ML20148T815

Forwards Preliminary Accident Sequence Precursor Analysis of Operational Event at Plant,Units 1 & 2 Which Occurred on 960628,for Review & Comment

ML20148T815
Person / Time
Site:	LaSalle
Issue date:	07/03/1997
From:	Skay D NRC (Affiliation Not Assigned)
To:	Johnson I COMMONWEALTH EDISON CO.
References
NUDOCS 9707090330
Download: ML20148T815 (18)

Text

m fe e

p

[N & UNITED STATES k dh Q'Lb 4

s j NUCLEAR REGULATORY COMMISSION WA6HINGTON, D.C. 30e50001 "o

July 3, 1997 l

Ms. Irene Johnson, Acting Manager Nuclear Regulatory Services Commonwealth Edison Company l Executive Towers West III 1400 Opus Place, Suite 500 l Downers Grove, IL 60515

SUBJECT:

REVIEW OF PRELIMINARY ACCIDENT SEQUENCE PRECURSOR ANALYSIS OF OPERATIONAL EVENT AT LASALLE COUNTY STATION, UNITS 1 AND 2

Dear Ms. Johnson:

! Enclosed for your review and comment is a copy of the preliminary Accident Sequence Precursor (ASP) analysis of an operational event which occurred at

LaSalle County Station, Units 1 and 2, on June 28, 1996 (Enclosure 1), and was reported in Licensee Event Report (LERs) Nos. 373/96-007 and 373/96-008. This analysis was prepared by our contractor at the Oak Ridge National Laboratory (ORNL). The results of this preliminary analysis indicate that this event may be a precursor for 1996. In assessing operational events, an effort was made to make the ASP models as realistic as possible regarding the specific features and response of a given plant to various accident sequence initiators. We realize that licensees may have additional systems and emergency procedures, or other features at their plants that might affect the analysis. Therefore, we are providing you an opportunity to review and l comment on the technical adequacy of the preliminary ASP analysis, including

the depiction of plant equipment and equipment capabilities. Upon receipt and

! evaluation of your comments, we will revise the conditional core damage j i probability calculations, where necessary, to consider the specific '

information you have provided. The object of the review process is to provide  !

as realistic an analysis of the significance of the event as possible.

In order to incorporate Comed's comments, perform any required reanalysis and l prepare the final report of NRC's analysis of this event in a timely manner, Comed is requested to complete its review and to provide any comments within i 30 days of receipt of this letter. NRC has streamlined the ASP program with /

the objective of significantly improving the time after an event in which the /

l l

final precursor analysis of the event is made publicly available. As soon as the final analysis of the event has been completed, NRC will provide (for [/_

Comed's information) the final precursor analysis of the event and the ,

! resolution of Comed's comments. In previous years, licensees have had to wait ' /

l until publication of the Annual Precursor Report (in some cases, up to 23 l months after an event) for the final precursor analysis of an event and the resolution of their comments.

Also enclosed are several items to facilitate your review. Enclosure 2 contains specific guidance for performing the requested review, identifies the 9707090330 970703 PDR ADOCK 05000373 P PDR NRC FlE CENTER COPY li!EIE.ElqEglIl i

l l

l 1

I. Johnson I criteria which NRC will apply to determine whether any credit should be given in the analysis for the use of licensee-identified additional equipment or ,

specific actions in recovering from the event, and describes the specific l information that Comed should provide to support such a claim. Enclosure 3 is a copy of LER Nos. 373/96-007 and 373/96-008, that documented the event.

Please contact me at (301) 415-1322 if you have any questions regarding this request. This request is covered by the existing OMB clearance number (3150-0104) for NRC staff followup review of events documented in LERs. Comed's response to this request is voluntary and does not constitute a licensing requirement.

Sincerely, ORIGINAL SIGNED BY:

Donna M. Skay, Project Manager Project Directorate III-2 Division of Reactor Projects - III/IV Office of Nuclear Reactor Regulation Docket Nos. 50-373, 50-374

Enclosures:

As stated cc w/encls: See next page Distribution:

Docket File PUBLIC l PDIII-2 r/f l J. Roe, JWR E. Adensam, EGAl R. Capra C. Moore i D. Skay i OGC ACRS M. Parker, RIII l S. Mays, AE0D P. O'Reilly, AEOD DOCUMENT NAME: G:\CMNTJR\LASALLE\LA. ASP Ta receive a copy of this document, indicate in the box: "C""Cpljwithoutenclosures"E" " Copy with enclosures "N" " No copy l OFFICE PM:PDIII-2 l6 JA:M-2 l C DaP[ll. F2 le l NAME DSKAY Jef74 > @ 0RE ) R$lF0 f,s DATE 07/ //97 07/( /97 07V5/97 0FFICIAL RECORD COPY

l .

l I. Johnson LaSalle County Station Comonwealth Edison Company Unit Nos. I and 2 cc:

l Phillip P. Steptoe, Esquire Robert Cushing Sidley and Austin Chief, Public Utilities Division l One First National Plaza Illinois Attorney General's Office l Chicago, Illinois 60603 100 West Randolph Street Assistant Attorney General 100 West Randolph Street Michael I. Miller, Esquire

. Nite 12 Sidley and Austin Chicago, Illinois 60601 One First National Plaza Chicago, Illinois 60603 U.S. Nuclear Regulatory Comission Resident inspectors Office LaSalle Station Document Control Desk-Licensing l 2605 N. 21st Road Comonwealth Edison Company  ;

