05000423/LER-1996-012-02, :on 960515,containment Leakage Exceeded Tech Spec Limit.Caused by Boric Acid Residue.Performed Flush of Line to Remove Boron Deposits

:on 960515,containment Leakage Exceeded Tech Spec Limit.Caused by Boric Acid Residue.Performed Flush of Line to Remove Boron Deposits

ML20117M216
Person / Time
Site:	Millstone
Issue date:	06/14/1996
From:	Brothers M, Temple W NORTHEAST NUCLEAR ENERGY CO., NORTHEAST UTILITIES SERVICE CO.
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
B15752, LER-96-012-02, LER-96-12-2, NUDOCS 9606180275
Download: ML20117M216 (9)

LER-1996-012, on 960515,containment Leakage Exceeded Tech Spec Limit.Caused by Boric Acid Residue.Performed Flush of Line to Remove Boron Deposits

Event date:
Report date:
Reporting criterion:	10 CFR 50.73(a)(2)(i)(B), Prohibited by Technical Specifications 10 CFR 50.73(a)(2)(ii)(A), Seriously Degraded
4231996012R02 - NRC Website

text

m, Mai nc om.. mye rerry u, weaora, er Northeast Utilities System

,,.o. n>> 128 Waterford, Cr 06385-0128 (201) 447-1791 June 14,1996 Docket No. 50-423 B15752 Re: 10CFR 50.73(a)(2)(i)(B) and 50.73(a)(2)(ii)(A)

U. S. Nuclear Regulatory Commission Document Control Desk Washington, DC 20555 a.

This letter forwards Licensee Event Report 96-012-00, documenting an event that occurred at Millstone Unit No. 3 on May 15,1996. This LER is submitted pursuant to 10CFR50.73(a)(2)(i)(B) and 50.73(a)(2)(ii)(A).

Very truly yours, NORTHEAST NUCLEAR ENERGY COMPANY i

b N

M. H. Brothers

\\

ggOlb Unit Director, Millstone Unit No. 3)

Attachment: LER 96-012-00 cc:

T. T. Martin, Region l Administrator A. C. Cerne, Senior Resident inspector, Millstone Unit No. 3

- V. L. Rooney, NRC Project Manager, Millstone Unit No. 3 9606180275 960614 1h PDR ADOCK 05000423 j

V S

PDR os3422 9 nov. : 9s

NRC FORM 366 U.S. NUCLEAR REGULATORY Commission APPROVED BY OMB NO. 31504104 EXPlRES 04130/98 go.gg Nr oNthi sffS Es (Nito N n^rio^"MTf?"^!L'a T=Ps".1%5a ^?uda

'R Ten"'/oT.^3MK2s".^o"J.T.T.,N" &

5!%^Ylitac1WKIMT1". 0?sfRM"&"na""'-

LICENSEE EVENT REPORT (LER) in?A (See reverse for required number of dig:ts/ characters for each block)

F ACluTY N AME lil DOCKET NWSER (2)

PAGE (3)

Millstone Nuclear Power Station Unit 3 05000423 1 of 8 fifLI I4)

Containment Leakage in Excess of Technical Specification Limits Due to Valve Leakage EVENT DATE (5)

LER NUMBER (6)

REPORT DATE (7)

OTHER FACfLITIES INVOLVED (8)

MONTH DAY YEAR YEAR SEQUENTIAL REvlsioN MONTH DAY YEAR FACIUTY NAME DOCKET NUMBER NUMBER

'^ " "*"'

05 15 96 96 012 00 06 14 96 OPERATING THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 1o CFR 5: (Check one or rnore) (til MODE (9) 5 2o.22o1(b) 20.2203(a)(2)(v>

X so.73(a)(2)(i)

So.73( )(2)(viii)

POWER 20.2203(a)(1) 2o.2203(a)(3)(i)

X so.73(a)(2)(ii)

So.73(a)(2)(x)

LEVEL (lo) 000 20.22o3(a)(2)(i) 20.2203(a)(3)(ii)

So.73(a)(2)(iii) 73.71 2o.2203(a)(2)(ii) 20.2203(a)(4)

So.73(a)(2)(iv)

OTHER 20.22o3(a)(2)(iii)

So.36(c)(1)

So.73(a)(2)(v)

