research Ctr, Franklin Institute |
| To: | Chow E Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20023A504 | List: |
| References | |
| CON-NRC-03-81-130, CON-NRC-3-81-130, RTR-NUREG-0737, RTR-NUREG-737 TER-C5506-170, NUDOCS 8210010152 | |
| Download: ML20023A503 (15) | |
Text
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TECHNICAL EVALUATION REPORT
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ECCS REPORTS (F-47)
TMI ACTION PLAN REQUIREMENTS i
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i CONSUMERS POWER COMPANY BIG ROCK POINT PLANT NRC DOCKET NO. 50-135 FRC PROJECT C5%6 i
FRC ASSIGNMENT 7 I
NRC CONTRACT NO. NRC-03-81 130 FRC TASK 170 Preparedby F. W. Vosbury Franklin Res3&rch Center Author:
G. J. Overbeck 20th and Race Streets B. Ludington Philadelphia, PA 19103 FRC Group Leader:
G. J. Overbeck Prepared for
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Nuclear Regulatory Commission Lead NRC Engineer:
E. Chow l
Washir'gton, D.C. 20555 Sepcember 28, 1982 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 l
employees, makes any warranty, expressed or implied, or astumes any legal liability or l
respensibility for any third party's use, or the results of such use, of any information, appa-l ratus, product or process disclosed in this report, or represents that its use by such third party would not infringe privately owned rights.
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Prepared by:
Reviewed by:
Approved by:
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TER-C5506-170 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
REVIER CRITERIA.
3 3
TECHNICAL EVALUATION 4
i 3.1 Review of Completeness of the Licensee's Report 4
3.2 Comparison of ECC System Outages With Those of Other Plants 6
3.3 Review of Proposed Changes to Improve the Availability of ECC Equipmer.t 9
4 CONCLUSIONS.
10 5
REFERENCES.
11 4
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%c FORENORD s
This 'hchnical Evaluation Report was prepared by Franklin research Center under a contract with the U.S. Nuclear Regulatory Cormaission (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. B. W. Ludington, and Mr. F. W. Vosbury contributed to the technical preparation of this report through a subcontract with WESTEC Services, Inc.
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TER-C5506-170 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 Consumers Power Company's (CPC) Big Rock Point Plant. Se purpose of this evaluation is to determine if the Licensee has submitted a report that is complete and satisfies the requirements of TMI Action Item II.K.3.17, " Report on Outages of Emergency Core-Cooling Systems Licensee Report and Proposed hchnical Specification Changes."
1.2 GENERIC BACKGROUND Following the 2ree 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 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 permitted plant operation for substantial periods with the HPCI system out ?f 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. m erefore, to ensure that the small break LOCA 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 Smal' Break 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:
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" 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), he outage data for ECC components should include all outages for-tne last five years of Jc. Franklin Research Center a cm a or n. r em.
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4 TER-C5506-170 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 intc' NUREG-0660 (2), "NRC Action Plan Developed as a Result of the TMI-2 Accident," for all light water reactor plants as TMI Action Item II.K.3.17.
In NUREG-0737 [3], " Clarification of TMI Action Plan Requirements," the NRC staff added a requirement that licensees propose changes 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:
"The report should contain (1) outage dates and duration of outages; (2) cause of the outage; (3) ECC systems or components involved in the outage; and (4) corrective action taken."
1.3 PLANT-SPECIFIC BACKGROUND y
On December 19, 1980, CPC submitted a report [4] in response to
- NUPEG-0737, Item II.K.3.17, " Report on Outages of Emergency Core-Cooling Systems Licensee Report and Proposed Technical Specification Changes." The i
report submitted by CPC covered the period from November 1,1975 to November I
i 1,1980 for the Big Rock Point Plant.
In their report, CPC did not tauggest any future modificationa to improve the availability of ICC systems, noting instead several improvements which had already been made.
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TER-C550 6-170 2.
REVIEW 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 [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 laut 5 years of operation. This 4
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 outagers.
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 cutages 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 [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 impro'ted l
reliability of diesel generators is under review by the NRC.
A Diesel j
Generator Interim Reliability Prgram has been proposed to effect improved l
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-C5506-170 3.
TECHNICAL EVALUATION 3.1 REVIEW OF COMPLETENESS OF THE LICENSEE'S REPORT Thc final hazards summary report (FHSR) for the ' Big Rock Point Plant stated that the core spray system is provided to prevent a core meltdown should the core become uncovered following an incident. Water from the fire' protection system is admitted into a circular sparger above the core which directs spray onto the fuel elements. Water can also be directed to redundant core spray nozzles above the core. The fire protection water is provided by motor-and diesel-driven fire pumps taking suction from Lake Michigan. After the water level in the containment vessel reaches elevation 587 f t, the core j
spray recirculation system (sometimes referred to as the post-incident recirculation system) is started and the water supply from Lake Michigan to the spray nozzles is isolated. The core spray recirculation system consists of two core spray pumps, a core spray heat exchanger, and piping and valves.