Marseilles, Illinois 61341-9756 1400 Opus Place, Suite 400 l Downers Grove, Illinois 60515 Chairman LaSalle County Board of Supervisors l LaSalle County Courthouse

! 0'.tawa, Illinois 61350 At tor.7ey General  ;

500 South Second Street 5pringfield, Illinois 62701 Chairman Illinois Comerce Comission Leland Building 527 East Capitol Avenue Springfield, Illinois 62706 Illinois Department of Nuclear Safety l Office of Nuclear Facility Safety 1035 Outer Park Drive l Springfield, Illinois 62704 Reqional Administrator

! U.S. NRC, Region III '

i 801 Warrenville Road l Lisle, Illinois 60532-4351 LaSalle Stat on Manager 4 LaSalle County Station '

Rural Route 1 P.O. Box 220 Marseilles, Illinois 61341 4

4 l

.. l i

f LER No. 373/96-007,-008 j i

I LER No. 373/96-007,-008 l, i

Event

Description:

Concrete scalant fouls cooling water systems Date ofEvent: June 28,1996 l Plant: LaSalle 1 and 2 i Event Summary  :

1 i

A foam scalant was inadvertently injected into the service water tunnel for Units 1 and 2 resulting in fouling j of nonessential service water, and potential fouling of essential (safety-related) coohng water systems. The  !

conditional core damage probability (CCDP) estimatad for this event is 2.3 x 104 at LaSalle 2. Because of i the location of the scalant and the location of the emergency service water system (ESW) pump intakes, the  ;

sI=ai&- of this event would be slightly lower for LaSalle I than for LaSalle 2. l

-l Event Description l

The Licensee Event Report (LER) for this event [Ref.1,2] reports that between May 21 and June.21,1996, contractors, as requested by the station's ConsaMared Facility Mamtenance group, began seahng cracks in i the walls and floors of the raw water intake building 'Ihe repair process involved dnihng holes into the side  ;

of each crack along its length and irsecting an expandable foam sealant ("Furmanite") into these holes to seal  !

the crack. While perfornung the repairs, workers started fixing cracks on the top of the service water tunnel. i This tunnel supplies coohng water to both the nanassential and eneantial service water systems Rcranee the l workers believed that they were workmg on a concrete floor laid over soil, they proceeded to drill five holes  !

through the ceiling of the service water tunnel. Consequently,instead ofinjecting scalant into a void under j the building floor, the material was injected into the service water tunnel. In all, personne: injected between >

80 and 120 cubic feet of scalant into the tunnel.  !

On June 19,1996, a high differential pressure developed across the non-essential service water (NESW) eramers for both units. An automatic backwash for the strainers for both units failed and differential pressure i amnes the strainers eve-lad the nonnal backwash setpoint by as much as 8 psid. Reactor power was reduced i to about 77% on both units to reduce loading of the service water system and the strainers were isolated one  !

at a time for repairs AAer the repairs were completed, operators were able to manually backwash each  !

stramer successfully. Initially, it was believed that the stramers had been fouled by " corn cob" material being used in manAlasting the exterior of the raw water intakeutding ,

On June 24,1996, normal surveillance tests were a**=a*~1 on the station's diesel fire water pumps (DFPs).

While DFP OA performed satisfactorily, DFP OB exponenced a high cooling water outlet temperature, I indicative of flow blockage, aAer about 5 min of operation. DFP OB was shut down and both pumps were j declared inoperable. Later on the same day, high differential pressures were experienced again across the

i

f

?- L I  !

Enclosure 1 l- i I, - ,

l

LER No. 373/96-007,-008 l

l non-essential service water stramers, which again failed to successfully automatically backwash. Operators  ;

reduced plant service water heat loads and manually backwashed the strainers On June 25 and June 26, test runs of the five emergency diesel generators and the four residual heat removal l systems were made, and no operability problems were noted. Unit 1 scrammed due to instrument calibration l problems late on June 26. On June 27, a Unit 2 service water stramer was P~8 and was found to be operabic. On June 28, divers performed an mspecten of the service water tunnel and discovered that a significant amount of debris remained present (this was later determmed to be the Furmanite sealant). .

, Essential service water systems taking sucten from the service water tunnel were declared inoperable. Unit l l 1 was scrammed from 1% power, and Unit 2 was reduced from 100% to 5% power and then scrammed j l

Additional Event-Related Information  !

Both units were mamtained in hot shutdown so that decay heat removal systems relying upon ESW would j not be demanded and the service water tunnel was extensively cleaned Between 80 and 120 cubic feet of l l scalant material were removed from the tunaci, including one large piece measuring about 8 ft wide by 15 l l- A long by up to 1 A thick. ]

l The Augmented Inspection Team report and other information provided by personnel at en==aawealth  !

%an [Ref. 3, 6] indicate that on July 4 and July 5,1997, ESW pumps were run for several hours for testing l and all performed adequately, except for the UI A fuel pool emergency makeup (FPEM) pump and the U2 i B FPEM pump, which were out of service. On July 5, the 2A RHRSW strainer was opened for inspection.  !

A large number of sealant pieces of significant size were found and the strainer was described as being filled  !

50-60% with debris. Examination of several tubes with a borescope revealed that all tubes ev=iH were i blocked with scalant.' Vendor information indicated that debris particles greater than 2 in. in size would not  ;

l be removed during strainer backwash because the internal diameter (ID) of the backwash arm was 2 in. Some  ;

l pieces of sealant debris were observed to have at least one dunension greater than the ID of the backwash  ;

arm. Only small amounts of debris were found in the balance of the ESW system. On July 7, both units were >

placedin cold shutdown  !

