ML20151T595
| ML20151T595 | |
| Person / Time | |
|---|---|
| Site: | Trojan File:Portland General Electric icon.png |
| Issue date: | 08/12/1988 |
| From: | Bauer S PORTLAND GENERAL ELECTRIC CO. |
| To: | |
| Shared Package | |
| ML20151T495 | List: |
| References | |
| NUDOCS 8808160416 | |
| Download: ML20151T595 (10) | |
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, LCA 169 Attachment Page 1 of 2 TABLE 3.4.6-1 REACTOR COOLANT SYSTEM PRESSURE ISOLATION VALVES (a)(b)(c)
(CHECK VALVES)
VALVE NUMBER FUNCTION 8948 A, B, C, O To RCS Cold Legs 8949 A, B, C, D To RCS Hot Legs 8818 A, B, C, D RHR to RCS Cold legs 8819 A, B, C, D SI to RCS Cold legs 8905 A B, C, O Si to RCS Hot Legs 8736 A, B RHR to RCS Hot Legs
!LOTES:
(a) Maximum Allowable _ Leakage (each valve):
- 1. Leakage rates less than or equal to 1.0 gpm are considered acceptable.
- 2. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered acceptable if the latest measured rate has not exceeded 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 gpm by 50 percent or greater.
- 3. Leakage rates greater than 1.0 gpm but less than or equal to 5.0 gpm are considered unacceptable if the latest measured rate exceeded 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 gpm by 50 percent or greater.
- 4. Leakage rates greater than 5.0 gpm are considered unacceptable.
- 5. When leakage tests are performed using a test differential pressure j lower than the function maximum dif forential pressure, observed leakage rates shall be adjusted to function maximum differential 1 nressure values. 1 (b) To satisfy ALARA requirements, leakage may be measured indirectly (as '
f rom 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.
(c) Minimum test dif ferential pressure shall not be less than 150 psid.
TROJAN - UNIT 1 3/4 4-15a Order dated 4/20/81 8808160416 880812 g() l PDR ADOCK 05000344 ;
p PDC .. I J
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. 1 LCA 169 Attachment Page 2 of 2 e
REACTOR COOLANT SYSTEM BASES 33 4.6.2 OPERATIONAL LEAKAGE Industry experience has shown that while a limited amount of leakage is expected from the RCS, the unidentified portion of this leakage can be reduced to a threshold value of less than 1 GPM. This threshold value is sufficiently low to ensure early detection of additional leakage.
The Surveillance Requirements for RCS Pressure Isolation Valves provide added assurance of valve integrity thereby reducing the probability of gross valve failure and consequent intersystem LOCA. Table 3.4.6-1 includes the requirement to adjust observed leakage rates for these isolation valves to function maximum dif ferential pressure values when testing at lower test pressures. This adjustment ensures conservatism since observed' leakage rates may be less at lower test differential pressures when the leakage channel opening remains constant with increasing pressure. Observed leakage rates are adjusted using the square root of the ratio of the function maximum dif ferential pressure and the actual test dif ferential pressure. Function maximum differential pressure is that differential pressure which the.
isolation valve is subjected to under normal Plant conditions during power operations. For isolation valves not subjected to a differential pressure under normal Plant conditions during power operations, RCS pressure is assumed on the high pressure side of the valve.
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Trojan Nucicar Plant Document Control Desk Docket 50-344 August 12, 1988 License NPF-1 Attachment Page 1 of 7 EVALUATION OF TROJAN EMERCENCY CORE COOLINC SYSTEM (ECCS) CHECK VALVE INTEGRITY I. INTRODUCTION Reactor Coolant System (RCS) pressure is isolated from the Safety Injection System (SIS) by two in-series check valves for each RCS loop (A-D). When failure of one valvo in a pair could go undetec-ted due to system configuration, periodic testing is required to verify valve integrity and reduce the overall risk of an inter-system Loss-of-Coolant Accident (LOCA). Test requirements are specified in Trojan Technical Specification Surveillance Require-ment 4.4.6.2.2 and Periodic Operating Test (POT) 2-4, "ECCS Pressure Boundary and Accumulator Check Valve Leakage Inservice Test".