The core spray pumps take suction from the containment vessel and discharge through a heat exchanger to the core spray sparger. The cooling water for the heat exchanger is supplied by the fire protection system.
The reactor depressurization system at the Big Ibck Point plant is provided to rapidly reduce the pressure of the primary system during LOCA conditions to allow the core spray system to put water into the reactor vessel and keep the core cooled. Since the core spray system has an operating l
l pressure of approximately 150 psig (from the fire pumps), it is essential to reduce the primary system pressure below t.his point to allow the core spray
' system to function properly.
In summary, the ECC systems at CPC's Big Rock Point Plant consist of the following three systems:
o core spray system (CS$)
o reactor depressurization system (RDS) o core spray recirculation system (CSRS).
In Reference 4, CPC also included data on diesel generators.
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e TER-C5506-170 In establishing unavailabilities, CPC considered only periods when the systems were required by the Technical Specifications to be operable for emergency core cooling. CPC did not include outages attributed to te;tting of i
protection system instrumentation and controls in which the degree of redundancy in the channels is greater than one. An example cited by CPC was a sensor channel bypassed for troubleshooting or repair on the reactor depressurizing system. The logic circuitry automatically reverts to a two out of three process instead of two out of four; however, the complete system is still capable of performing the required safety function. The logic circuitry as described is consistent with the requirements of IEEE Std 279-1971 for channel bypass or removal from operation. Specifically, a loss of function should not result from testing even if a single failure exists in one of the redundant channels which are not under test. CPC ee1"=W of. outages due to. testing of protection system instrumentation and controls as described is consistent with the intent of h"JREG-0737, Item II.K.3.17, in that the protection system is available to perform its safety function.
CPC also did not include instances which resulted in degradation of one of the RDS battery power supplies. CPC analysis concluded that the capability of the RDS to perform its safety function was maintained. However, events in which one or more battery cells were replaced were included, because one loop out of four was affected and the three unaffected loops must opertate without single f ailure to meet the safety function requirements.
i The ECC system outage data were extracted from the following plant records:
o operations logs o maintenance orders o Licensee Event Reports.
For each ECC system outage event involving forced or preventive maintenance outages, CPC provided the outage dates, the durations, the causes, and sufficient description to indicate the corrective action taken. Outages for surveillance testing activities were reported as estimated total times for the 5-year period for each system. The results of CPC's review were provided for the period from November 1,1975 to November 1,1980 for the Big Itck Point plant.
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TER-C550 6-17 0 On the basis of the preceding discussion, it has been established that the
.CPC report fulfills the requirements of review criterion 1 without exception.
3.2 COMPARISON OF ECC SYSTEM OUTAGES WITH THOSE OF OTHER PLANTS The outages of ECC systems can be categorized as (1) unplanned outages due to equipment failure on (2) planned outages due to surveillance testing or preventive maintenance. Unplanned outages are reportable as Licensee Event Reports (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 allowable outage times.
If an outage exceeds the maximum allowable time, then the plant operating 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, spgcified 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.
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 preliminary estimate of the unavailability of an ECC system was made b 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 I
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 cooling down is small compared to the time it is operating at power. In addition, an ECC j
system was considered down whenever an ECC system component was unavailable due to any cause, l
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4 TER-C5506-170 It should be noted that the ratio calculated in this, manner is not a true measure of the FCC 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 e.wpected if tne design capability of the system exceeded the capacity required for the system to fulfill its safety function. Pbr example, if cn ECC system consisting of two loops with multiple pumps in each loop is designed en 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 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 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. Se 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 th'e three ECC systems (RDS, CSS, and CSRS) and the emergency diesel generators were identified from the outage information in Reference 4 and are shown in number of days and as percentage of plant operating time per year in Table 1 for the Big Rock Point Plet. Outages that occurred during nonoperational periods were eliminated, as well as 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 Experience" [9-12], and from monthly reports,
" Licensed Operating Reactors Status Summary Reports" (13]. The remaining outages were segregated into planned and unplanned outages based on an interpretation of CPC's description of tne causes. We outage periods for each category were calculated by summing the individual outage durations.