Subsequent analysis determined that, given the layout of the senice water tunnel and the sealant injection l points, the vulnerability of Unit 1 ESW systems to fouling was low. Potentially vulnerable systems taking  ;

suction from the Unit 2 end of the senice water tunnel include the Unit I and Unit 2 NESW systems,2A and s l 2B RHRSW systems, the 2A diesel generator cooling water system,2A and 2B fuel pool emergency makeup pumps, and 0A and OB diesel fire pumps.  !

Modeling Assumptions l

A rJ"'% risk analysis of the event prepared by the licensee [Ref. 4] indicates that, based on the judgmect of personnel r==a~iiag to the events, the combined probability that either the June 19 or the June l l

L

'Not all tubes could be exammed using the borescope l i t i

k 2

i I

t

l l ,

l

( LER No. 373/96-007,-008 i

i 24 strainer failures could have resulted in a total loss of normal service water and subsequent scram was about  ;

r j 0.5. This assumption appears reasonable, and the event was analyzed as an at-power scram with non-safety-related service water unavailable. Unavailability of the NESW renders b i ='e=t systems unavailable, including the 4.* system (CDS), main feedwater (MFW), power conversion systems (PCS), the control rod drive hydraulic supply sysicen (CRD), and systems required for contamment venting (CVS).

l 1 l No test data could be identified to show that the potentially affected systems were fully operable aAer the I l June 19 event. Some testag was performed around June 25,1996, that indicated that ESW was operable aAer l the second foulmg event. An exhaustive system test was performed on July 4,19%, aAer a cleanup of the

. scalant material. Results from this testing also indicated that ESW was operable.

The 2A RHRSW system strainer was reported to be 50-60% fouled with scalant debris subsequent to the ,

l aforcemenhanart testing This fouling occurred despite the fact that the system had been tested aAer the two nported stramer fouling events and much of the original quantity of debris possibly removed from the tunnel.

- In addition, the OB diesel fire pump was found by surveillance testing on June 24 to be inoperable, due to i sealant foulmg ofits cooling water supply.

It is difficult to assess the likelihood that additional safety-related systems dependent on the senice water tunnel could have been rendered moperable by sealant fouling, but it is clear that the sealant material did present a potential concem In its probabilistic risk assessment of the event, Commonwealth Edison assumed a probability of failure of 0.01 for division 2 core standby coolant supply systems Mt upon ESW due to scalant fouhng. E '-,='==t failure of the ESW trains was incorporated into the ASP model fault trees with events DIVIFOUL and DIV2 FOUL. The primary impact on plant safety systems from the loss of a train of ESW (either by DIVIFOUL or DIV2 FOUL) is to fail the senice water supply to an RHR heat exchanger.

An ir.igt.kr.t failure probability of 0.01 was used for each basic event. Consistent with the common-cause strainer failure data presented in INEL-94/0064 (Ref. 5), a common-ctuse failure probability of 0.1 for the j second train of ESW, given that the first train has failed, was also added to the ASP model (basic event l ESW-CCF).

l l In addition, the B fuel pool emergency makeup pump was assumed to be failed based on additional informaten obtamed about the event [Ref. 6).

i' l

Analysis Results The CCDP estimated for this event is 2.3 x 10 .4 The dommant sequence highlighted on the event tree in Figure 1 involves l . the postulated scram with unavailability of PCS and main feedwater,

. failure of RHR, and

. unavailability of contamment venting.

As stated above, the loss of NESW fails CDS, MFW, PCS, CRD, and CVS.

! 3

~

\

. . . - -- . - = - . . - - - - . _. --_ --

l .

l'.

I  :

l .

LER No. 373/96-007,-008 Definitions and probabilities for basic events are shown in Table 1. The conditional probabilities associated with the highest probability sequences are shoan in Table 2. Table 3 lists the sequence logic associated with I the sequences listed in Table 2. Table 4 describes the system names associated with the dominant sequences. i Minimal cut sets associated with the dommant sequences are shown in Table 5.

Acronyms ASP accident sequence precursor i CCDP conditional core damage probability l CDS condensate storage system CRD control rod drive system CVS containment venting system DFPs dicsci fire water pumps l ESW essential service water HPCS high pressure core spray ID internal diameter l

LER licensee event report

! MFW main feedwater l NESW non-essential service water l PCS power conversion system RCIC reactor core isolation cooling RHR residual heat removal RHRSW residual heat removal system senice water SRVs safety relief valves TRANS transient i

References

l. Licensee Event Report (373/96-008) from Commonwealth Edison to the U.S. Nuclear Regulatory Commission," Foreign Material Injected Into Service Water Tunnel Causes Dual Unit Shutdown Due to Inadequate Work Control," November 25,1996.

2. Licensee Event Report (373/96-007) from C-manwealth Edison to the U.S. Nuclear Regulatory l Commission," Unit 1 Reactor Scram on Main Steam Flow High Trip Isolation Durmg Surveillance,"

l July 24,1996.

3. "NRC Region III Augmented Inspection Team Review of the Potential Loss of the Ultimate Heat Sink Due to Foreign Material in the Safety Related Service Water Intake Tunnel Inspection Report," U.S.

l Nuclear Regulatory Commission, August 2,1996.

4. "PRA Report of the impact of Foam Sealant Injection in the LaSalle County Nuclear Station Senice Water Tunnel," Commonwealth Edison, September 23,1996.