Spgi y in Ab t elow y

20.2203(a)(2)(iv)

So.36(c)(2)

So.73(a)(2)(vii)

W LICENSEE CONTACT FOR THis LER (12)

NAME TELEPHONE NUMBER Onclude A,ea Codel William J. Temple, Nuclear Licensing Supervisor (860)437-5904 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13)

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLE

CAUSE

SYSTEM COMPONENT MANUFACTURER REPORTABLT TO NPRDS TO NPRDS SUPPLEMENTAL REPORT EXPECTED (14)

EXPECTED MONTH DAY YEAR i

BMISSION YES NO 08 30 96 g

(If yes. complete EXPECTED Submission DATE).

ABSTRACT (Limit to 1400 spaces, i.e., approximately 15 single-spaced typewritten lines) (16)

On May 15,1996, while performing containment penetrat.5n Local Leak Rate Tests (LLRTs) with the plant in cold i

shutdown, the leak rate for containment isolation valves exceeded the Technical Specification limit. The leakage through a High Pressure Safety injection (SlH) valve, when combined with previously determined leakage for other penetrations, exceeded the Technical Specification limit of 0.6L. (294,800.2 secm).

Subsequent to May 15,1996, the LLRTs performed on other isolation valves revealed additional leakage that further contributed to exceeding the Technical Specification limit. The valves that were major contributors to the total leak rate were those located on a containment Recirculation Spray (RSS) line, a Hydrogen Recombiner (HCS) line, a containment Quench Spra) (OSS) line, the Containment Purge (HVU) supply line, and a Low Pressure Safety Injection (SIL) line.

On May 29,1996, during the performance of LLRTs, an additional"as found" leakage resulted in the total exceeding an analyzed condition. The leakage through valves on the HVU line was unquantifiable due to excessive leakage. Thus, an immediate notification was made on May 29,1996, pursuant to 10CFR50.72(b)(2)(i) for an event found while shutdown, that if found while operadng, would have resulted in an unanalyzed condition that significantly compromises plant safety.

The cause is attributed to boric acid buildup during the test sequence on the QSS and RSS valves. Foreign particulate is the attributed cause on the SlH valve, and seal overtightening is the cause on the HVU valve. The corrective action was flushing for the boron and particulates, and readjusting for the HVU seal. To prevent recurrence the testing and flushing procedures will be revised. A supplement LER will provide corrective action on the HCS and SIL valves.

i NRC FORM 366 (4 95)

U.S. NUCLEAR REoVLATORY COMMISSION (4 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAoE (3)

YEAR SEQUENTIAL REvlsloN Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 2 of 8 00 012 96 TEXT Uf more space is required, use additional copies of NRC form 366A) (il) 1.

Descriotion of Event On May 15,1996, while performing containment penetration Local Leak Rate Tests (LLRTs) with the plant in cold shutdown, the leak rate for containment isolation valves exceeded the Technical Specification limit. The leak rates i

through a high pressure safety injection isolation valve and a low pressure safety injection isolation valve, when combined with previously determined leakage for other penetrations, exceeded the Technical Specification 3.6.1.2.b Type C limit of 0.6L. (294,800.2 secm).

i a

The subsequent LLRTs performed on other containment isolation valves revealed additional *as-found* leak rates that further contributed to exceeding the Technical Specification limit. The isolation valves that were major contributors to the total leak rate were those associated with: a containment quench spray line, a high pressure safety injection line, a containment Recirculation spray line, a hydrogen Recombiner line, the containment purge supply line, and a low j

pressure safety injection line. Each of the major contributors to the totalleak rate is described below.

o On May 10,1996, the LLRT leakage through the inboard isolation check valve (3QSS*V4) on the Containment Quench Spray line (12-inch) penetration #100 was 156,000 secm. This leakage, added to the *as-found* total for all other valves as of May 10,1996, hvi not exceeded the Technical Specification limit, and by itself is not reportable.

