05000249/LER-1988-001, :on 880105,discovered That 41 Individual Type B & C Local Leak Rate Tests Performed During Previous Refueling Outage Had Individual Testing Intervals Exceeding 24 Months.Review Conducted

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:on 880105,discovered That 41 Individual Type B & C Local Leak Rate Tests Performed During Previous Refueling Outage Had Individual Testing Intervals Exceeding 24 Months.Review Conducted
ML17199U086
Person / Time
Site: Dresden Constellation icon.png
Issue date: 01/29/1988
From: Eenigenburg E, Lizalek J
COMMONWEALTH EDISON CO.
To:
NRC OFFICE OF ADMINISTRATION & RESOURCES MANAGEMENT (ARM)
References
88-063, 88-63, LER-88-001, LER-88-1, NUDOCS 8802080228
Download: ML17199U086 (17)


LER-1988-001, on 880105,discovered That 41 Individual Type B & C Local Leak Rate Tests Performed During Previous Refueling Outage Had Individual Testing Intervals Exceeding 24 Months.Review Conducted
Event date:
Report date:
Reporting criterion: 10 CFR 50.73(a)(2)

10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications
2491988001R00 - NRC Website

text

-

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LICENSEE EVENT REPORT (LER)~

Fac11 Hy Nam-.: ( 1) ocket Number (2)

~I Page (3) 0 I SI 0 I 0 I 0 I 21 41 9 1 I of I 1 I Dresden Nuclear Power Stat;on. un;t 3 6

Tltle (4)

LLRT Intervals D;scovered ;n Excess of 24 Months on an lnd;v;dua) Test Basis Due to Procedural Deficienc~

Event Date (5)

LER Number ( 6)

Reoort Date l7l Other Facilities Involved C8l Month Day Year Year I.II. Sequential ~~~ Revision Month Day Year Facility Names I Docket Number(s)

  • ~~~ Number Ill Number
  • Dresden Unit 2 ol SI OI OI ol 21 31 7 o I 1 o Is 8 I 8 8 I 8 --- o I o I 1 -- olo-o I 1 21 9 81 8 N/A OI SI ol ol ol I I OPERATING

.THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10CFR MODE (9)

(Check one or more of the following) (11}

N 20.402(b) 20.405(c)

SO. 73(a)(2)( iv)

73. 71(b).

POWER

20. 405(a)( 1 )( i)

S0.36(c)( 1)

SO. 73(a)(2)(v)

73. 7 l(c)

LEVEL o I 9 I 20.405(a)(l)(ii) -

S0.36(c)(2)

S0.73(a)(2)(vi;)

_Other (Specify

( 10) 6 20.40S(a)(l)(iii)..L S0.73(a)(2)(i)

S0.73(a)(2)(vi;;)(A)

n Abstract

~~~~~~~~~~~*~~~~~~~~~~~~*~~~ _. 20. 40S (a)( 1 )( iv)

S0.73(a)(2)(;i) -

S0.73(a)(2)(vii;)(B) below and in

~~~~~~~~~~~~~~~~~~~~~~~~~~ _

20.405(a)(l)(v) 50.73(a)(2)(i;i) -

S0.73(a)(2)(x)

Text)

LICENSEE CONTACT FOR THIS LER ( 12}

Name TELEPHONE NUMBER A~EA CODE Jerrv Lizalek. Technical Staff Ena;neer. X-421 8 I 1 I S 91 41 21 -I 21 91 21 o COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFAC-REPORT AB.LE ~~~~~~~ CAUSE SYSTEM COMPONENT MANUFAC-REPORTABLE ~~~~~~

TUR ER TO NPRDS

~~~~~~~

TURER TO NPRDS

~~~~~~

I I I I I I

~~~~~~~

I I

I I

I

~~~~~

I I I I I I 1111111 I*

I I

1 I

"/ "/"/"/"/

SUPPLEMENTAL REPORT EXPECTED (14)

Expected Month I Da~ I Year Subm;ss;on

- IYes (If ves ccrnolete EXPECTED SUBMISSION DATE)
  • 1XI NO Date ( lS)

I I I I I ABSTRACT (Limit to 1400 spaces, ;.e, approximately fifteen s;ngle-space typewritten l;nes) (16)

On January S, 1988, w;th Unit 3 at 9S.8% rated thermal power, wh;le reviewing past Local Leak Rate.Testing (LLRT) records, 1t was discovered that 41 individual type "B" and "C" LLRTs performed during the previous

  • refueling outage had individual testing intervals in excess of 24 months.

