ML20058D845

Separate Rept Re Local Leak Rate Test Results from Core 19/20 Refueling Outage

ML20058D845
Person / Time
Site:	Yankee Rowe
Issue date:	10/29/1990
From:	YANKEE ATOMIC ELECTRIC CO.
To:
Shared Package
ML20058D772	List: ML20058D770 ML20058D842 ML20058D845
References
NUDOCS 9011060290
Download: ML20058D845 (17)

Text

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SEPARATE REPORT CONCERNING I

LOCAL LEAK RATE TEST RESULTS I

FROM THE CORE 19/20 REFUELING OUTAGE I

,M SC ET S 0 Docket No,'50 29 I

h(AestEngineerN Ab Prepared by:

1 Reviewed byt en En nee ing g

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Approved by:

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React ngineeging Managet Approved by:

97( W+a M-M 44#'<#~

Technical Director *

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Approved by:

Plant Superintendent I

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i-TABLE OF CONTENTS SECTION llTLE PAGE Table of Contents i

list of Attachments ii References lii 1

Purpose 1.1 1

2 Summary 2.1 3

Test Results and Analysis 3.1 3.1 Test Methodology 3.1 l

3.2 Test Results

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3.3 Analysis of Results 3..'

3.4 Corrective Actions 3.4 I

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ATTACHMENTS TITLE i

1.A Yankee Nuclear Power Station Technical Data 2.A Licensee Event Report 89 001 1

2 3.A

'As Found' and ' As lef t' Type B and C Results I

1 3.8 Description of Type B and C Maintenance Requests Completed During the Core 19/20 Ref ueling Outage

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iii-REFERENCES 1.

10 CFR Part 50, Appendix J. Primary Reactor Containment Leakage Testing for Water cooled Power Reactors, January 1, 1989 2.

Procedure No. OP 4702, Vapor Container Type B & C Penetration I

Tests, Yankee Nuclear Power Station, Surveillance Test Procedure 3.

ANSI N45.4-1972, American National Standard, Leakage Rate Testing of Containment Structures for Nuclear Reactors, March 16, 1972 4.*

ANSI /ANS 56.8 1987 Containment System Leakage Testing Requirements, January 20, 1987.

5.

Licensee Event Report 89 001, Rev. O, Yankee Rowe Nuclear Power Station, Event Date 01/03/89, Report Date 02/09/89 6.

' June 1990 Reactor Containment Building Integrated Leak Rate Test And Periodic local Leak Rate Tests' Report, Yankee Atomic Electric Company, License No. OPR 3, Docket No. 50 29.

E 7.

EDCR #86'310, ' Water Clean up System Tie in Connections and VC Penetrations *, Yankee Atomic Electric Co., Nuclear Services Division I

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This document used on1 as a guideline.

Any reference to this document in no way imp les compliance to its requirements.

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SECTION 1 PURPOSE DuringtheCore19/20RefuelingOutabeattheYankeeNuclearPower Station (YNPS), the cumulative Type and C leak rate exceeded the allowable limit established by the plant's Technical Specifications.

In accordance with the requirements of 10 CFR 50. Appendix J. Paragraph I

V..B. this report presents the results of the Local Leak Rate T'.sts (LLRT's) performed during the outage and provides an analysis of those results.

YHPS is owned and operated by the Yankee Atomic Electric Company (YAEC).

Specific plant information and technical data is contained in Attachment 1.A.

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3-ATTACHMENT 1.A YANKEE NUCLEAR POWER STATION TECHNICAL DATA A.

Plant Information:

Owner end 0perator Yankee Atomic Electric Comrany Plant

' Yankee Nuclear Power Station Location Rowe, Massachusetts Containment Type Uninsulated Spherical Steel Shell Elevated Above Grade l

Operating License Number DPR 3 Docket Number 50 29

' i B.

Technical 01111, Containment Nct Free Air Volume 860,000 cu ft.

