ML20003H276

Primary Coolant Sys Pressure Isolation Valves,Fort Calhoun Unit 1, Technical Evaluation Rept

ML20003H276
Person / Time
Site:	Fort Calhoun
Issue date:	10/24/1980
From:	Noell P, Stilwell T FRANKLIN INSTITUTE
To:	Polk P Office of Nuclear Reactor Regulation
Shared Package
ML20003H200	List: ML20003H276
References
CON-NRC-03-79-118, CON-NRC-3-79-118 TER-C5257-222, NUDOCS 8105050477
Download: ML20003H276 (10)

Text

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ATTAcif4ENT 2 TECHNICAL EVALUATION REPORT PRIMARY COOLANT SYSTEM ,

1 PRESSU RE ISOLATION VALVES '

OMAHA PUBLIC POWER DISTRICT FORT CALHOUN UNIT 1 NRC DOCKET NO. 50-285 NRC TAC NO. 12888 FRC PROJECT C5257 NRC CONTRACT NO. NRC.03-79-118 FRCTASK 222 Prepared by Franklin Research Center Author: P. N. Noell The Parkway at Twentieth Street T. c. Stilwell Philadelphia, PA 19103 FRC Group Leader: P. N. Noell Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer: P. J. Polk

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October 24, 1980 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, apparatus, 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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. 0. Franklin Research Center A Division of The Franklin Institute The Beneemn Frenman Parkway. Pht!a. Pa 19103 (215)448-1000

1.0 INTRODUCTION

The NRC has determined that certain isolation valve configurations ie systema connecting the high pr, essure Primary Coolant System (PCS) to lower-pressure systems extending outside containment are potentially significant contributors to an intersystem loss-of-coolant accident (LOCA). Such configu-rations have been found to represent a significant factor in the risk computed for core melt accidents.

The sequence of events leading to the core melt is initiated by the con-current failure of two in-series check valves to function as a pressure isola-tion barrier between the high pressure PCS and a lower pressure system extend-ing beyond, containment. This failure can cause an overpressurization and rup-cure of the low pressure system, resulting in a LOCA that bypasses containment.

The NRC has cetermined that the probability of failure of these check valves as a pressure isolation barrier can be significantly reduced if the pressure at each valve is continuously monitored, or if each valve is periodi-cally inspected by leakage testing, ultrasonic examination, or radiographic inspection. The NRC has established a program to provide increased assurance that such multiple isolation barriers are in place in all operating Light

.k*ater Reactor plants designated by DOR Ceneric Implementation Activity B-45.

i In a generic letter of February 23, 1980, the NRC requested all licensees j to identify the following valve configurations which may exist in any of their plant systems communicating with the PCS: 1) two check valves in series or 2) two check valves in series with a motor-operated valve (MOV).

For plants in which valve configurations of concern are found to exist, licensees were further requested to indicate: 1) whether, to ensur + 'stegrity of the various pressure isolation check valves, continuous surveillance or periodic testing was currently being conducted, 2) whether any check valves of concern were known to lack integrity, and 3) whether plant procedures should be revised or plant modifications be made to increase reliability.

Franklin Research Center (FRC) was requested by the NRC to provide tech-nical assistance to NRC's B-45 activity by reviewing each licensee's submittal

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against criteria provided by the NRC and by verifying the licensee's reported findings from plant system drawings. This report documents FRC's technical review.

2.0 CRITERIA 2.1- Identification Criteria For a piping system to have a valve configuration of concern, the follow-ing five items must be fulfilled:

1) The high pressure system must be connected to the Primary Coolant System;

2) there must be a high-pressure / low pressure interface present in the line;

3) this same piping must eventually lead outside containment;

4) the line must have one of the valve configurations shown in Figure 1; and

5) the pipe line must have a diameter greater than 1 inch.

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$ A HF :  ; LP Figure 1. valve Configurations Designated by the NRC To Be Included in This Technical Evaluation

2.2 Periodic Testing Criteria For licensees whose plants have valve configurations of concern and choose to institute periodic valve leakage testing, the NRC has established criteria for frequency of testing, test conditions, and acceptable leakage races.