As 10 CFR SO Append;x*J requires that this testing be perfonned,on intervals no greater than 24 months, concern was raised as to whether compliance with 10 CFR SO Appendh J had been demonstrated.

The. root cause was attributed to a procedural deficiency within the scheduJ;ng of the station survei1lance track;ng program; ;t has been the station's practice to enter the test date of the last ;ndiv;dual LLRT as the entire surveillance completion date.

Corrective actions ;ncluded performing as many of the overdue LLRTs as poss;ble during reactor operation

  • and upgrading the track;ng program to verify the LLRT intervals on an ind;v;dual test basis in the future.

Management changes were also implemented regard;ng admin;stration of the surveillance program. A schedular exempt;on was requested fran the Office of Nuclear Reactor *Regulat;on (NRR) for the remci;ning untested valves. Further review revealed that 28 individual Type "B" and "C" LLRTs perfonned during a past Unit 2 refuenng outage had ;ndrvidual testing ;ntervals in excess of 24 months, although current compliance was verified. Safety s;gnHkance is min;ma1 s;nce a primary contairvnent Integrated Leak Rate Test (type "A" test) was performed on Unit 2 and un;t 3 pr;or to startup w;th sat;sfactory results.* A prev;ous event involving a surve;11ance program procedure def;c;ency ;s reported by LER 87-18 on Docket #OS0249.

8802080228 880129 PDR ADOCK 05000249 S

PDR

"'--------------=L=I=CE EVENT REPORT LER FACIL.ITY NAME ( l)

DOCKET NUMBER (2)

TEXT CONTINUATIO LER NUMBER 6 Page (3)

Dresden Nuclear Power Station Unit 3 Year

~~~ Sequential ~~~

Ill Number Ill 0

5 0

0 0

2 4 9 8 8

0 0

Revision Number 0

0 0 2 OF l 6 TEXT PLANT ANO SYSTEM IDENTIFICATION:

General Electric Boiling Water Reactor - 2527 foU.Jt rated core thermal power.

Energy Industry Identificaton System (EIIS) codes are identified in the text as [XX].

Nuclear Conmitment Tracking System* (NTS) tracking code numbers are identified in the text as (XX).

EVENT IDENTIFICATION:

While reviewing past Local Leak Rate Testing (LLRT) records it was discovered that 41 individual Type 118 11 and "C" LLRTs perfornied during the previous Unit 3 refueling outage had individual testing intervals in exces~* of 24 months due to a surveillance program procedure deficiency. Further review revealed that 28.

individual Type 118 11 and "C" LLRTs perfomied during a previous Unit 2 refueling outage had individual testing intervals in excess of 24 months, although current compliance was verified.

A.

CONDITIONS PRIOR TO EVENT

Unit: 3 Event Date:. January 5, 1988 Reactor Mode:

N Mode Name: Run Reactor Coolant System (RCS) Pressure: 997 psig B.

DESCRIPTION OF EVENT

Event Time:

1000.hours Power Level: 95.8%

On January 5, 1988 with Unit 3 at 95.8% rated thennal power, a review of LLRT.records for Type 118 11 and "C" testing performed during the previous Unit 3 refueling outage (End of Cycle 9) was conducted. This.review revealed that.the LLRT program had begun on September 27, 1985 and was completed on August 13, 1986. Therefore, 41 individual LLRTs were discovered to have individual testing intervals in excess of 2 years~ As 10 CFR 50 Appendix J requires testing intervals of no greater than 24 manths, concern was raised as to whether compliance with 10 CFR 50 Appendix J had been demonstrated. These individual LLRTs which were currently in excess of the 2 year interval are listed in Table *1.

This review of LLRT testing history was a result of investigation into a previous related surveillance interval problem.

On September 30, 1987, it was discovered that a fire barrier penetration surveillance had exceeded its 18 month Technical Specification surveillance interval in 1985, although current compliance was verified. The root cause of the fire barrier penetration surveillance problem was found to be a procedural deficiency within the surveillance tracking program.

The problem identified with the tracking program resulted from certain surveillances being misclassified as due on a refueling outage basis rather than on a fixed-interval basis.

As part of the corrective actions for this event, which was reported to the Nuclear Regulatory Conmission (NRC) on Licensee Event Report #87-018 on Docket #050249, a review was performed in order.to properly classify these types of surveillances *.

LIC EVENT REPORT LER TEXT CONTINUATIO FAi:l~ITY NAME ( 1)

Do* ET NUMBER (2)

LER NUMBER Page Pl Year