However, it is included in this LER because it was a major contributor to the total eventually exceeding the Technical Specification limit five days later. See Figure 1. The outboard isolation valve would have maintained containment integrity.

j o On May 15,1996, the combined LLRT leakage through both: the inboard isolation check valve (3SIL*V13) on the Low Pressure Safety injection line (6-inch) penetration #94; and through the inboard isolation check valve 1

(3SlH*V24) on the High Pressure Safety injection 1:ae (2-inch) penetration #98, was 43,900 secm. This event, added to the total for all previous LLRTs exceeded the Technical Specification limit. See Figure 2. The outboard l

isolation valve would have maintained containment integrity.

o Later on May 15,1996, the LLRT leakage through the inboard isolation check valve (3RSS*V3) on the Containment Recirculation Spray line (10-inch) penetration #107 was 111,100 sccm. See Figure 3. The outboard isolation valve i

would have maintained containment integrity.

o On May 24,1996, the LLRT leakage through the inboard isolation check valve (3HCS*V14) on the Hydrogen Recombiner line (2-inch) penetration #114 was 51,900 secm. See Figure 4. The outboard isolation valve would

]

have maintained containment integrity, o On May 29,1996, the LLRT leakage through isolation valves on the Containment Purge supply line was unquantifiable due to excessive leakage. The Containment Purge supply line (42-inch) penetration #86 has three isolation valves, one inboard (3HVU*CTV33A), one outboard (3HVU*CTV32A), and an additional outboard valve on

)

a 30-inch branch line (3HVU*VS). The inboard valve and both outboard valves are tested simultaneously because they cannot be tested independently. See Figure 5. The combined leakage through these three valves was unquantifiable due to excessive leakage that prevented the test pressure from being reached. This leakage alone exceeded the Technical Specification 3.6.1.2.c secondary containment bypass leakage limit of 0.042L. (20,833.9 sccm), and it exceeded the Technical Specification 3.6.1.2.b Type C limit of 0.6L. (294,800.2 sccm). It also resulted in the totalleakage exceeding an analyzed condition that had been previously determined. Accordingly, an immediate notification was made on May 29,1996, pursuant to 10CFR50.72(b)(2)(i) for an event found while shutdown, that if found while operating, would have resulted in an unanalyzed condition that significantly compromises plant safety. No immediate action was required because the plant was in a shutdown condition.

.~

' NRC FORM 3e:A U.S. NUCLEAR REoVLATORY COMMISSloN

{490 UCENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REvlsioN Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 3 of 8 00 012 96 TEXT (If rnore space is required. use additional copies of NRC form 366A) (17) e i

o On June 6,1996, the LLRT leakage through the inboard isolation check valve (3SIL*V6) on the Low Pressure Safety Injection line (6-inch) penetration #93 was unquantifiable due to excessive leakage that prevented the test pressure from being reached. See Figure 6. The outboard isolation valve would have maintained containment integrity.

The attached Table i provides a summary of the leakage rates for these valves which had excess leakage, and it provides a comparison to the Technical Specification limits.

j

11..Cause of Eve.rli i

The cause of the excessive leakage through the inboard isolation check valve 3QSS*V4 on the Containment Quench j

Spray line penetration #100 is attributed to boric acid residue which prevented the disc from fully closing. From the time the line is drained to perform the test, until the leak test is actually completed is approximately two days. It is believed that residual borated water in the bowl of the valve, which is the low point ir the system, evaporated prior to performing the test. This cause was verified by having the line flushed with demineralized water and retested. The as-left results were acceptable and did not Indicate any intemal damage when compared to previous leak results. The internal were examined during the mid 1995 refueling outage, with no disc or seat wear identified.

The cause of the excessive combined LLRT leakage through both: the inboard isolation check valve 3SIL*V13 on the Low Pressure Safety injection line penetration #94; and through the inboard isolation check valve 3SlH*V24 on the High Pressure Safety injection line penetration #98, was attributed to valve 3SlH*V24 either improperly seated or j

foreign material was present in the valve. A flush of the line was performed and the penetration was retested. The as-left results were acceptable. A review of previous leak results for this penetration indicate that the check valve shows a decreasing leakage trend over the last three outages.

I The cause of the excessive leakage through the inboard Isolation check valve 3RSS*V3 on the Containment i

Recirculation Spray line penetration #107 is attributed to boric acid residue which prevented the disc from fully closing.

From the time the line is drained to perform the test, until the leak test is actually completed is approximately two days.