This evaluation will demonstrate that test results obtained during the 1988 refueling outage, and Plant parameters monitored during subsequent power operation, provide adequate assurance of valve integrity, thereby reducing the probability of gross valve failure and consequent intersystem LOCA and establishing adequate justi-fication for continued operation through Operating cycle 11.
II. EVALUATION ,
This section summarizes the tests performed and Plant parameters monitored which relate to check valve leakage. Evaluation of the information focuses on identifying the current condition of the i check valves and assessing the consequences of postulated failures I of either in-series check valve. i A. Summary of Test Results During the 1988 Refueling Outage Prior to entering Mode 2 after the 1988 refueling outage, ;
testing was conducted in accordance with POT-2-4 and Technical Specification Surveillance Request 4.4.6.2.2 for pressure isolation valves identified in Table 3.4.6-1. Initial test results for Valves 8948A, B, and D [ Emergency Core Cooling System (ECCS) first off cold-leg check valves) exceeded maxi-mum allowable leakage limits specified in Table 3.4.6-1.
Total meascred leakage from Valves 8819A, B, C, and D (SIS second off cold-leg check valves) exceeded the maximum allow-able limit for each individual valve. Since the taat did not s:
Trojan Nuclear Plant Document C9ntrol Desk Docket 50-344 August 12, 1988 License NPF-1 Attachment Page 2 of 7 measure leakage through each valve, it is uncertain if allow-able limits were exceeded. Numerous leakage tests were con-ducted for each valve (s). A summary of results for POT-2-4 testing is as follows:
Valve Leakage (mpml Test Pressure (psi) 8819A-D 3.0 600 8819A-D 2.7 610 8819A-D 2.8 640 8819A-D 2.8 800 8819A-D 3.0 630 8819A-D 3.1 630 8819A-D 3.2 640 8819A-D 3.2 650 8819A-D 3.2 640 8819A-D 3.5 '640 8819A-D 3.5 640 8819A-D 3.0 1140 8948A 6.5 1575 8948A >10.0 1875 8948A 8.1 1725 8948A 3.4 1120 8948B 5.8 1875 8948B 6.4 1575 8948B 1.3 1120 8948D 3.3 1125 8948D 6.4 1525 8948D 7.0 1880
[Above data taken from POT-2-4 data sheets dated July 4-6, 1988. Other valves tested per POT-2-4, which are not listed
- above, met the allowable leakage limits.)
The Plant was placed in cold shutdown in accordance with I Action Statement 3.4.6.2.c. After flushing water through the ,
check valves and revising the test procedure to allow checking the 8819 valves two at a time rather than four at a time, testing was again performed per Temporary Plant Test (TPT) 266, "Safety Injection System First off and Second Off Check Valve Leakage Inservice Test", for the subject valves at a test differential pressure less than the initial test differential pressure. These results met the maximum allowable leakage limits specified in Table 3.4.6-1. The final test results are ,
summarized below:
Valve Leakage (Rpm) Test Pressure (psi) 8948A 1.4 350 1 8048B 1.4 350 l l
Trojtn Nuc1ccr Plcnt Document Contrcl Desk ,
Docket 50-344 August 12, 1988 ;
License WPF-1 Attachment ,
Page 3 of 7 l l
Valve _ Leakate (nom) Test Pressure (psi) 8948D 2.1 350 8819A&B 2.2 360 8819C&D 1.4 360 (Above data were taken from TPT 266 data sheets dated July 7, 1988.)