~~~ Sequential

~~~ Revision Dresden Nuclear Power Station Unit 3 0

5 0

0 0

2 4 9 8 8

Ill Number Ill NUIOOer 0

0 0

0 0 3 OF TEXT Further review revealed that a schedular exemption had been previously granted by the NRC concerning a past Unit 3 refueling outage LLRT program.

Upon perfonning a detailed review of this previous exemption and 10 CFR 50 Appendix J, the current LLRT concerns were raised..

Upon discovering that the 41 individual Unit 3 LLRT dates had exceeded a 24 month interval, the Corporate Office was contacted for assistance in detennining if a violation had occurred.

The Dresden site Senior Resident Inspec~or (USNRC, Region III) was also contacted and notified of this infonnation.

Further review revealed that 28 individual Type 118 11 and "C" LLRTs performed during a past Unit 2 refueling outage had individual testing intervals in excess of 24 months, although current compliance was verified. These individual LLRTs are listed in Table 2.

C.

APPARENT CAUSE OF EVENT:

Local Leak Rate Tests at Dresden Station are perfonned under the direction of five separate_

Technical Staff surveillances 1 ist_ed in Table 3.

Leak Rate Tests of similar type are grouped together and are controlled under one of these five surveillances.

As many as 85 Local Leak Rate Tests are listed under one survei_llance*and scheduling of these surveillances is tracked by the stat ion computer survei 11 ance program.

The completion of any of these Local leak Rate surveillances can take the duration of the entire refue.ling outage due to availability of equipment for test1rig and conflicts with inaintenance. It has_been the past practice of Dresden Station to enter the date of the last test perfonned as the completion date of the affected surveillance. This single completion date rather than individual test completion dates would then be used by the computer program to schedule the next surveillance interval for all of the penetrations affected. Since all of the individual tests listed under the surveillance were completed before the end of the refueling outage, and primary containment isolation volumes were not required to be operable until the end of the refueling outage, it has been past station practice to use the last individual test date as the representative completion date to be used for surveillance tracking.

Although the 69 primary containment isolation pathways (41 currently for Unit 3 and 28 previously for Unit 2) listed in Table 1 and_ Table 2 have individual test dates that have exceeded the 2 year 10 CFR 50 Appendix J test interval, the completion dates listed for each affected surveillance in the computer tracking system do not exceed the 2 year limit. Therefore, the cause has been attributed to a surveillance program procedure deficiency.

The Dresden Unit 2 and Unit 3 Technical Specifications Section 4.7.A.2. require perfonnance of Leak Rate Testing in accordance with 10 CFR 50 Appendix J. Therefore, this report is submitted in accordance with 10 CFR 50 (a)(2)(i)(B), which requires the reporting of any operation or condition prohibited.by the Technical Specifications.*

D.

SAFETY ANALYSIS OF EVENT:

The safety significance of the Unit 3 discrepancy is mitigated by the following factors:

1.

On July 26, 1986, a successful primary containment Integrated Leak Rate Test (ILRT) was performed on Unit 3.

The total "as left" leakage was.5874 weight %/day and included Local Leak Rate Test results for all systems not vented. This was well within the Technical Specification maximum allowable limit of 1.6 weight %/day. Twenty-nine of the 41 penetrations affected were vented during the ILRT.

6

LIC~ EVENT REPORT (LER) TEXT CONTINUATIO~

FACH,ITY NAME ( 1)

DOCKET NUMBER (2)

LER NUMBER (6)

Page Pl Year

~~~ Sequential ~~~ Revision Dresden Nuclear Power Ill Nurrber Ill Number Station. Unit 3 o I 5 I o I o I o I 21 41 9 8-1 8 o I o I 1 -

o I 0 0 14 OF TEXT

2.