It is believed that residual borated water in the bowl of the valve, which is the low point in the system evaporated prior i

to performing the test. This cause was verified by having the line flushed with demineralized water and retested. The as-left results were acceptable and did not indicate any internal damage when compared to previous leak results.

Further evidence of the cause was obtained by a visualinspection of the B train valve 3RSS*V6, performed this shutdcwn. Boron was present in the bowl of this valve.

The cause of the excessive leakage through the inboard isolation check valve 3HCS*V14 on the Hydrogen Recombiner line penetration #114 was foreign material present in the pipe and disk seating area. The materialis being evaluated and is believed to be hydrocarbon byproduct caused by the higher temperatures produced during the hydrogen Recombiner process.

The cause of the excessive combined leakage through the isolation valves 3HVU*CTV33A,3HVU*CTV32A, and 3HVU*VS on the Containment Purge supply line penetration #86 was a bulge in the outer valve 3HVU*CTV32A seat ring. The bulge is attributed to overtightening during installation performed in the mid 1995 refueling outage.

The cause of the excessive leakage through the inboard Isolation check valve 3SIL*V6 on the Low Pressure Safety injection line penetration #93 is under Investigation and will be reported in a supplement to this LER.

i

- Ill. Analvsis of Event The LLRT leakage through the inboard isolation check valve 3OSS*V4 on the Containment Quench Spray line penetration #100 was 156,000 secm. This leakage, added to the *as-found" total for all other valves as of May 10, NHC FORM 366A 44-90

.U.S. NUCLEAR REoVLAloRY Commission 84 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILrTY NAME {1)

DOCKET NUMBER (2)

LER NUMBER (6)

PACE 13)

YEAR SEQUENTIAL REvlsioN NUMBER NUMBER 4 of 8 Millstone Nuclear Power Station Unit 3 05000423 96 012 00 TEXT Uf more space is required, use additional copies of NRC form 366A) til) 1996, had not exceeded the Technical Specification limit, and by itself is not reportable. However, it is included in this LER because it was a major contributor to the total eventually exceeding the Technical Specification limit five days later. The leakage through the outboard valve 30SS*MV34A was 107.3 sccm, which would have maintained containment integrity.

The combined LLRT leakage through both: the inboard isolation check valve (3SIL*V13) on the Low Pressure Safety injection line penetration #94; and through the inboard isolation check valve (3SlH'V24) on the High Pressure Safety Injection line penetration #98, was 43,900 secm. This event, added to the total for all previous LLRTs exceeded the Technical Specification limit. The leakage through the outboard isolation valves 3SIL*MV88098 and 3SlH*MV8835 was 132 and 20 scem respectively, which would have maintained containment integrity.

The subsequent LLRTs performed on other containment isolation valves revealed additional"as-found" leak rates that contributed to exceeding the Technical Specification limit. The additional isolation valves that were major contributors to the totalleak rate were those associated with: a containment Recirculation spray line, a hydrogen Recombiner line, the containment purge supply line, and a low pressure safety injection line. These additional contributors to the total leak rate are described below.

The LLRT leakage through the inboard isolation check valve (3RSS*V3) on the Containment Recirculation Spray line penetration #107 was 111,100 sccm. The outboard valve 3RSS*MOV20D leakage was 1850 sccm, which would have maintained containment integrity.

The LLRT leakage through the inboard isolation check valve (3HCS*V14) on the Hydrogen Recombiner line penetration #114 was 51,900 sccm. The outboard valve 3HCS*V13 leakage was 20 sccm, which would have maintained containment integrity.

The LLRT leakage through isolation valves on the Containment Purge supply line was unquantifiable due to excessive leakage. The Containment Purge supply line penetration #86 has three isolation valves, one inboard (3HVU*CTV33A),

one outboard (3HVU*CTV32A), and an additional outboard valve on a 30-inch branch line (3HVU*VS). The inboard and both outboard valves are tested simultaneously because they cannot be tested independently. See Figure 5. The combined leakage through these three valves was unquantifiable due to excessive leakage that prevented the test pressure from being reached. This leakage alone exceeded the Technical Specification 3.6.1.2.c secondary containment bypass leakage limit of 0.042L. (20,633.9 secm), and it exceeded the Technical Specification 3.6.1.2.b Type C limit of 0.6L. (294,800.2 sccm). It also resulted in the total leakage exceeding an analyzed condition.