ECCS check valve testing in 1988 per POT 2-4 and TPT-266 has shown that the 8819A-D valves have a total leak rate in the ,
range of 2.7 to 3.6 spm. Tests have been perforned with a -
differential pressure ranging from 360 psi to over 1000 psi. . ,
Test results do not indicate an increasing leak rate with !
increasing pressure.
B. Plant Parameters and Special Test During Power Operation
- 1. Observation of Plant parameters on August 4 and 5, 1988 l and review of Plant loss are summarized below. Pressure ;
on the accumulator side of check valva 8948 (Figure 1) is l st accumulator pressure of 650 psis. The SI pump dis- !
charge header is also at accumulator pressure, while the !
SI pump suction header is approximately 30 pais. This indicates that at least one 8819 valve may be leaking and l that the SI pump discharge valves are not leaking. The ;
pressurizer relief tank (PRT), which accepts discharge ;
from ECCS relief valves, has not given indication of !
relief valve lifting. This indicates that ECCS check j valve back leakage is not pressurizing piping systems to i their relief valve setpoints (1750 pois for the SIS).
Accumulator levels have remained steady, ,
Based upon a review of data sheets from POT-1-3, "Leakage Evaluation", RCS leakage and inventory checks have shown l the following:
Identiflod Leakage Unidentified Leakage Date (Allowed 10 p.pm) (Allowed 1 tom) 7-16-88 0.12 0 )
7-19-88 0.06 0.05 7-22-88 0.05 0.02 7-29-88 0.18 0 7-31-88 0.12 0 8-03-88 0.05 0 l
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Triojcn Nucic0r Plant Document CIntesi Desk !
, Docket 50-344 August 12, 1988 ,
- License NPF-1 Attachrtent page 4 of 7 [
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- 2. On August 5,1988. TPT 267, "Safety Injection System First :
< Off and Second off Check Valve Integrity Verification", i was performed to provide additional assurance of check. :
valve integrity. The test was performed while in Mode 1 !
at normal operating prossure and temperature. The first 3 portion of the test satisfactorily verified first off check valve integrity by again measuring pressure between the first off and second off valves to be at SI accumu-lator prescure. The second portion of the test provided-a ,'
j qualitative indication of leakage across the second off i check valves. Specific leakage measarements were not 4
possible due to system lineup constraints while in -
Mode 1. Test results indicated that some leakage across the second off valves does exist, but that it is nst in excess of about 3.5 spm at a differential pressure t.?
approximately 625 psi. i C. Evaluation and Conclusions
! 1. Test results for Valves 8948A, B, and D and 8819A-D !
I summarized in II.A above, indicate that no gross failures [
j exist (ie, the check valves function and limit back leak- ,
age as shown in the above test results). ;
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- 2. Evaluation of ECCS First Off Check Valves (8948A, B, and D) 1
! Observed Plant parameters indicate that there is no leak- ;
ago through the first off check valves (8948A, B, and D) l as summarized below:
I I a. SI accumulator levels have remained steady.
- b. Accumulator pressure, as well as the pressure of the !
piping between the in-series check valves has remained (
] steady at the expected value of approximately 650 psig. ;
! If 8948 valves leaked, the piping between the in-series -
check valves could rise to RCS pressure. >
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- c. The PRT level has remained steady indicating that the ,
ECCS relief valves for lower pressure syst. ems have not l lifted. This indicates that first off check valve leakage is not occurring simultaneous with second off l check valves. l d
Since there is no leakage, the ECCS first off check valves ;
meet their design function for SI accumulator availability ,
as described in Section 6.3.2.2.7.4, "Accumulator check i valves (swing-disc)", of the Final Safety Analysis Report. !