A review was perfonned of the "as found" and "as left" leakage rates for the 41 _affected penetr.ations during the last three refueling outages. The results are listed in Table 4.

In general, almost all "as found" leakages for the 41 penetrations were small, with the worst "as found" leakage of 87.9 SCFH found on valves 220-57A and 62A during the 1984 refueling outage.

This is well below the Technical Specification limit of 493. 116 SCFH.

The total "as found" leakage for all 41 penetrations is listed at the bottom of Table 3 for each outage.

The worst total "as found" leakage was 214.25 SCFH which was also below the Technical Specification 1 imit.

The safety significance of the Unit 2 discrepancy is mitigated by the following factors:

  • 1.
  • performed on Unit 2.

The total "as left" leakage was.4732 weight %/day and included Local Leak Rate Test results for all non-vented systems. This was well within the Technical Specification.maximum allowable.limit of 1.6 weight %/day.

Fifteen of the 28 penetrations affected were vented during the ILRT.

2.

A review was perfonned of the "as found" leakage rates for the 28 penetrations affected during the end of Cycle 10 refueling outage. It was determined that the "as found" maximum pathway leakage was 185.008 SCFH; this is well below the Technical* Specification limit of 493. 116 SCFH.

Additionally, the "as found" minimum pathway leakage for all Type "8" and "C" LLRTs was determined to be 236.789 SCFH.

This is also below the Technical Specification limit of 493. ll6 SCFH.

3.

Currently, Dresden Station is in compliance with the required Appendix J 24 month test interval for all Unit 2 Type "8" and "C" primary containment penetrations on an individual basis.

E.

CORRECTIVE ACTIONS

The inmediate corrective actions included tabulating the already past due 41 Unit 3 LLRTs (Table 1) and determining which of these could have LLRTs performed during reactor power operation.

Subsequent to a review, it was determined that 22 of the 41 past due surveillances could be tested.

Table 5 identifies which primary containment/valve volumes could be tested. These volumes. were inmediately Local Leak Rate Tested and the identified leakage can be seen on Table 5.

  • The 19 remaining volumes could not be tested at reactor power for the reasons stated in Table 6.

Secondly, a review was performed to determine which individual LLRTs would becane overdue if not performed prior to the scheduled.March 26, 1988 outage start date. It was determined that 43

- additional volumes (Table 7) would becane overdue prior to the March 26, 1988 outage start date. A review of these volumes revealed that 37 could be Local Leak Rate Tested during power operation.

These were inmediately tested and their _leakage can be seen on Table 8. The 6 remaining volumes could not be tested at reactor power for the reasons stated in Table 9.

A total of 25 penetrations/volumes that were overdue or would be overdue prior to the March 26, 1988 outage start date could not be tested. A request for schedular exemption from the 10 CFR 50, Appendix J, Type "B" and "C" test interval for these 25 tests was submitted on January 10, 1988 in a letter from the Dresden Nuclear Licensing Administrator to the Director of Nuclear Reactor Regulation.

In addition to these short term corrective actions, the following long term corrective actions have been initiated to restore full confidence in the Technical Specification Surveillance Program.