Accordingly, an immediate notification was made on May 29,1996, pursuant to 10CFR50.72(b)(2)(i) for an event found while shutdown, that if found while operating, would have resulted in an unanalyzed condition that significantly compromises plant safety. No immediate action was required because the plant was in a shutdown condition.

The LLRT leakage through the inboard isolation check valve (3SIL*V6) on the Low Pressure Safety injection line penetration #93 was unquantifiable due to excessive leakage that prevented the test pressure from being reached.

The outboard valve 3SIL*MV8809A leakage was 403 sccm, which would have maintained containment integrity.

IV. Corrective Action

The corrective action for the inboard isolation check valve 30SS*V4 on the Containment Quench Spray line penetration #100 was to perform a flush of the line to remove any boron deposits. Inspection results from the mid-1995 refueling outage did not indicate any adverse conditions including seat or disc damage. The as ! eft results from the mid 1995 refueling outage were comparable to the as-left results from this outage. Therefore it is concluded that no rew damage to the seating surface is present. However, the historical review of this valve as compared to the B train valve: 3OSS*V8 shows that 30SS*V4 has consistently higher leakage. This valve will be reworked during the MC FORM 366A (4-95)

.U.S. NUCLEAR REGULATORY COMMISSloN 14 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FActLITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REvlsioN Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBER 5 of 8 96 012 00 TEXT Vimore space is required. use additional copies of NRC Form 366A) (17) next refueling outage, to further improve the seating surfaces. A clean water flush of the QSS system will be evaluated to determine if the flush should be performed on a routine basis for each penetration after each as-found LLRT test. In addition, the number of times the system is drained and refilled is being reduced. Currently the line is drained each quarter to ahow cycling of the outboard isolation valves: 3OSS*MV34A/B. In this condition it is possible for boron to enter into the containment side of the penetration, due to the lower pressure condition inside containment (maintained at a slight vacuum), The draining frequency will be reduced to each cold shutdown, which will reduce the potential for boron to be present in the system.

The corrective action for the inboard isolation check valve 3SlH*V24 on the High Pressure Safety injection line penetration #98, was flushing the line with water. A review of previous leak results indicates an improving trend of overall decreasing leakage. Therefore, it is determined that the valve internals are not adversely affected. Valve 3SlH*V24 will be disassembled for inspection during the next refueling outage if as-found leak rates increase above the current outage as-left results. deviating from the current trend.

The corrective action for the inboard isolation check valve 3RSS*V3 on the Containment Recirculation Spray line penetration #107 was flushing the line with water. The histor/ of this valve does not indicate any decreasing valve leakage. An inspection of the other train valve 3RSS*V6 indicated boron in the valve body. Based on the limited data available, a flush of each penetration is being evaluated. It is believed that performing a closed cooling water flush after each as-found leak test will reduce the level of boron present in the valve.

The corrective action for the inboard isolation check valve 3HCS*V14 on the Hydrogen Recombiner line penetration

1. 114 is being evaluated and will be reported in a supplement to this LER.

The corrective action for the isolation valves 3HVU*CTV32A,3HVU*CTV33A, and 3HVU*V5 on the Containment Purge supply line penetration #86 was to repair the seating surfaces of the 32A/33A valves. The t ring seat of outboard valve 3HVU*CTV32A was fully adjusted after it was identified that a bulge existed in one segment of the t-ring. There are 14 segments each containing three set screws. The maintenance procedure was compared to the vendor recommendations to ensure full compliance with the adjustment process. Due to the design of this penetration it is difficult to identify if both valves or only one valve caused the excessive leakage. Minor adjustments were performed on the inboard valve t-ring seat. A retest produced very low leak rates.

The corrective action for the inboard isolation check valve 3SIL*V6 on the Low Pressure Safety injection line penetration #93 is being evaluated and will be reported in a supplement to this LER.

V.

Additional Information

Similar Events LER 91-004-01 and LER 95-009-02 discuss similar events of containment leakage in excess of Technical Specification limits due to containment isolation valve leakage.