Design Pressure 31.5 psig Containment Max. Allowable Pressure 34.5 psig Calculated Peak Accident Pressure 31,6 psig Design Temperature 249 of

)

Calculated Peak Accident Temperature 245 *F Containment Diameter 125 ft.

Containment Min. Shell Thickness 7/8 in.

Containment Shell Material A300 Carbon-Silicon Steel, Class A 201, Grade B

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o SECTION 2

SUMMARY

- on alf Containment Isolation Barriers (Cbe B and C tests were performed Durin the Core 19/20 Refueling Outage T

's) required to be tested by Yankee-Rowe Technical Specification (YRTS) Table 3-61.

The LLRT's were performed in accordance with Reference 2 using the pressure decay method.

All the barriers were tested using air or nitrogen to pressurize the test boundary to greater than or equal to the calculated peak accident pressure of 31.6 psig.

The $. ant's' cumulative Type B and C leak rate is required by YRTSor 0.12 wt%/ day w l

3.6.

2.b to be less than er equal to 0.60 L the plant is in Modes 1 4 If this leak rate limit is exceeded the leak rate must be reduced to less than or equal to the limit before the Main I

Coolant System temperature can be increased above 200oF.

On January 3. 1989 it was determined that the cumulative Type B and C E

leak rate had exceeded the 0.12 wt%/ day limit. At the time of the fI:

discovery the cumulative leak rate was 0.1492 wt%/ day and the plant was in Mode 5.

Repairs to the valves deemed to be the major contributors to the excessive,eak rate were completed.

The leakage was reduced to less than the acceptat,le limit prior to entering Mode 4.

Reference 5 was-issued on. February 9, 1989 to report that the "As-Found" cumulative leak rate had exceeded the YRTS 3.6.1.2.b limit. The

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specifics of this event and the associated corrective actions are

~g delineated on Attachment 2.A. a copy of Reference 5.

The"AsFound"cumulativeTheBandCleakratefromtheCore19/20 Refueling Outage was determ ed to be 0.1762 wt%/ day.

This leak rate I

includes the initial results of the LLRT's performed on the four valves installed in a spare containment penetration during the refueling per Reference.7 for einew system. The initial -leak-rate for the four valves t

was, determined to be 0.0413 wt%/ day.

Subtracting this leakage from the

- As Found" cumulative leak rate results in a leak rate.of 0.1349

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wt%/ day. This is the cumulative-leak rate that existed at the end of

-i Core 19 operations.

This end-of life Core 19 leak rate is well below L, or 0,20 wt%/ day.

L t

'is the maximum. allowable leak rate that assures containment integrity.,

-Maintaining the' containment-leak rate at less than or equal to L ensures that site boundary doses would be within the 10 CFR 100 limits-1 during a design bases accident. Therefore, the health-and safety of the public<would not have been affected if a design basis accident had oc.urred dur.ing Core 19 operations, kI3 At' the end of' the outage. following completion of all repairs, the " As i

.Left? cumulative Type B and C leak rate had been reduced to 0.0848

.wt%/ day. This :is substantially below the technical specification limit f

of 0.12 wt%/ day.

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t 5-I ATTACHMENT 2.A LICENSEE EVENT REPORT 89-001 r

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==w TANKEE NUtt[AR POWER 51 A110N. Rowe. HA. 01367 o16Iotoqet 01219 i loel0 t 2 Type B & C Test Coseined Lookoge Exceeded I

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!v On January 3,19B9, at !?00 hours0 days <br />0 hours <br />0 weeks <br />0 months <br />, while in Mode 5, during Containment Type B L C surveillance testing, the contined leakage rate for all penetrations subject to Type B & C testing onceeded-the Technical Specification 3.6.1 2.b-limit of 0.12 wtX/24 hours. The total leak rate was 0.1492 wtX/24 hours. The primary contributors to the combined Isakage rate exceeding the T.S. limit were i^

til the Containment Service Water supply line isolation valve, SW-TV-412. and (El the Containment Heating Steam returg line isolation valve. HC-TV-409. The root cause was the leakage of SW.TV-412 due to eucessive'esea' of the valve disk i

and seat and the leakage of HC-TV-409 due to a buildup of a black gritty

i deposit in the portion of the valve body that accomecdates the stem follower.