They are as follows:

11 6

EVENT REPORT LER FACl~ITY NAME (1)

ET NUMBER (2)

TEXT CONTINUATI0~1 LER NUMBER (

Page (3)

Year

~~~ Sequential ~~~ Revision Dresden Nuclear Power

  • Station Unit 3 0

5 0

0 0

2 4 9 8 8

"//'/

Number

"//'/

Number 0

0 0

0 0 5 OF TEXT

1.

A Conrnonwealth Edison Task Force has been assigned to perfonn an independent review of the surveillance program's adequacy. Appropriate improvements wi 11 then be implemented (249-200-88-00101).

2.

An INPO assistance visit has been scheduled for February 8, 1988.

The purpose of this visit is to provide the station and Task Force with INPO industry input on effective surveillance programs (249-200-88-00102).

3.

In order to improve the effectiveness and control of the surveillance program, management changes concerning its administrati.on were initiated. Overall control of surveillance tracking will be assigned to the Work Planning Department, which is headed by an Assistant Superintendent. Scheduling of surveillances will be integrated into daily work. scheduling meetings in order to ensure they are perfonned in a timely manner.

Each working group will be held responsible to comply with appropriate scheduling requirements (249-200-88-00103).

F.

PREVIOUS OCCURRENCE:

Licensee Event Report #87-018 on Docket #050249, Fire Stop 18 Month Surveillance Interval Exceeded Due to Procedural Deficiency.

As stated in Section B., this event also involved a surveillance program deficiency. Corrective actions involving reclassification of the surveillances as due on a fixed-interval basis as opposed to an operating cycle basis did not correct the underlying problem with Local Leak Rate Testing procedures in t.hat they were logged as completed upon performance of the entire group of tests.

G.

COMPONENT FAILURE DATA

Since no component failures were associated with this event, this section is not applicable.

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Dresden Nuclear Power Station Unit 3 TEXT DTS-250-1 DTS-1600-1 DTS-1600-2 DTS-1600-4*

DTS-1600-15 TEXT CONTINUATIO LER NUMBER b Year

~~~ Sequential

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Number

~~~ Revision

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Number Page (3) 0 5

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0 2 4 9 8 8

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OF 1 6 TABLE 3 LOCAL LEAK RATE TEST SURVEILLANCE PROCEDURES Main Steam Isolation Valve Local Leak Rate (Dry) Test

. Local Leak Rate Testing of Primary Containment Isolation Valves Local Leak Rate Testing Procedure Bellows Seal Penetrations Local Leak Rate Testing For Electrical Penetrations Local Leak Rate Testing of Double Casketed Seals

FACILITY NAME ( 1)

Dresden Nuclear Power Station Unit 3 TEXT VALVES East Torus Drain West Torus Drain 4720 & 4721 4722 & Check 220-57A & 62A 1101-1 & 15 1101-1 & 16 1501-18A & 19A 1501-20A & 38A 1501-18B & 19B 1501-20B & 38B 1402-4A,8A,25A,36A 1402-4B,8B,25B,36B 1301-1 & 2 220-44'& 45 Penetration Bellows X-105A X-105B X-105C.

X-1050 X-106 X-107A X-108A X-lllA X-lllB X-128 X-123 X-124 X-125 X-126 X-138 X-147 X-149A X-149B X-107B X-109A 3702 & 3799-126 1301-17 & 20 220-1 & 2 2301-4 & 5 2301-35 & 36 Drywell Head Hanway TOTAL AS FOUND 0

AS EVENT REPORT LER TEXT CONTINUATIO ET NUMBER (2)

LER NUMBER Year