I 4

l

=

i l

l LER No. 373/96-007,-008

5. Marshall and Rasmuson, Common-Cause Failure Data Collection and Analysis System Volume 6 - ,

Common-Cause Failure Parameter Estimates, INEL-94/0064, December 1995. \

6. Teleconference involving personnel from the U.S. Nuclear Regulatory Commission, Oak Ridge National Laboratory, and C====weahh FAenri, April 11,1997 1

1 l

)

l 1

I I

i t

I i

4 l 5 l

I l

l 4

5 i

I

i l

i l

l l

LER No. 373/96-007,-008 l

l r

g ......................................................................I I 1 ----nuu m. n= == =umu , = =. a .. u = =. n= == = === == .

II,I [ 3 ill '

l I j .. _. . . _. _.

hlh I , -- . _ - -- - -- --

l l -. __ . .

I E i l F

g .

llI 8 ,

ow ll '

e lwa i

a  : <

n i

. .s 2

u>

' I i i;l il il 5 h il I ,

^

l l i I ,

' l s L \

i

! Figure 1. Dommant core damage sequence for LER Nos. 373/96-007,-008. (The loss of i NESW fails the following systems: CDS, MFW, PCS, CRD, and CVS.)

l 6 l l

' af LER No. 373/96-007, 408 l

Table 1. Definitions and Probabilities for Selected Basic Events for LER No. 373/96-007,-cos i

! Modified l Event Base Current for this

name Description probability probability Type event IE-1.OOP Loss of Offsite Power Immates s.5 E 006 0.0 E+000 Yes Event IE-SLOCA SmallI. mas of Coolant jamates 1.7 E 006 0.0 E+000 Yes

! Event IE 11 TANS-NESW Tromment(TRANS)Immatug 5.0 E401 5.0 E 001 NEW Yes Event-Lase of-NESW ADS-XHE-XE-ERROR Operator Fails to Depressurize 1.0 E402 1.0 E-002 No Using the Automatic Depressurusten System CDS-TNK-HW-CST Caulensate Storage Tank Fails 1.0 E-005 1.0 E-005 No CSS-XHE-XE-NOREC Operator Fails to Recover 1.0 E+000 1.0 E+000 No h- Spreys .

DIVIFOUL Failure of Division 1 Safety 1.0 E-002 1.0 E-002 NEW Yes Systems due to the imes of ESW DIV2 FOUL Failure of Division 2 Safety 1.0 E 002 1.0 E 002 NEW Yes Symems due to the less of E5W ESW CCF Common Cause Failure of the 1.0 E 001 1.0 E401 NEW Yes ESW HCS-MDP-FC TRAIN High Pressure Core Spray (HPCS) 13 E 002 13 E402 No Train Imvol Failures -

HCS-XHE-XE-NOREC Operator Fails to Recover HPCS 7.0 E-001 7.0 E-001 No PPR-SRV OO.2VLVS One or Two Safety Relief Valves 3.2 E 002 3.2 E-002 No (SRVs)Failto Close RCI-TDP-FC-TRAIN Reactor Core Isolation Coolms 3.8 E 002 3.8 E-002 No (RCIC) Train Components failures RCI-XHE XE-NOREC Operator Fails to Recover RCIC 7.0 E 001 7.0 E 001 No RH-PC-LT Operator Fails to Recover the 2.8 E 002 2.8 E-002 No Residual Heat Removal (RHR)

System and the Power Conversion System (PCS)Over a 12 h Period RHR-MDP CF-MDPS Common Cause Failure of RHR 1.0 E404 1.0 EA04 No Pumps 7

LER No. 373/96-007,-008 ,

i Table 1. Definitions and Probabilities for Selected Basic Events for LER No. 373/96-007,-008 i

Modified  ;

Event Base Current for this name Description probability probability Type event RHR-MDP-FC-TRNA RHR Train A Components Fail 3.8 E 003 3.8 E-003 No l

RHR MDP-FC-1RNB - RHR Train B Components Fail 3.8 E-003 3.8 E403 No l

RHR-MOV OO-BYPSA RHR Heat Exchanger Bypass 3.0 E 003 3.0 E-003 No Valve Fails to Close RHR-MOV OO-BYPSB RHR Heat Exchanger B Bypass 3.0 E 003 3.0 E-003 No Valve Fails to Close SDC-XHE XE-NOREC Operator Fails to Recover RHR I.0 E+000 1.0 E+000 No l

SPC-XHE-XE-NOREC Operator Fails to Recover RHR 10E+000 1.0 Em No SRV One or More of the SRVs Fail to 3.1 E-002 3.2 E 002 No Close i SSW-MDP-FC-FPEMU Failure of the Fuel Pool 3.3 E-003 1.0 E+000 Yes l Emergency Makeup Pump Train 5

l i

i

. 8

f

• s LER No. 373/96-007,-008 h

! Table 2. Sequence Conditional Probabilities for LER 373/96-007, 008 I Conditional core

Event tree Sequence - damage Percent name number probability (CCDP) contribution TRANS-NESW 07 1.9E-005 83.9 TRANS-NESW 62 1.5E-006 6.4

TRANS-NESW 31 1.2E-006 5.1 TRANS-NESW 37 6.5E-007 2.7 i

i Total (all sequences) 2.3E-005 1

1 4

i Table 3. Sequence Logic for Dominant Sequences for LER 373/96-007,-008 Event tree name ,"gr Logic TRANS-NESW 07 /RPS, PCS, /SRV, MFW, /HCS, RHR, CVS TRANS-NESW 62 /RPS, PCS, P2, HCS, ADS TRANS-NESW 31 /RPS, PCS, /SRV, MFW, HCS, RCI, ADS, CRD TRANS-NESW 37 /RPS, PCS, P2, /HCS, CDS,

/LCS, RHR, CVS 9

LER No. 373/96-007,-008 ,

f Table 4. System names for LER 373/06-007.-008 j b3tem name Logie ADS Automatic Depressurization Fails l CDS Failure of the Condensate System j CRD Insufficient CRD Flow to the RCS ,

CVS Containment (Suppression Pool) Venting HCS HPCS Fails to Provide Suflicient Flow to the Reactor Vessel 1  !