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Planned and Unplanned (Forced) Outage Times for Big Rock Point Plant
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Core Spray Recirculation RDS Diesel Generator
,3 Days of Plant Outage in Days Outage in Days Outage in Days Outage in Days lU Year Operation Forced Planned Forced Planned Forced Planned Forced Planned E
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[Q 1976 183.54 0.0 0.13 0.0 0.50 0.17 0.0 2.89 0.69 8{
(0.1%)
(0.3%)
(0.1%)
(1.6%)
(0.4%)
I 1977 267.96 0.0 0.13 0.0 0.50 0.58 0.0 1.12 0.05
(<0.1%)
(0.2%)
(0.2%)
(0.4%) (<0.1%)
1978 284.13 0.02 0.13 0.0 0.50 3.99 0.29 1.90 0.67
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(<0.1%) (<0.1%)
(0.2%)
(1.4%)
(0.1%)
(0.7%)
(0.2%)
1979 85.96 0.0 0.13 0.0 0.50 0.0 0.0 0.0 0.04 I
(0.2 % )
(0.6%)
(<0.1%)
I 1980 299.00 0.05 0.13 0.0 0.50 3.17 0.0 0.21 0.38
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(<0.1%) (<0.16)
(0.2 %)
(1.1%)
(0.1%)
(0.1%)
i Total 1111.59 0.07 0.65 0.0 2.5 7.91 0.29 6.12 1.83
(<0.1%) (<0.11)
(0.2 %)
(0.7%) (<0.1%)
(0.6%)
(0.2 %)
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- Percentages in parentheses are ratios of outage times to total operating times.
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TER-C5506-170 Big Rock Point Plant is an older plant, and only the core spray portion of the ECC system is comparable to systems in other plants. The observed outage j
time of the core spray system was compared with those of other BWRs. Based on this comparison, it was concluded that the historical unavailability of the core spray system has been consistent with the performance of those systems throughout the industry. The observed unavailability was less that the industrial mean, assuming that the underlying unavailability is distributed lognormally.
The other two ECC systems, RDS and recirculat' ion, are not directly comparable to systems in other plants. However, the observed unavailability for both systems is comparable to unavailability observed in other types of ECC systems. All outage times were consistent with existing technical specifications, n e outages of the emergency diesel generators were not included in this comparison.
3.3 REVIEW OF PROPOSED CHANGES TO IMPROVE THE AVAILABILITY OF ECC EQUIPMENT In Reference 4, CPC indicated that technical specification changes 'to impose cumulative outage limitations were not necessary. CPC identified several modifications to ECC equipmerit already in place and plans to continue improvement. Changes already completed within the RDS include improvements in the automatic test pulse circuitry, battery cell replacement, and an increase in the battery float voltage. Replacement of Icop C RDS battery cells is scheduled. CPC also cited improvements to the emergenty diesel generators I
which included modifications to the cooling water pump seal, several changes associated with the governor, and use of premium diesel fuel.
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TER-C5506-170 4.
CONCLUSIONS Conscmers Power Company (CPC) has submitted a report for the Big Rock Point Plant that contains (1) outage dates and durations of outages, (2) causes of the outages, (3) ECC systems or components involved in the outages, and (4) correc-tive actions taken. It is concluded that CPC has fulfilled the requirements of NUREG-0737, Item II.K.3.17.
In addition, the histcrical innavailability of the core spray system has been consistent with the performance of those systems throughout the industry. The observed unavailability was less than the industrial mean. The obsserved unavailability of the RDS and recirculation systems are comparable to unavailability observed in other types of ECC systems. All outage times were consistent with existing technical specifications.
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REFERENCES 1.
NUREG-0 626
" Generic Evaluation of Feedwater Transients and Small Break Ioss-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 'IMI-2 Accident" NRC, March 1980 3.
" Clarification of TMI Action Plan Requirements" NRC, October 1980 4.
D. P. Hoffman (CPC)
Letter to D. M. Crutchfield, (Operating Reactor Branch No. 5).
Subject:
Response to NUREG-0737 December 19, 1980 5.
J. N. Donohew, Jr. (NRC)
Letter to Dr. S. P. Carfagno (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 Review Criteria and Scope of Work July 26, 1982 i
8.
WASH-1400 l
" Reactor Safety Study" NRC, October 1975 9.
" Nuclear Power Plant Operating Experience 1976" NRC, December 1977 10.
" Nuclear Power Plant Operating Experience 1977" NRC, February 1979
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" Nuclear Power Plant Operating Experience 1978" NRC, December 1779 12.
" Nuclear Power Plant Operating Experience '1979" NRC, May 1981 13.
" Licensed Operating Reactors Status Surmnary Report" Volume 4, Nos.1 through 12, December 1923, and Volume 5, No.1, January 1981 i
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