ML20058G171

Forwards Prioritization Evaluation & Gimcs Info Sheets Re GI 165 Spring-Actuated Safety & Relief Value Reliability

ML20058G171
Person / Time
Issue date:	11/26/1993
From:	Beckjord E NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
To:	Minners W NRC OFFICE OF NUCLEAR REGULATORY RESEARCH (RES)
References
REF-GTECI-***, REF-GTECI-MI, REF-GTECI-NI, REF-GTECI-SC NUDOCS 9312090116
Download: ML20058G171 (12)

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UNITED STATES

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l NOV 2 61993 i

MEMORANDUM FOR:

Warren Minners, Director Division of Safety Issue Resolution Office of Nuclear Regulatory Research FROM:

Eric S. Beckjord, Director Office of Nuclear Regulatory Research

SUBJECT:

GENERIC ISSUE NO.165, " SPRING-ACTUATED SAFETY AND RELIEF VALVE RELIABILITY" The prioritization of Generic Issue No. 165, " Spring-Actuated Safety and -

Relief Valve Reliability," shows that the issue has a HIGH priority ranking.

This memorandum approves RES/DSIR taking appropriate actions, within current resource allocations, to resolve the issue. The evaluation of the subject issue is provided in Enclosure I.

In accordance with RES Office Letter No. I, "Procedere for Identification, l

Prioritization, and Tracking of the Resolution of Generic Issues," the resolution of this issue will be monitored by the Generic Issue Management Control System (GIMCS). The information needed for this system is indicated i

on the enclosed GIMCS information sheet (Enclosure 2). As stated in the Office Letter, the information needed should be provided within 6 weeks.

The enclosed prioritization evaluation will be incorporated into NUREG-0933, "A Prioritization of Generic Safety Issues," and is being sent to the regions, other offices, the ACRS, and the PDR, by copy of this memorandum, to a310w others the opportunity to comment on the evaluation. Any changes as L result l

I of comments will be coordinated with you. However, the schedule 4r the resolution of this issue should not be delayed to wait for it.ese comments.

The information requested should be sent to the Engineering Issues Branch, DSIR, RES (Mail Stop NL/S-314). Should you have any questions pertaining to the contents of this memorandum, please contact Ronald Emrit (492-3731).

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4 Eric S. Beckjohc, Director l

Office of Nuclear Regulatory Research

Enclosures:

I.

Prioritization Evaluation 2.

GIMCS Information Sheet i

9312090116 931126 I

PDR OTECI GMISC I

PDR

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cc:

T. Murley, NRR E. Jordan, AE0D T. Martin, Reg. I S. Ebneter, Reg. II A. Davis, Reg. III J. Milhoan, Reg. IV J. Hartin, Reg. V ACRS PDR DCS -

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PRIORITIZATION EVALUATION Issue 165: Spring-Actuated and Safety Relief Valve Reliability

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ISSUE 165: SPRING-ACTUATED SAFETY AND RELIEF VALVE RELIABILITY DESCRIPTION Spring-actuated safety and relief valves provide overpressure protection for a number of systems in both PWRs and BWRs. However, failure of these valves in safety-related support systems could cause a significant diversion of flow from these systems, and thus, prevent the systems from performing their designed funr. tion. It is estimated that perhaps 3-5, out of a total of 55-60 spring-actuated safety and relief valves installed in such safety-related systems of a typical PWR or BWR plant, would be significant contributors to core-melt frequency. Also, due to the size of these valves (< 4 inches), it is believed that most of them could be tested at the plant site, and many of them in situ, thus reducing the time and cost for testing. For these reasons, this issue addresses the unreliability of spring-actuated safety and relief valves in safety-related support systems.

Historical Backcround

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On a number of occasions, spring-actuated safety and relief valves failed to meet setpoint criteria within the desired tolerance. Other incidents included more seriously degraded performance of safety and relief valves.*' In AE00 Special Study S92-02, April 1992, the staff concluded that most pressurizer j

safety valves (PSVs), main steam safety valves (MSSVs), and BWR safety / relief valves (SRVs) do not meet the one percent setpoint drift tolerance and many are above three percent.

These results suggest that other systems which have safety and relief valves could be adversely affected by setpoint drift. The staff discussed some of these systems in IHs 90-05 and 92-64, and in NUREG/CR-6001. More importantly, at the Shearon Harris Nuclear Power Plant, the failure of a high head safety injection RV to operate at a very low setpoint resulted in the undetected loss of the entire system and would have resulted in inadequate emergency core coolant injection if a small or intermediate break loss-of-coolant accident (LOCA) occurred. This event is discussed in detail in LER 91-008-01 and IN 92-61.

l Although the staff established generit. safety issue (GSI) B-55 for increasing the reliability of Target Rock two-stage pilot-operated SRVs and GSI 70 to j

address the reliability of power-operated relief valves (PORVs) and PORY block l

valves, it has not established a similar GSI for spring-actuated SVs and RVs.