LCS Low Pressure Core Spray MFW Main Feedwater System j P2 One or Two SRVs Fail to Close l

l l PCS Power Conversion System Fails RCI RCIC Fails to Provide Suflicient Flow to the i RCS )

4 RHR Residual Heat RemovalFails l RPS Reactor Shutdown Fails SRV None of the SRVs Failto Close l

I i

10 i

E 4

LER No. 373/96-007, 008 i

Table 5. Conditional Cut Sets for Higher Probability Sequences for LER No. 373/96-007,-008  ;

Cut set Percent number Contribution CCDP' Cut sets

• TRANS-NESW Sequence 07 1.9 E-005 m >%  %% ,

J 68.8 1.3 E-005 ETRANS-NESW, /SRV, DIVIFOUL, ESW-CCF, SDC-XHE-XE-NOREC, SPC-XHE-XE NOREC, RH-PC-LT, CSS-XHE-XLNOREC

2 6.8 1.3 E-006 ILTRANS-NESW, /SRV, RHR-MDP-CF-MDPS, SDC-XHLXE-NOREC, SPC-XHLXLNOREC, RH-PC-LT, i CSS-XHLXLNOREC 3 6.8 1.3 E-006 ILTRANS-NESW,/SRV,DIVIFOUL, DIV2 FOUL, SDC-XHE-XE-NOREC, SPC-XHE-XE-NOREC, RH-PC-LT, CSS-XHLXE-NOREC 4 2.6 5.1 E-007 IE-TRANS-NESW, /SRV, RHR-MDP-FC-TRNA, DIV2 FOUL,

/HCS, SDC-XHE-XE-NOREC, EPC-XHE-XLNOREC, RH-PC-LT, CSS-XHE-XE-MOREC 5 2.6 5.1E-007 IE TRANS-NESW, /SRV, DIVIFOUL, RHR-MDP-FC-TRNB,

/HCS, SDC-XHLXLNOREC, SPC-XHE-XE-NOREC, RH PC LT, CSS-XHE-XE-NOREC 6 2.0 4.0 E-007 IE-TRANS-NESW, /SRV, DIVIFOUL, RHR *MOV-OO-BYPSB,

/HCS, SDC-XHLXLNOREC, SPC-XHLXE-NOREC, RH-PC-LT, CSS-XHLXE-NOREC

, 7 2.0 4.0 E-( 07 ILTRANS-NESW, /SRV, RHk-MOV-OO-BYPSA, DIV2 FOUL,

/HCS, SDC-XHE XLNOREC, SPC-XHLXLNOREC, RH-PC-LT, CSS-XHLXLNOREC

^

TRANS-NESW Sequence 62 1.5 E-006 ,

• 6 ,

m 4u 1 97.0 1.4 E-006 IE-TRANS-NESW, PPR-SRV OO-2VLVS, HCS-MDP-FC TRAIN, HCS-XIEXLNOREC, ADS XHE-XLERROR TRANS-NESW Sequence 31 1.2 E-006

• 1% -

%4 1 96.6 1.1 E-006 ETRANS-NESW, /SRV, HCS-MDP-FC-TRAIN, HCS-XFEXLNOREC, RCI-TDP-FC-TRAIN, RCI-XHE-XLNOREC, ADS-XHE-XLERROR 4

11

l

! i 4

? LER No. 373/96-007, 408  ;

4

, i Table 5. Conditional Cut Sets for Higher Probability Sequences for LER No. 373/96-007,-008 1 1 i Cut set Percent i 2

aussber Contribution CCDP* Cut sets *

+

J- 2 1.9 2.3 E-008 ETRANS-NESW,/SRV, CDS-1NK-HW CST,

HCS-XHE-XE-NOREC, ADS-XHLXLERROR

~

TRANS-NESW Sequence 37 6.5 E-007 $[Ml!# E ypWf5f%gM$5MI.gy 4

4 1 68.8 4.4 E-007 IETRANS-NESW, PPR SRV OO 2VLYS, DIV! FOUL, j ESW CCF, SDC-XHLXLNOREC, SPC-XHE-XE-NOREC,

! RH-PC-LT, CSS-XHLXE-NOREC 1

2 6.8 4.4 E-008 ETRANS-NESW, PPR-SRV-OO 2VLVS, RHR-MDP CF-MDPS, SDC-XHLXLNOREC, SPC-XHLXE-NOREC, RH-PC-LT, CSS-XHLXE-NOREC 3 6.8 4.4 E-008 IE TRANS-NESW, PPR-SRV-OO-2VLVS, DIVIFOUL, DIV2 FOUL, SDC-XHLXLNOREC, SPC-XHE-XE-NOREC, RH-PC-LT, CSS-XHE-XLNOREC 4 2.6 1.7 E-008 IE 11 TANS-NESW, PPR-SRV OO-2VLVS, RHR-MDP-FC-TRNA DIV2 FOUL, SDC-XHLXE-NOREC, SPC-XHE-XENOREC, RH-PC-LT, CSS-XHE-XLNOREC 5 2.6 1.7 E-008 ILTRANS-NESW, PPR SRV OO-2VLVS, DIVIFOUL, RHR-MDP-FC-TRNB, SDC-XHE-XE-NOREC, SPC-XHE-XLNOREC, RH-PC-LT, CSS-XHE-XE-NOREC 6 2.0 1.3 E-008 ILTRANS-NESW,PPR SRV OO-2VLVS,DIVIFOUL, RHR-MOV OO-BYPSB, SDC-XHE-XE-NOREC, i SPC-XHE-XE-NOREC, RH-PC-LT, CSS-XHE-XE-NOREC I l

