ML20003G103
| ML20003G103 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 10/24/1980 |
| From: | Noell P, Stilwell T Franklin Research Ctr |
| To: | Polk P Office of Nuclear Reactor Regulation |
| Shared Package | |
| ML20003G098 | List: |
| References | |
| CON-NRC-03-79-118, TAC 12883 TER-C5257-217-R01, NUDOCS 8104280257 | |
| Download: ML20003G103 (7) | |
Text
- _ _ _
THIS REPORT SUPERSEDES ISSUE OF AUGUST 22, 1980 O
TECHNICAL EVALUATION REPORT PRIMARY COOLANT SYSTEM PRESSURE ISOLATION VALVES FLORIDA POWER CORPORATION CRYSTAL RIVER UNIT 3 NRC DCOXET P.O.
50-302 NRC TAC NO.
12883 FRC PROJECT C5257 NRC CONTRACT NO. NRC-03-79-118 FRC TASK 217 Prepared by Franklin Research Center Author:
P. N. Noell The Parkway at Twentieth Street T. C. Stilwell Philadelphia, PA 19'.t'3 FRC Group Leader:
P. N. Noell Prepared for Nuclear Regulatory Commission Washington, D.C. 20555 Lead NRC Engineer:
P. J. Polk 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.
A l.
. Franklin Research Center l
A Division of The Franklin Institute S104280 M 1 h Beernn Frankhn Parkway. PMa., Pa. 19103 (21S) 448-1000 l
l 1
1.0 INTRODUCTION
The NRC has determined that certain isolation valve configurations in systems connecting the high-pressure 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 =elt 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-ture of the low-pressure system, resulting in a LOCA that bypasses containment.
The NRC has determined that the probability of fail tre of these check valves as a pressure isolation barrier can be significantly reduced if the pressure at each valve is continucualy 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 Water Reactor plants designated by DCR Generic Inplementation Activity 3-45.
In a generic letter of February 23, 1980, the NRC requested all licensees to identify the following valve configurations which =ay 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).
7or plants in which valve configurations of concern are found to exist, licensees were further requested to indicate: 1) whether, to ensure integrity of the various pressure isolation check valves, continuous surveillance or periodic testing was currently being conducted, 2) whether any check valves of concern were snown 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 3-45 activity by reviewing each licensee's submittal 1
~
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 syste.n 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 Primarv Coolant Systes;
- 2) there must be a high-oressure/ 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) che pipe line =ust have a diameter greater than 1 inch.
PCS :
psowl I
'/*
'A
=
I h
/
'A
=
I
'[
'/
psowl m
l A
~
=
HP :
LP Figure 1.
Valve Configurations Designated by the NRC To 3e 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 1e:kage testing, the NRC has established criteria for frequency of testing, test conditions, and acceptable 1erkage races.
These criteria may be summarized is follows:
2.2.1 Frequency of Tes ting Periodic hydrostatic leakage testing
- on each check valve shall be accom-plished every ti=e the plant ia placed in the cold shutdown condition for refueling, each time che plant is slaced 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 a:ccmplished in. :he preceding 9 zonths,
each time any check valve may have =oved from the fully closed position (i.e., any time the dif feren-tial pressure across the valve is less than 100 psig), and prior to returning the valve to service af ter =aintenance, repair, or replacement work is performed.
2.2.2 Hydrostatic Pressure Criteria Leakage tests involving pressure differentials lower than function pres-sure dif ferentials are permitted in those types of valves in which service pressure will tend :o diminish the overall leakage channel opening, as by pressing the disk into or onto the seat with greater force. Ga:e valves, check valves. and globe-type valves, having fune:irn pressure dif ferential applied over the seat, are examples of valve applications satisfying this r equireme r.t.
- 4hcn leakage tests are made in such cases using pressures lower :han function maxi =um pressure differential, the observed leakage shall be adjusted to function maxi =um pressure differential value.
This adjust =ent shall be made by calculation appropriate :o :he test media and the ratio between test and function pressure dif ferential, assuming leak-age to be directly propor:ional to the pressure differential to the one-half power.
2.2.3 Acceptable Leakage Rates:
Leakage rates less than or equal to 1.0 gpm are considered accept-able.