Therefore, because significant NRC and industry resources have been spent in the past on both evaluating the risk and improving the reliability of PSVs, PORVs, MSSVs, and BWR SRVs, the focus of this GSI is limited to spring-actuated RVs in safety-related support systems and the effects of their l

unreliability on plant operation.

Safety Sionificance Spring-actuated relief valve reliability is significant to safety for light water reactors, because their unreliability can lead to a core-melt from loss 1 l

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. l of core cooling and inventory makeup. Possible sources of loss include the j

following:

1) failure of a valve to close after opening,

2) failure of a valve to open when challenged, resulting in overpressure conditions that precipitate a LOCA, and

3) premature opening of a valve below setpoint resulting in a LOCA.

Possible Solutions The proposed safety issue resolution (SIR) involves the improved periodic l

inspection and testing of spring-actuated RVs in risk-significant systems.

PRIORITY DETERMINATION Assumotions This safety issue is assumed to affect all operating plants.

Implementaticm i

of the SIR, if found necessary, can be achieved at future plants with minimal incremental costs, and thus, a forward-fit evaluation was not performed.

l Therefore, 71 operating plants were assumed: 47 PWRs and 24 BWRs with average I

remaining lives of 27.7 and 25.2 years, respectively. This corresponds to the number of plants existing or planned at the time of the publication of i

NUREG/CR-2800, not the actual number to date.

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The Surry PRA"** was used to model PWR RVs in SARA 4.0.""

The Grand Gulf l

PRA" was primarily used to model BWR RVs. The Peach Bottom PRA"** was used to support the Grand Gulf results.

Failure of a RV to operate within the allowable opening and closing setpoints was considered a failure of the valve. However, not all valve failures necessarily fail the train of the system in which they operate. Therefore, it was conservatively assumed that 10 percent of the valve failures would fail their trains. NPRDS was used to obtain values of RV unreliability for various systems throughout the plant that had spring-actuated RVs. From these data, a best-estimate probability of the RV to fail its train was calculated to be 5.0E-03/ demand (based on 524 valve failures out of 10,063 events multiplied by a 10 percent train failure probability). The upper bound probability is 5.0E-02/ demand assuming the RV failure always results in train failure. A lower bound probability was estimated by using the AE0D/S92-02 report which considered nine valve failures out of 1100 events, equaling a probability of 1.0E-03/ demand including the 10 percent train failure probability.

FreauencY Estimate Because the Surry PRA does not include RVs in every system, modifications to the PRA are required to model their effects on a particular system.

For those systems where RVs are included with a component in a single train whose unavailability could fail the entire system, the failure probability of the RV i

was simply added to the component's failure probability. On the other hand, for those systems where RVs are included w;th components in two trains, and thus, common mode failure can occur, the failure probability of the RV had to

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be added by taking into account the use of beta factors in the component's failure probability. A beta facto: is defined as the conditional probability of a component failure given that a similar component has failed.

It can be shawn that P, the component failure probability including the RV reliability, and B, the beta factor for the RV and component, are given by:

P = P, + P, Eq. (1) and p

Oc, + 0,P, P

Eq. (2)

(P,+P,)

where B. and B. are the beta factors, and P, and P. are the failure probabilities, for the component and RV, respectively. This analysis used a value of 7.0E-02 for B., which was obtained from the beta factor for a SRV in the PRA. The values of B. and P. were obtained from the applicable caponent in the PRA. Using Eq. (1) and Eq. (2), the values of P and # were calculated and then inserted into SARA for those systems that had dual trains.

The effect of the SIR would be to improve the reliability that the valves operate as designed. To reflect this, it was assumed that the SIR will reduce the probability for a failure of a safety or relief valve to a negligible amount and thus bring the core-melt frequency to the values predicted by the plant-specific PRAs. As a result, in SARA the base-case core-melt frequency value represents the value after SIR implementation, and the adjusted-case core-melt frequency represents the increased risk from including the effects of safety and relief valve unreliability. Therefore, the change in core-melt frequency computed in SARA gives the result of improving safety and relief reliability. The changes in core-melt frequency for various systems in the Surry PRA are sumarized in Table 3.165-1. Diesel and Emergency Power includes RVs in the emergency diesel generator air start system (see IN 90-018). The change for the Component Cooling Water, Containment Spray, Main Feedwater, and Essential Service Water systems were negligible.