7 2.0 1.3 E-008 ILTRANS-NESW, PPR-SRV OO-2VLVS, RHR-MOV OO-BYPSA DIV2 FOUL, SDC-XHE-XE-NOREC, SPC-XHLXE-NOREC, RH-PC-LT, CSS-XHE-XE-NOREC Total (all sequences) 2.3 E-005 M 'ca ~s@ Mf5 N$ ND ' $@ j

• The conditional pmbebility for each cut set is determined by multiplying the pmbebility of the initiating event by the pmbebilities of the basic events in that sainimal cut est. The probabilities for abe initiating events and abe basic events are given in Table 1.

b laitiating events, such as LTRANS-NESW, are not normally included in the output of the fault tree reduction process This event has been added to aid in understan&g the sequences to potential core damage associated with the event. Unavailability of the NESW nutianos the transant and senders numerous systems that are dependent ce service water unavadable, including the CDS, MfW PCS, CRD, and CVS.

I I

l 12 l 1

.?m y es -m e 3- ** - er ,

GUIDANCE FOR LICENSEE REVIEW OF PRELIMINARY ASP ANALYSIS i

Background

The preliminary precursor analysis of an operational event that occurred at i your plant has been provided for your review. This analysis was performed as  !

a part of the NRC's Accident Sequence Precursor (ASP) Program. The ASP j Program uses probabilistic risk assessment techniques to provide estimates of 1 operating event significance in terms of the potential for core damage. The types of events evaluated include actual initiating events, such as a loss of off-site power (LOOP) or loss-of-coolant accident (LOCA), degradation of plant conditions, and safety equipment failures or unavailabilities that could increase the probability of core damage from postulated accident sequences.

This preliminary analysis was conducted using the information contained in the plant-specific final rafety analysis report (FSAR), individual plant examination (IPE), and the licensee event report (LER) for this event.

l Modeling Techniques j The models used for the analysis of 1995 and 1996 events were developed by the Idaho National Engineering Laboratory (INEL). The models were developed using the Systems Analysis Programs for Hands-on Integrated Reliability Evaluations (SAPHIRE) software. The models are based on linked fault trees. Four types of initiating events are considered: (1) transients, (2) loss-of-coolant accidents (LOCAs), (3) losses of offsite power (LOOPS), and (4) steam generator tube ruptures (PWR only). Fault trees were developed for each top event on the event trees to a supercomponent level of detail. The only support system currently modeled is the electric power system.

The models may be modified to include additional detail for the systems /

components of interest for a particular event. This may include additional equipment or mitigation strategies as outlined in the FSAR or IPE.

Probabilities are modified to reflect the particular circumstances of the event being analyzed.

Guidance for Peer Review Comments regarding the analysis should address:

e Does the " Event Description" section accurately describe the event as it occurred? -

• Does the " Additional Event-Related Information" section provide accurate additional information concerning the configuration of the plant and the operation of and procedures associated with relevant systems?

e Does the "Modeling Assumptions" section accurately describe the modeling done for the event? Is the modeling of the event appropriate for the events that occurred or that had the potential to occur under the event conditions? This also includes assumptions regarding the likelihood of equipment recovery.

Enclosure 2

. e l

Appendix H of Reference 1 provides examples of comments and responses for I previous ASP analyses. l Criteria for Evaluating Comments Modifications. to the event analysis may be made based on the comments that you provide. Specific documentation will be required to consider modifications to i the event analysis. . References should be made to portions of the LER, AIT, or other event documentation concerning the sequence of events. System and component capabilities should be supported by references to the FSAR, IPE, plant procedures, or analyses. Comments related to operator response times and capabilities should reference plant procedures, the FSAR, the IPE, or applicable operator response models. Assumptions used in determining failure probabilities should be clearly stated.

L l Criteria for Evaluating Additional Recovery Meaiures Additional systems, equipment, or specific recovery actions may be considered

'for incorporation into the analysis. However, to assess the viability and effectiveness of the equipment and methods, the appropriate documentation must be included in your response. This includes:

l - normal or emergency operating procedures.*

piping and instrumentation dia l

- electrical one-line diagrams,' grams (P& ids),*

- results of thermal-hydraulic analyses, and l

- operator training (both procedures and simulator),* etc.

! Systems, equipment, or specific recovery actions that were not in place at the time of the event will not be considered. Also, the documentation should address the impact (both positive and negative) of the use of the specific recovery measure on:

- the sequence of events, l

the timing of events, ,

- the probability of operator error in using the' system or I equipment, and

- other systems / processes already modeled in the analysis (including l operator actions). l For example, Plant A (a PWR) experiences a reactor trip, and during the subsequent recovery, it is discovered that one train of the auxiliary i feedwater (AFW) system is unavailable. Absent any further information regrading this event, the ASP Program would analyze it as a reactor trip

' with one train of AFW unavailable. The AFW modeling would be patterned after information gathered either from the plant FSAR or the IPE.