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7 Trojen Nuclsar Plent Document Control Dssk Docket 50-344 August 12, 1988 License NPF-1 Attachment
.Page 5 of 7
- 3. Evaluation of SIS Second Off Check Vatves (8819A-D)
As s'tated in II.C.1, recent testing his verified check valves are functional for both first off (8948) and second off (8819) check valves. Therefore, the postulated fail-ure condition for these in-series check valves is that the lowest leakage valve is presumed to: fail and the highest leakage series valve is exposed to its mcximum differen-tial pressure. For the current condition, this case is a failure of one 8948 valve and the presumed leakage of either 8819A or 8819B at a rate of 2.2 gpm with a pressure differential of 360 psi. This assumes all the leakage is through one valve. The maximum possible pressure differ-ential across the 8819 valve is RCS prescure at the cold-leg (2280 psig) with no pressure in the SI system (0 psig) for a differential pressuro of 2280 psi. Assuming a con-stant area opening, the flow rate is ratioed by the square" root of the pressure difference:
Maximum Flow = Maximum AP x Test Flow Test AP MaximumFlow=/2280 x 2.2 = 5.5 spm
/ 360 l l
This is well within the 20 gpm relief capability of each ,
of three relief valves in the SI System (Note: Figure 1 !
shows one of the relief valves, PSV-8851). Also, this leak rate from the RCS is well within the normal RCS-makeup capabilities and thus, a.LOCA (as defined in Part 50.46 to Title 10 of the Code of Federal Regulations (10 CFR 50.46)) would not occur.
A more realistic assumption for pressure differential assumes that the SI pump discharge check valves (8922A/B) remain fuactional and the SI system oressurizes to the l PSV-8851 relief valve setpoint of 1750 psig. Then the I pressure differential across 8819A or B is 2280 psig - 1750 psig = 530 psi. The leak rate would then be:
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1 2.2 X 530 = 2.7 spm .
360 1 This is again well within the 20 gpm relief capability of I PSV-8851 (FSAR Table 6.3-9).
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Tsojen Nuclsar Plant Docum:nt Control D2sk Docket 50-344 August 12,'1988 License NPF-1 Attachment.
Page 6.of 7.
-III. CONCLUSIONS Review of test.results and current operating Plant' parameters
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shows that there is no undetected gross failure of any of the evaluated in-series check valves. The evaluation.has also demon-strated that a singlo failure of one of the in-series check valves would not J ead to the overpresuurization and. rupture of the SI system. Therefore, the postulated overpressurization and rupture of a' lower pressure system connected to the RCS (intersystem LOCA) is not credible based on test leak rates for check valves at Trojan. Continued operation through Operating Cycle 11 is i justified.
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Trojan Nuclotr Pitnt Docum:nt Control Dzsk Docket 50-344 August 12, 1988
-Licenso NPF-1 Attachment Page 7 of 7 'l
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- 8956 .
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SV-8851 REACTOR *
- e S I I 8948 8819 8922 s
SAFEIY INJECTION PUMP EMERGENCY CORE COOLING SYSTEM (ECCS) CHECK VALVE CONFIGURATION FIGURE 1
- l. . . .
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION l
l In the Matter of ) i
) I PORTLAND GENERAL ELECTRIC COMPANY, ) Docket 50-344 THE CITY OF EUGENE, OREGON, AND ) Operating License NPF-1 PACIFIC POWER & LIGHT COMPANY )
)
(TROJAN NUCLEAR PLANT) )
CERTIFICATE OF SERVICE I hereby certify that copios of License Change Application 169 to the Operating License for Trojan Nuclear Plant, dated August 12, 1988, have J been served on the following by hand delivery or by deposit in the United States mail, first class, this 12th day of August 1988:
Mr. Bill Dixon State of Oregon ,
Department of Energy 625 Marion St NE Sslem OR 91310 j l
Mr. Michnol J. Sykes !
Chairman of County Comissioners l Columbia County Courthouso j St. Helons OR 97051 ;
h, re - l
'S . A. Bauer, Manager '
Nuclear Regulation Branch Nuclear Safety & Regulation i 1
Subscribed and sworn to before no this 12th day of August 1988.
- dsd b r/m Notary Public of O/cgon OMyCommission hW[e -
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