The significant changes in core-melt frequency for various systems in the Grand Gulf PRA are summarized in Table 3.165-2. The change for other systems studied - which included the RHR/LPI, Feedwater, Condensate, Standby Liquid Control, Control Rod Drive, Nuclear Steam Supply Shutoff, and Low Pressure Core Spray systems - were negligible. The Peach Bottom PRA was used in SARA to further validate the change from the Essential Service Water system computed in the Grand Gulf PRA. These results support that findirig.

Conseauence Estimate The containment failure probabilities and base consequences were taken from NUREG/CR-2800" for similar accident sequences. The results from the

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. Table 3.165-1 Chance in Core-Melt Frecuency for Various PWR Systeu PWR System Valve Failure Probability Estimate Best Estimate Lower Bound Upper Bound (5.0E-03)

(1.0E-03)

(5.0E-02)

High Pressure 1.0E-05 2.0E-06 1.0E-04 Injection Diesel and Emergency 7.3E-06 1.5E-06 9.2E-05 Power Accumulator 5.0E-06 1.0E-06 4.8E-05

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Reactor Coolant 2.3E-06 4.7E-07 2.1E-05 e val /L ressure l

Injection Auxiliary Feedwater 6.7E-07 1.3E-07 9.2E-06 Chemical and Volume 3.3E-07 6.7E-08 3.3E-06 Control System PWR Total 2.6E-05 5.3E-06 2.9E-04 Table 3.165-2 Chance in Core-Melt Frecuency for Various 6WR Systems BWR System Valve Failure Probability Estimates Best Estimate Lower Bound Upper Bound j

(5.0E-03)

(1.0E-03)

(5.0E-02) 1 Essential Service 1.6E-06 3.2E-07 1.4E-05 i

Water Diesel and Emergency 3.8E-07 7.5E-08 7.2E-06 Power RCIC 3.6E-08 7.2E-09 3.5E-07 HP Core Spray 1.7E-08 3.3E-09 1.7E-07 Main Steam 0

0 2.9E-08 BWR Total 2.0E-06 4.0E-07 2.2E-05

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. 1 per-plant calculations for the changes in public risk, and also the changes in core-melt frequency, are shown in Table 3.165-3 for the three different estimates of valve failure probability. The total public risk reduction is 1.0E+05 person-rem with a lower bound estimate of 2.0E+04 person-rem and an upper bound estimate of 1.0E+06 person-rem. These values would increase by about 50% if three-fourths of the plants had their license renewed for a 20-year period.

Table 3.165-3 PWR and BWR Results for Chances in Core-Melt Freauency and Public Risk Change in Core-Melt Change in Public Risk Frequency (per RY) for (person-rem /RY) for Various Valve Failure Various Valve Failure Probabilities Probabilities 5.0E-03 1.0E-03 3.0E-02 5.0E-03 1.0E-03 5.0E-02

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I PWR 2.6E-05 5.3E-06 2.9E-04 7.3E+01 1.5E+01 7.7E+02 BWR 2.0E-06 4.0E-07 2.2E-05 5.8E+00 1.2E+00 6.2E+01 Cost Estimate Industry Cost: As stated previously, resolution of this issue would require improved periodic inspection and testing of RV systems. Assuming these activities were required every year and could be performed in about two man-days, it was determined that the total annual test and inspection requirements for each system would be about 2 man-days /RY. Assuming five affected systems per plant results in a total labor of 2 man-week /RY. At a labor cost of

$2270/ man-wk, the per-plant industry cost for inspection and testing would be

$4540/RY [(2 man-wk/RY) ($2270/ man-wk)], which results in a total industry cost for inspection and testing of $8.7E+06 {($4540/RY)[(47 PWRs)(27.7 yr) +

l (24 BWRs)(25.2 yr)]}. Because the industry already does testing every 10 years, the value calculated here for industry costs is conservatively _ high.

1 NRC Cost: NRC costs to review SIR test and inspection requirements are anticipated to be minimal. Thus, a total of 3 man-days /RY (0.6 man-wk/RY) was estimated for these reviews. At an assumed labor. cost of $2270/ man-wk, the total NRC cost to review SIR tests and inspections was calculated as $2.6E+06

{(0.6 man-wk/RY) ($2270/ man-wk)[(47 PWRs)(27.7 yr) + (24 BWRs)(25.2 yr)))

Other costs, such as work with ASME Code Committees for revisions to increase the valve testing frequency, were estimated to be negligible.

The total cost is the sum of the industry cost of $8.7E+06 and the NRC cost of

$2.6E+06 for a total of $1.lE+07. The major elements of these costs are shown in Table 3.165-4.