However, if information is received about the use of an additional system (such as a standby steam generator feedwater system) in recovering from this event, the transient would be modeled as a reactor trip with one train of AFW unavailable, but this unavailability would be

! i i

!

• Revision or practices at the time the event occurred. 4

. _. ._ ~ .. . - . . - ..

1 mitigated by the use of the standby feedwater system. The mitigation effect for the standby feedwater system would be credited in the analysis provided that the following material was available:

standby feedwater ,ystem characteristics are documented in the FSAR or accounted for in the IPE,

? - procedures for using the system during recovery existed 7.t the time of the event, the plant operators had been trained in the use of the system prior to the event, a clear diagram of the system is available (either in the FSAR,

IPE, or supplied by the licensee),

previous analyses have indicated that there would be sufficient

time available to implement the procedure successfully under the i circumstances of the event under analysis, the effects of using the standby feedwater system on the operation a and recovery of systems or procedares that are already included in the event modeling. In this case. use of the standby feedwater system may reduce th'e likelihood of recovering failr;d AFW l equipment or initiating feed-and oleed due to time and personnel constraints.

Materials Provided for Review The following materials have been provided in the package to facilitate your review of the preliminary analysis of the operational event, e The specific LER, augmented inspection team (AIT) report, or other pertinent reports.

• A summary of the calculation results. An event tree with the dominant

, sequence (s) highlighted. Four tables in the analysis indicate: (1) a summary of the relevant basic events, including modifications to the i probabilities to reflect the circumstances of the event, (2) tha  ;

i dominant core damage saquences, (3) the system names for the systems i cited in the dominant core damage sequences, and (4) cut sets for the

dominant core damage sequences.

Schedule j Please refer to the transmittal letter for schedules and procedures for submitting your comments.

I References I 1. L. N. Vander. Heuvel et al., Precursors to Potential Severe Core Damage Accidents: 1994, A Status Report, USNRC Report NUREG/CR-4674 (ORNL/NOAC-232) Volumes 21 and 22, Martin Marir.tta Energy Systems, Inc., Oak Ridge National Laboratory and Science Applications International Corp.,

December 1995.

l

• Atillmottu rJttil 1.tlevin ( Aintl1Jtit jJ%Jlle (sof teralstig %tJtie sti Jool %rth 21st H ud

\l trwilles llhl34lM*4~

jrlbl% O~O~hl l

. July 24,1996 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 l l

Licensee Event Report #96-007-00, Docket #050-373 is being submitted to your office in accordance with 10 CFR 50.73(a)(2)(iv). I l

Respectfully,

~

)

\

l

" D.J. Ray Station Maneger LaSalle County Station Enciosure cc: H. J. Miller, NRC Region ill Administrator  !

M. P. Huber, NRC Senior Resident inspector - LaSalle C. H. Mathews, IDNS Resident inspector - LaSalle F. Niziolek, IDNS Senior Reactor Analyst INPO - Records Center DCD - Licensing (Hardcopy: Electronic: )

"!07300374u60724 PDR [fi OE I S ADOCK 05000373 PDR Enclosure 3

(5-72) EXP!RES 05/31/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFORMATION COLLECrlON REQUEST: 50.0 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMAE TO ME INFORMAMN LICENSEE EVENT REPOR1 EER) AND RECORDS M AN AGEMENT BRANCH (MNBB 7734). U.S NUCLEAR REGULATORY COMMIS$10N. W ASHINGTON. DC 20555-0001. AND TO THE PAFERWORK REDUCTION PROJECT (3150-0104). OFFICE OF MANAGEMENT AND BUDGET WASHINGTON.DC 20503.

FACILITY NAME (1): DOCKET NUMBER (2) PAGE (3)

LaSalle County Station Unit Ono 05000373 1 of 5 TITLE (4)

Unit 1 Reactor Scram on Main Steam Flow High Trip Isolation during Surveillance EVENT DATE (5) l LER NUMBER (6) REPORT DATE(7) OTHER FACILITIES INVOLVED (8) i Mom oAv yLAa l yLAa suavLyne uvmoN Momu DAY YLA8 FACILITY NAME NUMBLR DOCKET NUMBER NUMBLR 06 26 96 96 007 00 07 24 96 FACILITY NAME DOCKET NUMBER OQ"y l 1 l THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR O (Check one or more)(11)

M*18 l 100 % l (DTI flei f f

y. 20.2201(b) 20.2203(aH3 x0 1 50.73(ax2Hin) _

73 71(b)

V "- - = 20.2203(aH I) 20.2003(ax3 xn)

" E 50.73(nx2Hav) 73 71(c) o 20.2203(aM2 Ns) 20.2003(ax4) 50 73(a N2Hv) OTHER 20.2203(ax2Xu) 50.36(cx 1) 50.73(aK2Xvu) 20.2203(a M2Xm) 50.36(e x2) 50.73(aX2M vmX A) (Specify m Abstract below 20.2203(a H2Hav) 50 73(ax2x0 50 73(ax2 xvm M B) and m 'lext. NRC Form 366A) h 20.2003(ax2H v) 50 73 aN2 Hun 50.73(aH2H z)

LICENSEE CONTACT FOR THIS I ER (12)

NAME TELEPHONE NUMBER (include Area Code)

William Kirchhoff, Site Engineering (815) 357-6761 Extension 2927 COMPLETE ONE LINE FOR EACH COMPONENT FAILURE DESCRIBED IN THIS REPORT (13) i CAUhL sYsTLM COMPONENT MANUFACTVRLR kLMJRT ABL.L I d .. CAUit $YSTLM COMPONLKT MANVI ACTURfA REPORI ABLA M " D5 QW1 ;3, m mDs MC..W3 SUPPLEMENTAL REPORT EXPECTED (14i EXPECTED MO* DA' YLA" O YEs (If yes. complete EXPECTED SUBMISSION DATE)