Leakage rates greater than 1.0 gpm but less than or equal :o 5.0 gym are considered acceptable if the lates t measured rate has not exceeded the rate determined by the previous test by an amount
- To satis fy ALARA requirements, leakage may be measured indirectly (as from the performance of pressure indicators) if accomplished in accordance wi:h approved procedures and supported by computations showing : hat the method is capable of demonstrating valve compliance with the leakage criteria.
' t
that reduces the margin between measured leakage rate and the maximum permissible rate of 5.0 gpm by 50% or greater, Leakage rates greater than 1.0 gpa but less than or equal to 5.0 e
gpa are considered unacceptable if the latest measured rate ex-coeded the race determined by the previous test by an amount tha t reduces "the margin between measured leakage rate and the maximum permissible rate of 5.0 gym by 50% or greater.
Leakage, rates greater than 5.0 gym are considered unacceptable.
e
3.0 TECHNICAL EVALUATION
3.1 Licensee's Response to the Generic Letter In response to the NRC's generic letter (Ref.1], the Florida ' tower Corporation (FPC) stated (Ref. 2] that a valve configuration of concern does exist in the Decay Heat / Low-Pressure Injection System at Crystal River Unit 3.
FPC described the system as follows: "The decay heat / low pressure injection system is isolated from the RCS by two check valves in series with a motor-operated valve."
The licensee then itemized the four check valver which exist in the A and 3 trains of the Decay Heat / Low-Pressure Injection System, namely; DHV-1, DMV-2, CFV-1, CFV-3.
The Licensee further stated "No tests are being accomplished at this time."
It is FRC's understanding that, with FPC's concurrence, the NRC will direct FPC to change its Plant Technical Specifications as necessary to ensure that periodic leakage testing (or equivalent testing) is conducted in accor-4 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 Decay Heat / Low-Pressure Injection piping lines.
I
_4
- ~.
In its review of the P& ids [Ref. 3] for the Crystal River Unit 3 FRC found the following piping system to be of concern:
The Decay Heat / Low-Pressure Injection System (DH/LPIS) is composed of two piping trains (A and B) each connected to the reactor ves-sel. Each train has two check valves and a motor-operated valve in one of the series configurations of concern. In each train the high-pressure / low-pressure interface is located on the upstream side of the motor-operated valve (MOV). These valves are listed below:
Decay Heat / Low-Pressure Injection System Train A high-pressure check valve, CFV-1 high-pressure check valve, DRV-2 high-pressure check MOV, DHV-6 (DH-V4B), normally closed Train 3 high-pressure Check valve, CTV-3 high-pressure ch sk valve, DHV-l high pressure MOV, DHV-5 (DH-V4A), normally closed In accordance with the criteria of Section 2.0, FRC has found no other valve configurations of concern existing in this plant. These findings con-firm the licensee's response [Ref. 2].
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 j
intersystem LOCA occurring. FRC found that introducing r. program of check valve leakage testing in accordance with the criteria summarized in Section i
2.0 will be an..lective measure in substantially reducing the probability of an intersystem LOCA occurring in these lines, and a means of increasing the probability that' these lines will be able to perform their safety-related functions. It is also a step toward achieving a corresponding reduction in the plant probability of an intersystem LOCA in the Crystal River Unit 3.
4.0 CONCLUSION
Crystal River Unit 3 has been determined to have valves in one of the con- __
i figurations of concern in both A and B trains of the Decay Heat / Low-Pressure Injectio2 System.
If IPC modifies the Plant Technical Specification for Crystal River Unit 3 to incorporate periodic ta'...ng (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
- Decay Heat / Low-Pressure Safety Injection Train A CFV-1 DEV-2 i
Train B CFV-3 DRV-1
5.0 REFERENCES
l
[1]. Generic NRC letter, dated 2/23/80, from Mr. D. G. Eisenhut, De partment of Operating Reactors (DOR), to Mr. R. M. Bright, Florida Power Corporation (FPC).
[2]. Florida Power Corporation's response to NRC's letter, dated 3/14/80, from Mr. R. M. Bright (FPC) to Mr. D. G. Eisenhut (DOR).
[3]. List of examined P& ids:
Gilbert Associates Drawings:
E-318-641, (Rev.2)
E-318-631, ( Rev.2)
E-318-661, ( Rev.2)
C-318-702 (Rev.2)
- To be provided by licensee at a future date in accordance with Section 2.2.3. _..