The overall Type B & C leakage rate was reduced within the allowable I

Technical Specification.*talt prior to entering Noce 4 Similar occurrences i

have been recorted in I R B4-06, 84-10 and B7-II.. The health and safety of the putilic were not affected as a result of this event.

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IT ATTACHMENT 2.A LICENSEE EVENT REPORT 89-001

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.aseuty anansa ne eucas,.nman 538 Les apua.eaa est poet is YANKEC NUCLEAR POWER STATION "tt?.?

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On January 3,1989, at 1700 hours0.0197 days <br />0.472 hours <br />0.00281 weeks <br />6.4685e-4 months <br />. while in Mode 5, conducting Containment Type B & C surveillance testing, in accordance with plant procedure DP-4702, 9

the combined leakage rate for all penetrations subject to Type B & C testing exceeded the Technical Specification 3.6.1.2.b limit. At the time of this discovery the total leak rate was 0.1492 wt%/24 hours. The Technical Ii Specification !! alt is 0 12 wt%/24 hours. The primary contributors to the combined leakage rate esteeding the T.S. limit were (1) the' Containment Service Water supply line inclation valve, SW-TV-412, and (El the Containment Heating Steam return line isolation valve, HC-TV-409 Previous instances of enceeding I.

Technical Specification Type B & C acceptance criteria have been reported in LER 84-06. 04-10 and 07-11.

Corrective action taken to reduce the excessive Ieakage to within acceptable limits consisted of removing and disassembling SW-TV-412. A new disc was installed and the seat ring was lapped. The valve was reassembled L

using new studs, nuts and gasket and the valve was then retested. The Type C

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retest indicated an acceptable improveecat in the valve leaka's from 0.0369 to 0,0116 wt%/24 hours. The root cause of the excessive valve leaka(# has been attributed to wearing of the valve disc and seat. Also. HC-TV-409 was removed I:

and disassembled. Foreign material was removed from the section of the valve body that accommodates the stem follower. The disc seating surface was machined and lapped. The valve body seat was also lapped. The salve was reassembled and then retested. The Type C retest indicated a significant E'.

cause of the excessive valve leakage has been attributed to the buildup of a leprovement in the valve leakage from 0.0252 to 0.0009 wt%/24 hours. The root black gritty deposit from the heating steam in the section of.the valve body.

that accommodates the stem follower. Thus the follower could not be fully i

estended into the valve body which prevented the disc from properly seating..

Additionally, the plant will evaluate replacing SW-TV-412 and the downstream test boundary isolation valve,' SW-V-1060 and HC-TV-409 and the upstream test boundary isolation valve, HC-V-061, with valves whtCh'are better I;

't designed to undergo air leakage testirig The reduction in leakage. rates associated with the repair of SW-TV-412 and -

HC-TV-409 reduced the overall Type D & C leakage rate.to within the allowable Technical Specification limit prior to entering Mode 4 The overall leak rate is now 0.0848 wt%/24 hours.

The monimum Type B & C leakage rate of 0.1492 wt%/24 hours was within the Technical Specification limit for containment integrity (0.20 wt%/24 hoursh lI1 which ensures-that site boundary doses would be within 10 CFR 100 limits.

. Therefore, the health and safety of the public were not affected as a result of '

this event.

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SECTION 3 TEST RESULTS AND ANALYSIS 4

3.1 Test Methodoloov_

Each CIB required to be Type B or C tested as delineated on YRTS Table l

3.6-1 was addressed by a separate Attachment to Reference 2.

Although

'L the test procedure for each barrier is unique, the method for performing each of the Type B and C tests is essentially the same.

i All LLRT's were conducted using the pressure decay method. The pressure decay method of leak testing uses a test rig to measure the reduction in test boundary pressure over a measured period of time.

For liquid filled systems the test boundaries were drained prior to pressurization.