Impact /Value Assessment The Impact /Value Assessment is based upon the total public risk reduction and 4

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_ the total SIR cost. The best estimate of the public risk reduction was 1.0E+05 person-rem, therefore the Impact /Value Assessment is 1.1E+02 $/ person-rem (with upper and lower error bounds of 1.1E401 and 5.5E+02 $/ person-rem, i

respectively, calculated from the previous lower and upper bound estimates for public risk reduction).

Table 3.165-4 NRC and Industry Costs Inspection and Testing

$ 8.7E+06 TOTAL INDUSTRY COSTS

$ 8.7E+06 l

Review of SIR tests and inspections 5 2.6E+06 TOTAL NRC COSTS

$ 2.6E+06 TOTAL COSTS

$ 1.1E+07 Other Considerations Occupational DoLei l

The total occupational dose increase for SIR implementation was calculated to be 3.8E+02 person-rem for all plants (with upper and lower error bounds of 1.1E+03 and 2.5E+02 person-rem, respectively).

Accident Avoidance Doses The occupational dose reduction due to accident avoidance was calculated to be 7.2E+02 person-rem for all plants (with upper and lower error bounds of 7.5E+03 and 0 person-rem, respectively).

Accident Avoidance Costs The industry cost savings due to accident avoidance were calculated to be

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$6.0E+07.

CONCLUSION Based on the impact /value score and the changes in risk, this issue is rated a HIGH priority issue.

REFERENCES 64.

NUREG/CR-2800, " Guidelines for Nuclear Power Plant Safety Issue Prioritization Information Development," U.S. Nuclear Regulatory Commission, February 1983, (Supplement 1) May 1983, (Supplement 2)

December 1983, (Supplement 3) September 1985, (Supplement 4) July 1986.

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_7 1318. NUREG/CR-4550, " Analysis of Core Damage Frequency from Internal Events,"

U.S. Nuclear Regulatory Commission, (Volume 1, Rev.1) January 1990, (Volume 2) April 1989, (Volume 3, Rev.1) April 1990, bolume 4, Rev.1)

December 1990, (Volume 5, Rev.1) April 1990, (Volume 6) April 1987, (Volume 7 Rev.1) May 1990.

1456. NUREG/CR-5303, " System Analysis and Risk Assessment System (SARA)

Version 4.0," U.S. Nuclear Regulatory Commission, (Volume 1) February 1992, (Volume 2) January 1992.

i 1520. Memorandum for E. Beckjord from T. Murley, " Request to Prioritize a New

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Generic Issue for Spring-Actuated Safety and Relief Valve Reliability,"

4 October 8, 1992.

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ENCLOSURE 2 Page 1 of 2

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t Managment and contr-1 indicators used in GIMCS are defined as follows:

Generic Issue Number 1.

Issue No.

Generic issue Title 2.

Title Date the issue was identified 3.

Identification Date The date that the prioritization evaluation 4.

Prioritization Date l

was approved by the RES Director l

Generic Safety (GSI), Licensing (LI), or 5.

Type Regulatory Impact (RI) l High (H) or Medium (H) l 6.

Priority Name of assigned individual responsible for 7.

Task Manager resolution l

The Office, Division, and Branch of the Task 8.

Office /Div/Br Manager who has lead responsibility for resolving the issue.

Technical assistance funds Active 9.

Action Level appropriated for resolution and/or Task Manager actively pursuing resolution No technical assistance funds Inactive appropriated.for resolution, Task Manager assigned to more important work, or no Task Manager assigned l

All necessary work has been Resolved completed and no additional resources will expended Coded summary as follows:

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Status NR - (Nearly-Resolved);

3A - (Resolved with requirements);

3B - (Resolved with No requirements);

5 - (Licensing or Regulatory Impact issued that should be assigned resources for completion)

Task Action Control (TAC) number assigned j

11. TAC Number to the issue Scheduled resolution date for the issue

12. Resolution Date 1

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l ENCLOSURE 2 Page 2 of 2 i

1 Who or what authorized work _to be done on

13. Work Authorization the issue Financial identification number assigned to 14.

FIN contract (if any) for technical assistance

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Contractor name

15. Contractor Contract Title (if contract issued)

16. Contract Title Describes briefly the work necessary to

17. Work Scope technically resolve and complete the generic issue Describes current status of work

18. Status Identifies documents into which the

19. Affected Documents technical resolution will be incorporated j

Identifies problem areas and describes what

20. Problem / Resolution actions are necessary to resolve them Selected significant ailestones:

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21. Milestones (a) the " original" scheduled dates reflect the original Task Action Plan plus addi-tional milestone dates added during task l

resolution; (b) c.hanges in the the original scheduled dates are listed under " Current";

(c) actual completion dates are listed under

" Actual"

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