These criteria may be sirmmarized as follows:

2.2.1 Frequency of Testing Periodic hydrostatic leakage testing

• on each check va*ve shall be accom-plished every time the plant is placed in the ccid shutdown condition for refueling, each time the planc is placed in a cold shutdown condition for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> if testing has not been accomplished in, the preceding 9 months ,

each time any check valve may have moved from the fully closed position (i.e., any time the differen- tial pressure across the valve is less than 100 psig), and prior to returning the valve to. service af ter maincenance, repair, or replacement work is performed.

2.2.2 Hydrostatic Pressure Criteria Leakage tests involving pressure differentials lower than function pres-sure differentials are permitted in those types of valves in which service ,

pressure vill tend to diminish the overall leakage channel opening, as by l pressing the disk into or onto the seat with greater force. Cate valves, l check valves , and globe-cype valves , having function pressure differencial applied over the seat, are examples of valve applications satisfying this requirement. '4 hen leakage cests are made in such cases using pressures lower than function maximum pressure differential, the observed leakage shall be adjusted to function maximum pressure differencial value. This adjustment shall be made by calculation appropriate to the test media and I the ratio between test and function pressure differential, assuming leak- )

age to be directly proportional to the pressure differential to the one- l half power.

2.2.3 Acce"peable Leakage Rates:

e Leakage races less chan or equal to 1.0 gpm are considered accept-able.

l e Leakage races greater than 1.0 gym but less chan or equal to 5.0 gym are considered acceptable if the latest measured race has not l exceeded the race determined by the previous test by an amount

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• To satis fy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance with approved procedures and supported by computations showing that the method is capable of demonstrating valve compliance with the leakage criteria.

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that reduces the margin between the measured leakage rate and the  ;

maximum permissible rate of 5.0 gym by 50% or greater. '

e Leakage rates greater than 1.0 gpa but less than or equal to 5.0 gym are considered unacceptable if the latest measured rate ex-caeded the rate determined by the previous test by an amount that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gym by 50% or greater.

e Leakage rates greater than 5.0 gpa are considered unacceptable.

3.0 TECHNICAL EVALUATION

3.1 Licensee's Response to the Generic Letter In response to the NRC's generic letter (Ref.1], the Omaha Public Power District (OPP) stated (Ref. 2] that, " piping configurations at the Fort Calhoun station do not conform to the Event V configurations of Figures I-4-6 and V-4-3 of Wash-1400."

The licensee also itemized the various piping safety design features and valve leakage surveillance techniques in use in the Fort Calhoun Unit 1 in order to assure the integrity of all high-pressure boundary barriers.

It is FRC's understanding that, with OPP's concurrence, the NRC will direct OPP to change its Plant Technical Specifications as necessary to ensure that periodic leakage testing (or equivalent testing) is conducted in accor-dance with the criteria of Section 2.2.

3.2 FRC Review of Licensee's Response FRC has reviewed the licensee's response against the plant-specific Piping and Instrumentation Diagrams (P& ids) (Ref. 3] that might have the valve con-figurations of concern.

FRC has also reviewed the efficacy of instituting periodic testing for the check valves involved in this particular application with respect to the re-duction of the probability of an intersystem LOCA in the High- and Low-Pressure Safety Injection System pipe lines.

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In its review of the P& ids (Ref. 3] for Fort Calhoun Unit 1, FRC found the following two piping systems to be of concern:

The Eigh- and Low-Pressure Safety Injection Systems are connected to the Primary Coolant System by a single, common piping line to each of the cold-leg sides of the four PCS loops, IA, IB, 2A, and 2B, The Eigh-Pressure Safety Injection System has a two check valve in-series configuration of concern with the high pressure / low-pressure interface located on the upstream side of the check valve farthest from the reactor vessel.