Each test boundary was pressurized to greater than the calculated peak I.

accident pressure of 31.6 psig using air or nitrogen.

The test pressure was applied to each barrier in the direction that the accident pressure would be applied to the barrier (i.e. from the

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The downstream side of each barrier was either vented to atmosphere or, where this was not possible, the test boundary.

pressure was increased to assure that greater than 31.6 psig was applied across the barrier throughout the test period.

Once the test boundary was pressurized as described above, the air or nitrogen was disconnected from the test volume.

Initial values of test boundary pressure and. temperature were taken and the test period

.I started. The pressure decay of the test boundary was generally measured over a test period'of fifteen minutes.

At the end of the test period, final measurements of test boundary pressure and temperature were made.

If the test boundary pressure dropped precipitously during a test because of a large leak rate, then the time it took-for boundary pressure to fall from the initial pressure to 32 psig was measured.

calculated by applying the ideal The test bours v mass leak rate a

I gas law to t s measured pressure. cay rate.

The volumes of each of the

. test boundaries had previously been calculated so the mass leak rate could be directly calculated as follows:

Mass Leak Rate (wt%/ day) -;

144 inch'

'60 min' '24 hr'100%)

[y P3 - P, x ft hr day,

g 3

53.35 1bf (186. 903 1 bm)

] mR s

Y 8~ 'x Hass Leak Rate (wt%/ day) =

t, - t2 T + T'~

i 3

(0.481) 2

.f s

where:

Pi - initial pressure of test volume (psig)

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P, - final pressure of test volume.(psig) l t: - initial-time (min.)

t, - final time (min. )

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Ti - initial temperature of test volume (*R)

T, - final temperature of test volume (*R)

V-calculated test volume (f t')

186,903 lbm - calculated minimum post-LOCA dry air mass in containment Each test boundary's mass leak rate was calculated from the test data.

The total test boundary leakage was assigned to the barrier being tested.

The individual test results were entered into a computer I--

computer program sums the individual leak rates for each tested barrier program that calculates the cumulative Type B and C leak rate. This using the most recent leak rate entry for that barrier.

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3.2 Test Results.A lists the 'As Found* and 'As left' test results for the Core 19/20 Refueling Outage for all containment barriers required to be I

included in the cumulative leek rate. The differences between the 'As Found" and "As left' data reflect any repairs, removals and reinstallations, adjustments, and/or retests of the barriers that occurred during the outage.

The Maintenance Request numbers for any repairs made to the CIB's during the outage are also listed on.A.

The 'As Found' cumulative Type B and C leak rate for the Core 19/20 I-Refueling Outage was 0.1762 wt%/ day. This leakage rate exceeded the limit for cumulative Type B and C leakage.

On January 3, 1989 0.60 L, determined that the leak rate had exceeded the 0.12 wt%/ day it was limit.

At the time of the discovery the cumulative leak rate was 0.1492

'IJ wt%/da and the plant was in Mode 5.

Repairs to the valves deemed to be

.the ma or contributors to the excessive leakage rate were completed and the le kage was reduced to less than the acceptable limit prior to entering Mode 4.

Reference 5 was issued on February 9, 1989 to report that the "As Found" cumulative leak rate had exceeded the YRTS 3.6.1.2.b limit.

The specifics of.this event and the associated corrective actions are I

delineated on Attachment 2.A..a copy of Reference 5.

After completion of all repairs the 'As Left" cumulative leak rate for the Core 19/20 Refueling Outage was 0.0848 wt%/ day.

3.3 Analysis of Results-As: stated in the previous section, the "As Found" cumulative leak rate for all CIB's tested during the outage was 0.1762 wt%/ day.

This leak rate-includes the initial LLRT results for the four valves installed in spare containment penetration "FC' per Reference 7 during the refueling outage.