~~~ Sequential Iii' Number 5

0 0

0. 2 4 9 8 8

0 0

TABLE 4 UNIT 3 HISTORICAL LLRT DATA 1982 1984 FOUND AS LEFT AS FOUND AS LEFT SCFH SCFH SCFH SOFH 0

0 0

0 0

0 0

0 0

0 4.14 4.14

.24

.24 3.92 3.92 25.9 25.9 87.9 9.33

.96

.96 li.115 11.115 23.14 23.14 0

0 0

0 1.185 1.36 7.83 7.83 4.05 4.05

.38

.38 0

0 2.48 2.48 0

0

.56

.56 0

0 0

0

.918

.918 0

0 2.15 2.15 0

0 0

0 ALL BELLOW SEALS ALL BELLOW SEALS INITIAL FINAL INITIAL FINAL 4.8 4.8 8.33.

8.33

.289

.289 1.07 1.07

.649

.649 6.49 2.84 7.67 7.67 1.021 1.021 2.12 2.12 0

0 79.82 19.73 0

0 0

0 74.87 SCFH 214.25 SCFH Page Pl

~~~ Revision i'i'i' Number 0

0 1 0 OF 1 6 1985 AS FOUND AS LEFT SCFH SCFH

.1991

.1991

.2109

.2109 2.62 2.62

. 1. 949

1. 949 10.85 10.85 6.01 6.01 11.81

. 11. 81 1\\.534 4.534 1.94 1.94

.5917

.5917 1.294 1.294 3.44 3.44 1.821 1.821

.9365

.9365

.033

.033 3.974 3.974 0

0 0

0 0

0 0

0

864

.864 0

0 3.024 3.024 0

0 0

0 0

0 0

0 0

0 0

0

.432

.'132 0

0

.216

.216 1.08 1.08 0.0 0.0

.o 0

0 0

1. 74
1. 74 54.328
1. 736 3.44 3.44 1.437 1.437 0

0 118. 77 SCFH

FACIL.ITY NAME ( l)

Dresden Nuclear Power Stat ion Unit 3 TEXT LER NUMBER LICE'I EVENT REPORT LER TEXT CONTINUATIO DO ET NUMBER (2)

Year

~~~

Ill Sequential ~~~

Number IT/

Revision Number 0

5 0

0 0

2 4 9 8 8

0 0

Penetration Number X-105A.

X-105B X-105C X-105D X-106 X-107A X-107B X-108A X-109A X-lllA X-lllB X-128 X-123 X-124 X-125 X-126 X-138 X-147 X-149A X-149B X-313A X-313B TABLE 5 UNIT 3 OVERDUE LLRTs THAT HAVE BEEN PERFORMED AS OF 1/5/88 Voluine Being Tested Main Stearn Main Steam Main Stearn Main Steam Main Steam Drain Feedwater Feedwater Iso. Cond. Steam Iso. Cond. Condensate Shutdown Cooling Shutdown Cooling HPCI Steam RBCCW Inlet RBCCW Outlet Vent From DW Vent To OW SBLC Rx head Spray*

Core Spray Core Spray E. Torus Drain Vlvs w. Torus Drain Vlvs 0

0 1

Recent Test Leakage SCFH 3.412 0

0 0

0 0.604 0

0 0

1.9 0

0 0

0 0

0 0.562 0

0.432 0.734

  • o~o 0.0 Page (3)

OF l

6

VALVE OR PENETRATION TES'I: PROBLEMS A0-4720 & 4721 A0-4722 & Check Valve 220-57A & 62A 1101-1 & 15 1101-1 & 16 1501-lSA & 19A 1501-20A & 38A 1501-188 & 198 1501-208 & 388 H0-1402-4A, SA, 25A, 36A H0-1402-48, SB, 258, 368 H0-1301-1 & 2 A0-220-44*& 45 Drywall Pneumatic Supply Isolation Drywall Pneumatic Supply Isolaition Feedwater Check Isolation standby Liquid Control Injection Standby Liquid Control Injection LPCI Loop A LPCI Loop A

.LPCI Loop B LPCI Loop B Core Spray Injection Core Spray Injection Isolation Condenser

  • steam Supply Recirc Loop Sample Isolation REASON FOR NOT TESTING AT POWER T~sting pathway could cause main steam isolation valves to close.

Testing pathway will cause main steam isolation* valves lo close.

Testing would require isolation of the reaclor feedwater system and a unit shutdown.

Test isolation valves are located in the inerted drywall and are inaccessible during operation.

Test isolation valves are located in the inerted drywall and are inaccessible during operation.

Testing pathway would require entry into a limiting condition of operation.

Testing pathway would require entry into a limiting condition of operation.

Testing pathway would require entry into a limiting concHtlon of operation.

Testing pathway would require entry into*a limiting condition of operation.

Testing pathway would require entry irito a limiting condition of operation. One.test boundary would be pressurized at 1000 psig preventing valid testing.

Testing pathway would require entry into a limiting condition of operation.