The Low-Pressure Safety Injection System contains a two check valve and a motor-operated valve (MOV) in one of the series configuration of concern. The high-pressure / low-pressure interface is located on the upstream side of the MOV. The valves comprising each configura-tion of concern are listed below:

' Righ-Pressure Safety Injection Loop 1A, cold leg high pressure check valve, SI-216 high pressure check valve, SI-201 Loon IB, cold leg high pressure check valve, SI-220 high pressure check valve, SI-204 Loop 2A, cold leg high-pressure check valve, SI-208 high-pressure check valve, SI-195 Loop 2B, cold leg high pressure check valve, SI-212 high pressure check valve, SI-198 Low-Pressure Safety Injectica Loop 1A, cold leg high pressure check valve, SI-216 high pressure check valve, SI-200 high-pressure MOV, HCV-329, normally closed (n.c.)

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Loop 15, cold leg high-pressure check valve. SI-220 i high pressure check valve,51-203 high-pressure MOV, HCV-327, n.c.

Loop 2A, cold leg high-pressure check valve, SI-208 high-pressure check valve, SI-194 high-pressure MOV, HCV-331, n.c.

Loco 25, cold leg high-pressure check valve, SI-212 high-pressure check valve, SI-197 high-pressure MOV, HCV-333, n.c.

In accordance with the criteria of Section 2.0, FRC found no other valve configurations of concern existing in this plant.

FRC reviewed the effectiveness of instituting periodic leakage testing of the check valves in these lines as a means of reducing the probability of an intersystem LOCA occurring. FRC found that introducing a program of check valve leakage testing in accordance with the criteria summarized in Section 2.0 will be an effective measure in substantially reducing the probability of in intersystem LOCA occurring in these lines, and a means of increasing the probability that these lines will be able to perform their safety-related func t ions . It is also a step toward achieving a corresponding reduction in the plant probability of an intersystem LOCA in Fort Calhoun Unit 1.

4.0 CONGLUSION Based on the previously decketed information and drawings made available l for FRC review, FRC found that the cold-leg branches of the High and Low-Pressure Safety Injection Systems in Fort Calhoun Unit 1 contain valving in two l

of the configurations of concern (identified in Figure 1). Thus , if the licensee's review of the valving configurations contained in the cold-leg branches of the High and Low-Pressure Safety Injection Systems confirms FRC's l

I finding, then the valve configurations of concern existing in Fort Calhoun Unit 1 incorporate the valves listed in Table 1.0.

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If OPP modifies the Plant Technical Specifications for Fort Calhoun Unit 1 to incorporate periodic testing (as delineated in Section 2.2) for the check valves itemized in Table 1.0, then FRC considers this an acceptable means of achieving plant compliance with the NRC staff objectives of Reference 1.

Table 1.0 Primary Coolant System Pressure Isolation Valves System Check Valve No. Allowable Leakage

• High-Pressure Safety Injection Loop 1A, cold leg SI-216 SI-201 Loop 1B, cold leg SI-220 SI-204 Loop 2A, cold leg SI-208 SI-195 Loop 2B, cold leg SI-212 SI-198 Low-Pressure Safety Injection Loop 1A, cold leg SI-200 Loop 1B, cold leg SI-203 Loop 2A, cold leg SI-194 Loop 23, cold leg SI-197

5.0 REFERENCES

1. Generic NRC letter, dated 2/23/80, from Mr. D. G. Eisenhut, Department of Operating Reactors (DOR), to Mr. 4. C. Jones, Omaha Public Power District (OPP).

• To be provided by licensee at a future date in accordance with Section 2.2.3.

2. Omaha Public Power District's response to NRC's letter, dated 3/28/80, from Mr. k'. C. Jones (OPP) to Mr. D. G. Eisenhut (DOR).

3. List of examined P& ids:

Combustion Engineering Drawings of Fort Calhoun Unit 1:

E-23866-210-110, (Rev. 6)

E-23866-210-120, (Rev. 8) Sh.1 of 2 E-23866-210-120, (Rev. 8) Sh. 2 of 2 E-23866-210-121, (Rev. 9)

E-23866-210-130, (Rev. 9) Sh.1 of 2 E-23866-210-130, (Rev. 9) Sh. 2 of 2 Gibbs, Hill, Durham & Richardson, Inc. , Drawings of Fort Calhoun Unit 1:

11405-M-5 11405-M-10 11405-M-12 .

11405-M-40 t

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