The initial LLRT's for these four valves were performed in I

accordance with Attachment NH to OP 4702 on' December 24, 1988. As can be.seen from Attachment 3. A, the total "As Found" leak rate on this date for the four valves was 0.0413>wt%/ day. This leak rate can be

'i subtracted from " As Found". cumulative leak rate of 0.1762 wt%/ day to determine the cumulative leak rate that existed at the-end of Core 19

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operations. This results in a end-of-life Core 19' leak rate of 0.1349 wt%/ day. This leak rate exceeds the technical specification limit for acceptable cumulative Type B and~C leakage: however, this does not mean I

that containment integrity was not maintained during Core 19 operations.

3.2 I

the overal1~ containment leak rate less than or equal to L Maintainink/dayguaranteesthatthetotalcontainmantleakagewillnol or 0.20 wt exceed the value assumed in the Yankee Rowe accident analysis at the peak accident ressure.

This assures the ublic health and safet is adequately pro ected in the event of a des gn basis accident.

YR S 3.6.1.2.b limits the cumulative Type B and C leak rate to less than or e ual to 0.60 L or 0.12 wt%/ day. This added margin of conservatism is i cluded to account for additional leakage pathes not identified by I

LLRT's. This margin also allows for possible degradation during the plant's operating cycle of containment barriers which are required to be Type B & C tested.

If the cumulative Type B and C leak rate exceeds the 0.12 wt%/ day limit then the Action Statement of YRTS 3.6.1.2 requires the leak rate to be restored to less than or equal to 0.12 wt%/ day prior to increasing Main Coolant System temperature above 2000F.

e B and C leak rate, based The Core 19 beginning of life cumulative Typ/ day. As stated previously I

on the Core 18/19 LLRT data, was 0.0761 wt%

the Core 19 end-of life cumulative Type B and C leak rate was 0.1349 wt%/ day. Thus containment integrity was maintained throughout Core 19 operations.

The health and safety of the public was not adversely

'I-af fected as a result of exceeding the 0.12 wt%/ day technical

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specification limit during a portion of Core 19 operations.

airs were completed to reduce As described in Section 3.4 all needed rep %/ day limit.

All repairs were 3

the cumulative leak rate below the 0.12 wt completed prior to entering Mode 4 at the end of.the refueling outage.

The 'As Lef t" cumulative Type B and C leak rate at the end of the Core 19/20 Refueling was 0.0848 wt%/ day.

As reported in Section 4 of Reference-6. the "As Found* cumulative Type B and C leak rate for the Core 20/21 Refueling Outage was 0.0718 wt%/ day.*

Thus acceptable leak rate margins and containment integrity I;

'were maintained throughout Core 20 operations.

3.4 Corrective Actions j

I The. major contributors to the "As Found* combined leak rate exceeding the technical specification limit were the Vapor Container (VC) Service Water supply line isolation valve SW TV 412, and the VC Heating Steam return line isolation' valve, HC TV-409.

Immediate corrective actions to I

repair these valves were implemented as described on Attachment 2. A.

The result-of these repairs was a 0.0496 wt%/ day reduction in the.

. combined.' leak rate of the two valves..

Where deemed appropriate, additional repairs to other CIB's were made

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'throughout-the course of the outage to further reduce the cumulative leak rate.

The associated YNPS'Haintenance Request (MR) numbers for these repairs are' listed next to the affected' barriers on Attachment I

3.A..B lists these Maintenance Requests, identifies the specific components that were repaired, and briefly describes the-repairs that were completed. The net effect of all the repairs undertaken during the outage was to reduce the cumulative Type B and C an leak rate to 0.0848 wt%/ day.

g As a long term corrective action the plant committed-in Reference 5 to evaluate replacement of HC TV-409 and SW-TV-412 and their associated

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test boundary valves, HC-V 861 and SW-V 1060, with valves which are

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better designed,to preclude air leakage.

Currently the plant is evaluating the replacement of these four valves during the next

-refueling outage with ball valves.

The type ~of. ball valves to be

~As can be seen from these figures the cumulative leak rate actually decreased during Core 20 operations.

This is mostly attributable to the large improvement in the leakage rate of SW-L:

TV 412 brought about by the flushing of its valve seat throughout Core 20 operations.