One test boundary would be pressurized at 1000 psig preventing valid testing.

Testing pathway would require entry into a limiling condition of operation.

One test boundary would be pressurized at 1000 psig preventing valid testing.

  • one test boundary would be pressurized at 1000 psig preventing valid testing.
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VALVE OR PENETRATION TEST PROBLF.HS H0-3702 &

3799-126 RBCCW to Drywell Coolers Drywell Head Hanway Double Gasket Seal A0-1301-17 & 20 Isolation Condenser Vents H0-220-1 & 2 Hain steam Line Drain

. H0-2301..,.4 & 5 HPCI Steam Supply H0-2301-35 & 36 HPCI Pump Suction REASON FOR NOT TESTING AT POWER Testing of pathway results in loss of drywell cooling and could cause a reactor scram.

Test boundary inaccessible during operation.

One test boundary will be pressurized at 1000 psig preventing valid testing.

One test boundary*will.be pressurized at 1000 psig preventing valid testing.

Testing pathway would require entry into a limltlng condition of operation.

Testing pathway would require entry into a limiting condition of operation.

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DATE PREVIOUSLY CAN BE TESTED RECENT m*

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DESCRIPTION

PERFORMED DURING POWER TEST DATE Ill w-,

A0-3-301-156A,157 Scram Discharge Volume Ven ls and Drains 1-11:..86 N

None A0-3-301-160A,161A Scram Discharge Volume Ven ls and Drains l-ll-86 N

None 1c::

0 :c A0-3-301-1568,1578 Scram.Discharge Volume Ven ls and Drains 1-11-86 N

None z.

Ill H

A0-3-301-1608,1618 Scram Discharge Volume Venls and Drains 1-11-86 N

None 1-':l A0-2001-5,6 Drywall Equipment Drain Sump Discharge 1-11-86 y

1-9-88 VJ A0-2001-105,106 Drywall Floor Drain Sump Discharge 1-21-86 y

1-9-88 L'

A0-1599-61,62 Torus to Condenser Drain 1-21-86 y

1-8-88 0

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X-136A Tip Monitor Flange 2-2-86 y

1-9-88 r-X-1368 Tip Monitor Flange 2-2-86 y

1-9-88 L'

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m X-136C Tip Monitor Flange 2-2-86 y

1-9-88 L'

Cz:J X-136D Tip Monitor Flange 2-2-86 y

1-9-88 1.11 X-136E Tip Monitor Flange 2-2-86 y

1-9-88 x

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X-136F 2-2-86 y

0 c:: m 3: z A0-205-2-4 & Flange Reactor Head Cooling 2-6-86 H

None 1-':l CX1 -l 0

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A0-205-2-7 & Flange Reactor Head Cooling 2-6-86 H

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A0-2499:-1A,2A CAM H2 to Sensor 2-16-86 y

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A0-2499-18,28 CAM H2 to Sensor 2-16-86 y

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A0-2599-4A,5A ACAD Vent 2-23-86 y

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A0-2599-48,58 ACAD Vent 2-23-86 y

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X-200C Electrical Penetrations 3-2-8.6 y

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X-2018 Electrical Penetrations 3-2-86 y

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1-9-88 1-':l L' Al m 0

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X-202F Electrical Penetra.tions 3-2-86 y

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X-202W Electrical Penetrations 3-2-86 y

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1-9-88 Ill X-20U Electrical Penetrations 3-2-86 y

1-8-88 t<

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1-8-88

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X-204H Electrical Penelralions 3-2-86 y

1-8-88 X-204M Electrical Penetrations 3-2~86 y

1-9-88 H

'° X-204N Electrical Penetrations 2:

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1-9-88 1-':l

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X-204Q Electrical Penetrations 3-2-86 y

1-9-88 tZJ Ill X-204S Electrical Penetl."ations 3-2-86

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Ill

.y 1-9-88 0, 0 X-2058 Electrical Penetrations F:

j 3-2-86.