Manufacturer Data Ells Equipment Code: Isolation Valves - ISV NP.C FORM 366A (4 94

- _____ _ _ _U.S. NUCLEAR REGULATORY COMMISSION I4 95)

LICENSEE EVENT REPORT (LER)

TEXT CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENTIAL REVISloN NUMBER NUMBER 6 Of 8 Millstone Nuclear Power Station Unit 3 05000423 012 00 96 TEXT (11 more space is required, use additional copies of NRC Form 366A) til)

TABLE 1 Containment Type C Leakane Data (in secm)

Penetration Valve As-Left As-Found As-Left (Number)

(ID)

(Refuel 5)

{Mid-Cvele 6)

(Mid-Cvele 6)

Quench Spray 30SS*V4 3,740 156,000 8,090 (100)

Safety injection (94) 3SIL*V13 1,176 43,900 1,480 (98) 3SlH*V24 Recirculation Spray 3RSS*V3 1,533 111,100 1,283 (107)

Hydrogen Recombiner 3HCS*V14 1,800 51,900

[?](')

(114)

Low Pressure injection 3SIL*V6 544 Indeterminate

[?](*)

(93)

Other Penetrations Sub-Total 56,278.1 136,059.1 112,143.9 All Penetrations Total 56.976.3 498.959.1 122.996.9 Technical Specification Allowable Limit (0.6LJ 294.800.2 294.800.2 294.800.2 Containment Bypass Leakane Data (in secm)

Penetration Valve As-Left As-Found As-Left (Number)

(ID)

(Refuel 5)

(Mid-Cycle 6)

(Mid-Cycle 6)

Purge Supply 3HVU*CTV33A 664 Indeterminate 20 (86) 3HVU*CTV32A 3HVU'V5 Other Penetrations Sub-Total 3,246 4,253 4,253 All Penetrations Total 3.910.6 4.253 4.273 Technical Specification Allowable Limit (0.042LJ 20.633.9 20.633.9 20.633.9

• Note: The As-Left leak rate will be provided in an LER supplement.. _ _ _ _ _ _ _ _ _ _ __-_-- - _____ -__________ _ _ _ _ _ _

..U.S. NUCLEAR REGULATORY COMMISSION H 961 UCENSEE EVENT REPORT (LER)

TEX T CONTINUATION FACILITY NAME (1)

DOCKET NUMBER (2)

LER NUMBER (6)

PAGE (3)

YEAR SEQUENT (AL REVIStoN Millstone Nuclear Power Stat,on Unit 3 05000423 NuMern Nvuern 7 of 8 i

012 00 96 TEXT (If more space is required, use additional copies of NRC form 3GGA) (17)

Figura 1, Containment Quench Spray Penetration #100 n

vE" T"

.h % *E'E c

5 a.)

R*w

~ 'vI f~ Figure 2, Low Pressure / High Pressure Safety injection Penetration #98 04 LkiC/0 to M kV0M vaF6 vett ' M-M-wcn { wen ( cx e V\\\\ u.,a ~ k "r,He. /j u*'** vm 's Avsm t-H vva.ie e,,,,, ve l-- y y ~~

2, -cg

,,4 m..., LWJD Figure 3, Containment Recirculation Spray Penetration #107 I --- a vnr 10 7 l LtdC/V ) LWC/V \\ H3 $[, D'81 t/

• )j O

NRC FOAM 366A (4-95i

WRC, FORM 3G6A U.S. NUCLEAR REGULATORY COMMISSION 84 9W LICENSEE EVF.NT REPORT (LER) TEXT CONTINUATION FACILITY NAME (1) DOCKET NUMBER (2) LER NUMBER (0) PAGE (3) YEAR SEQUENTI AL REVISION Millstone Nuclear Power Station Unit 3 05000423 NUMBER NUMBE R 8 Of 8 96 012 00 TEXT IIIrnore space is required. Lse additional copies of NRC Form 366A) (17) Figure 4, Hydrogen Recombiner Penetration #114 14 P ,a tuv. rs -y W lj Lhc l n e 50 Figure 5, Containment Purge Supply Penetration #86 .g. EJA 88 CTV)2A-g g,3p m,....r H H h LJ at n y LWC Figure 6, Low Pressure Safety injection Penetration #93 u ste v6 y

k_8, UI n..,

r n.,, w \\ nu

- 3 =a

)l vm un Cv,6am Wy, i L a o'.,o SL e.9 a 28 u wa.... _ _ }}