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The plant revent air leakage.t of other trip valves installed are specifically designed tohas used this type of valve fo with good results.

t boundary composed of SW-A review of the Type C data below f or the teshave not demonstrated TV-412 and SW-V-1060 indicates that these valves rates.

consistently high leak V_C Service Water S. d u

Trio Valve _Tyne ( C9]Jt Leak Pete (wt%/ day 1 Test Da.1.g 0.0026 11-27-82 0.0006 04-24-84 0.0 10-19-85 0.0 I

06-03-87 0.0369 01-03-89 0.0116 01-04-89 0.0 1989 were an 06-24 90 I

l in January, f the VC The high leak rates measured through the va ves ical cleaning o anomaly most likely attributable to the chemService Water The these tests.

of a gritty particulate material cleaning resulted in the dispersaThis material apparently prevented the valve I

L from properly seating and was the cause of t e the ' ore 20 operating cycle notheir latest Ty throughout the system.

After being continuously flushed during leakage was measured through the valves dur'ng>f both SW-TV-412 and SW I

Thus, although replacement i nperative that the it is not in June, 1990.

1060 is still being considered, refueling outage.

modifications be completed during the next f our valves, the plant I

In addition to the possible replacement of theseis evalu l es to further reduce The VC Heating Steam Supply trip boundary valve, HC V-1200, the cumulative Type B and C leak rate.

Consideration valve, HC-TV-413, and its associated testl de air leakage. valves during I

were not specifically designed to prec uis being given to replac ith ball the next refueling outage.

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- - - -,....... - _, _ _..... ATTACHMENT 3.A "AS F0VHD" AND "AS LEFT" TYPE B AND C RESULTS Core 19/20 Leak Rate Repaired Per OP-4702 Test Boundary Description (wt%/ day)

Maintenance Request As Found As left A

Cavity Fill Line (CS-V-601) 0.0 0.0 g

B, Port A VC Hydrogen Vent (8 Valves) 0.0110 0.0091 89-111,69-112 h

B, Part B VC Hydrogen Vent (HV-50V-42) 0.0001 0.0001 J

C VC Service Water Return (TV-408) 0.0 0.0002 89-076,89-149 l

0 Fuel Chute Blank Flange 0.0001 0.0

_]

E Fuel Chute Pumpback Blank Flonge 0.0 0.0 l

F Main Coolant Sample (TV-206) 0.0001 0.0001 G

Neutron Shield Tk Sample (TV 207) 0.0001 0.0001 E 1 VS Leak off (Tv-204 & Sv-223) 0.0001 0.0001 J

Main Coolant Vent (TV-203) 0.0001 0.0001 g K VC Htg Stm Return (TV-409) 0.0252 0.0')09 88-2489, 89-031/081 k L Part A Air Part. Tap (VD-V-1108) 0.0 0.0 L, Part B Air Part. Tap (VD-V-1107) 0.0 0.0 M

Main Coolant Drains (TV-202) 0.0004 0.0004 0

Component Cool. Return ( TV-205) 0.0001 0.0001 0

VC Drain (VD TV-209 & CC-SV-227) 0.002 0.002 R

ECCS Retirc (651, 516 & 517)

O.0014 0.0004 89-109 s

Electrical Penetrations 0.0531 0.0524 89-024 V

8' Air Bypass (HC-V-602) 0.0001 0.0001 B

W VC Personnel Hatch & CA-V-755 0.0 0.0025 88-571, 88-2443 E

Demin Water Blank Flange 0.0 0.0 Z

AA LP Vent Hdr Blank Flange 0.0001 0.0001 CC NST Tell Tales (VD-V-754) 0.0001 0.0 89-0001-1 DD VC Air Charge (746 & 1277) 0.0001 0.0001 FF Component Cool Supply (TV-208) 0.0001 0.0001 GG VC Htg Stm Supply (TV-413) 0.0035 0.0035 HH VC Service Water Supply (TV 412) 0.0369 0.0116 89-016 II, Part A 30" Blank Flange (lolet) 0.0 0.0 11, Part B 30" Blank Flange (Outlet) 0.0 0.0 JJ VC SV Disch Hdr (TV-214) 0.0001 0.0001 I

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2-ATTACHMENT 3,A "AS FOUND" AND "AS LEFT" TYPE B AND C RESULTS Core'19/20 Leak Rate Repaired Per OP 4702 Test Boundary Description (wt%/ day)

Maintenance Request Att.