y 1-9-88 M0-1601-57,58,59 H2 Makeup To Drywall 3..,.16-86 y

1-9-88 X-313A E. Torus Drain 3-18-86 y

1-9-88 A

X3138 w. Torus Drain 3-18-86 y

1-9-88 Ill 0

Ill w

°'

LICE~ EVENT REPORT (LER) TEXT CONTINUATION~

FACH.ITY NAME ( 1)

DOCKET NUMBER (2)

LER NUMBER (6)

Page {3)

Year

~~~ Sequential ~~~ Revision Dresden Nuclear Power

"//"/

Number

"//"/

Number Station. Unit 3 o I 5 I o I o I o I 21 41 9 a I a -

o I o I 1 -

o I 0 1 15 OF 11 6 TEXT TABLE 8

. UNIT 3 PENETRATIONS TESTED SO AS NOT TO EXCEED 24 MONTH INTERVAL Penetration Leakage Number Volume Being Tested SCFH X-200C

.LV Power & Control 8.130 X-201B HV Power 1.016 X-202B CRD Indication 0.103 X-202BB CRD Indicator

  • o.o X-202D HV Power o.o X-202F Thermocouples 8.965 X-202J Neutron Monitor o.o X-202N Neutron Monitor o.o X-202Q Instrumentation o.o X-202S CRD Indicators 2.571 X-202W CRD Indicators 3.074 X-203B HV Power o.o X-204A HV Power 2.376 X-204E Neutron Monitor o.o X-204H Neutron Monitor 0.103 X-204M LV Power 6.125 X-204N CRD Indicator o.oo X-204Q..

CRD Indicator 0.561 X-204S LV Power & Control 14.062 X-20SB CRD Indicator o.o X-136A Tip Flange o.o x..:.136B Tip Flange

.082 X-136C Tip Flange 0

X'."1360 Tip Flange 0

X-136E Tip Flange

.661 X-136F Tip Flange 0.0 X-313A E. Torus Drain o.o X-313B w. Torus Drain o.o NA 1S99-61 & 62 3.77 X-126,304 1601-S7, 58, 59 0.421 X-118 2001-5 & 6

.854 X-117 2001-105 & 106

.281 X-202V 2499-lA & 2A o.o X-204B 2499-lB & 2B o.o X-202V 2599-2A & 23A 3.285 X-125,318 2S99-4A & SA 9.235 X-125,318 2599-4B & SB 1.558

LICE


'-r"---------'==

EVENT REPORT LER TEXT CONTINUATION FAC~LITY NAME (1)

Dresden Nuclear Power Station Unit 3 TEXT.

DOCKET NUMBER (2) 0 5

0 0

0 2

LER NUMBER Year

~~~

'Ill 4 9 8 8

TABLE 9 6

Page Sequential

~~~ Revision Number Ill Number 0

0 0

0 1 6 OF LIST OF PATHWAYS NOT YET OVERDUE THAT COULD NOT BE TESTED (UNIT 3)

Valve or Penetration 205-2-4 & Flange 205-2-7 & Flange 3-301-156A, 157 3-301-160A, 161A.

3-301-1568, 1578 3-301-1608, 161B Reactor Head Cooling Reactor Head Cooling Scram Discharge Volume Scram Discharge Volume Scram Discharge Volume Scram Discharge Volume Test boundary inaccessible during operation Test boundary inaccessible during operation Testing of pathway could cause a reactor scram Testing of pathway could cause a reactor scram Testing of pathway could cause a reactor scram Te$ting of pathway could cause a reactor scram p) 1 6

Common~h Edison Dresden Nucl~ower Station R.R. #1 Morris, Illinois 60450 Telephone 815/942-2920 EDE LTR 1188-063 U.S. Nuclear Regulatory Commission Document Control Desk Washington, D.C. 20555 January 29, 1988 Licensee Event Report /188-001-0, Docket /1050249 is being submitted as required by Technical Specification 6.6, NUREG 1022 and 10 CFR 50.73(a)(2)(i)(B).

Een' en Station ger Dresden Nuclear Power Station EDE/kj 1 Enclosure cc:

A. Bert Davis, Regional Administrator, Region III File/NRC File/Numerical 0192k