(MR)

As Found As left l

% KK LPST SV Disch Hdr (W/ 917) 0.0 0.0 N L'L HC Bleed Line (CH LCV 222) 0.0 0.0

.kHM,PartA VC Breathing Air (BA V 63) 0.0001 0.0001 HM,' Part'B' VC Breathing Air (BA-V 61)-

0.0 0.0

- lNN,Part'A'-

Water Clean Up (VD V-1171) 0.0001 0.0001-

' ] NN, Part B Water Clean Up (WC-V 621) 0.0001' O.0001

$NN.PartC Water Clean Up (VO V 1170) 0.0001 0.0001 h NN, Part D Water Clean Up (WC V-622) 0.0.41 0.0003 88 2440 i

j Totals 0.1762-0.0848 5'

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& ATTACHMENT 3.B-DESCRIPTION OF TYPE B AND C MAINTENANCE RE0 VESTS COMPLETED DURING THE CORE 19/20 REFUELING OUTAGE

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REPAIRED 8ARRIER EFFECTED DESCRIPTION OF COMPONENT BY Rt0 AIR WORK REPAIRS89-111 HV V 5 VC Hydrogen Vent Removed reach rod &

Valves (8 valves) bonnet lapped disc seat: cleaned internals:

reassembled 1

1 89 112 CA V 688 VC Hydrogen Vent Removed bonnet &

Valves (8 valves) gate: lapped gate seat; cleaned s

internals:

reassembled 89 076 SW TV 408 VC Service Water Removed actuator:

I-Return (SW TV-408) disassembled valve:

cleaned internals:

changed seats:

reassembled valve &

actuator 89 149 SW-TV-408 VC Service Water Changed actuator:

Return (SW TV-408) verified proper stroke 89 031-HC TV-409 VC Heating Steam Hachined & lapped t

Return (HC-TV-409) valve & disc seats l

88 2489--

HC V 861 VC Heating Steam Removed bonnet:

Return (HC-TV-409) inspected internals:

lapped seats:

reassembled I,89-081 HC V-861 VC Heating Steam Removed, inspected,-

i Return (HC TV-409) reassembled internals of valve-89-109 SI H0V 517-ECCS Recirculation Adjusted-valve Line Isolation packing: checked t

'~

Valves stroke of actuator

~

89-024 Blister 6E VC Electrical-Removed penetration Penetrations 8 Penetrations canisters; installed

& 15 blank-flanges: tested blank flanges I{

~

88 571.

VCPH VC Personnel Hatch Installed new door and CA V 755 gasket 88 2443 VCPH VC Personnel Hatch Cleaned, adjusted, &

and CA-V-755 lubed outer door pistons 89-0001 1 VD-V-754

~NST-Tell Tales Removed bonnet:

=

(VD-V-754) replaced diaphram:

reassembled valve I

1

2 ATTACHMENT 3.B DESCRIPTION OF TYPE B /.ND C MAINTENANCE RE0 VESTS COMPLETED DURIN9

.THE CORE 19/20 REFVELING_0UTAGE MR #

REPAIRED BARRIER EFFECTED DESCRIPTION OF COMPONENT BY REPAIR WORK REPAIRS89-016 SW TV-412 VC Service Water Disassembled &

Supply (SW-TV-412) cleaned; installed new disc, studs, nuts, & gasket:

verified proper actuator stroke i

l' 88 2440 WC V 622 Water Clean-up.

Disassambled, cleaned l

(WC V-622)

& inspected valve:

lapped seats:

l' reassembled I

s 1"I

)

l:

1 I

LI ll K:

L,.

e l.

.. -