B NO SUBMISSION D ATE (l5)

ABSTRACT (Lanut to 1400 spaces.s.c..approximately fafteen smgic space typewntten Ames 86)

At 20:56 hours on June 26, 1996, a full Main Steam Isolation valve (MSIV) isolation was received on Unit I during the Instrument Maintenance Department (IMD) instrument mechanics (IM's) performance of a surveillance for calibrating the Main Steamline High Flow Isolation switches. The MSIV isolation trip resulted in an automatic Reactor Scram of Unit I due to the Reactor Protection System (RPS) trip signal from

~

the MSIV not .ull open trip logic. All Safety systems functioned as expected and the reactor safely shut down.

A Primary Containment Isolation System (PCIS) Group 1 half isolation trip on the A2 channel was in place due to the calibration of the IB21-N010C switch and had not been reset. The IM had just completed actions for prepressurizing the high flow differential pressure switch to near reactor pressure, and was in the process of throttling open the flow switch high side isolation valve when a trip was received from the PCIS Group 1 channel b2 for main steam flow high trip. The A2 instrument and B2 instrument lines are shared. The combination of the A2 channel and B2 Channel trips resulted in the full PCIS Group 1 isolation of the MSIVs.

The root cause of the event was an Instrument Maintenance Department work practice deficiency in the proper technique for prepressurizing instruments. A contributing factor was a procedural weakness. The procedure did not include steps to reset the 4 main steam high flow isolation trip channel prior to returning the instrument to service. This action would have reduced the .nz Mability of receiving a full PCIS Group 1 isolation from a pressure spike induces wnile valving in the instrument.

- - - ~ . . - - _ . . - - - . _ - - ~ - .- - .-. .- - - - _ - _ .

i. .

I i

I

~

NRC FORM 3#

g592) ES. NUCLEAR REGULATORY COMMISSION , MWOYFD BY OMB NO 3150-0104 FXPIRES 05/31/95

. 'ESTik '.TED BURDENPEli HESPONSE TO COMPLY WITH THis INFOT v.'. TION C;1.*ffCDON REQUEST: 50.0 HRS. FORWARD LICENSEE EVENT REPORT (LER) COMdFNTS REGARDING BURDEN ESTIMATE TO THE INFORM ATION TEXT CONTINUATION AND RECORDS MANAGWENT BRANCH (MNBB 7714L ES NMAR

, REGULATORY COMMISSION. WASHINGTON, DC 20555 0001. ANDTO THE PAPERWORK REDUCTION PROJECT (31504104k OFFICE OF M ANAGEMENT AND BUDGET, WASHINGTON. DC 20503.

j FACILITY NAME W DOCKET NUMBE.1 (2) LER NUMBER (6i PAGE (3)

YLAR SEQllLNT&AL REWEION NUM8FR NUMBER

{ LaSalle County Station Unit One 96 007 00 2 of 5 050Q373 (If more space is required, use additional copies of NRC Form 366A)(17) i PLANT AND SYSTEM IDENTIFICATION General Electric - Boiling Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as [XX).

A. COSIDITIOIf PRIOR TO EVERIT I

Unit (s): One Event Date: 06/26/96 Event Time: 20:56 Hours  !

Reactor Mode (s): 1 Mode (s) Name: Run Power Level (s): 100%

D. DEACRIPTION OF EVEBIT At 20:56 hours on June 26 1996, the Instrument Maintenance Department (IMD) was performing Instrument Su::veillance LIS-MS-102,* Unit 1 Main Steam Line (MSL) High Flow MSIV Isolation Calibration." Ten of the sixteen Main Steam Isolstion Valve  !

(MSIV, MS) [SB) high flow differential pressure switches had already been successfully calibrated and returned to service. The calibration of the 1B21-N010C flow switch I had just been' completed and the instrument mechanics were in the process of returning the switch to service. i There are four instrument racks, and each instrument rack contains one Primary l Containment Isolation System (PCIS, PC) (NH] subchannel. Each PCIS subchannel '

consists of four flow switches; one from each main steam line. These subchannels are A1 (A switches), B1 (B switches), A2 (C switches) and B2 (D switches). The switches - are configured in a one-out-of-two twice logic. To trip a PCIS subchannel from the high flow switches, at least one switch in a channel must trip on high flow. To receive a Group 1 isolation, one of the A or C channels must trip along with one of the B or D channels.

1 A Primary Containment Isolation System (PCIS) Group 1 half isolation trip on the A2 channel was in place during the calibration of the 1B21-N010C switch ond had not been reset. An Instrument Mechanic had just completed actions for pref >ressurizing l the hi flow dif ferential pressure switch to near reactor pressure, and was in the '

process of throttling open the flow switch high side isolation valve whea a trip was received from the PCIS Group 1 channel B2 for Main steam flow hi trip. The combination of the A2 channel and B2 Channel trips resulted in the full PCIS Group 1 isolation of the MSIVs. The MSIV insolation trip resulted in an automatic Reactor Scram of Unit 1. All control rods fully inserted and Operations established control of Reactor Vessel water level o.2d pressure using the Motor Driven Feed Pump (MDRFP) and Safety Relief Valves (SRVs). Both loops of suppression pool cooling were started to remove the added heat to the suppression pool from the SRVs and the Reactor Core Isolation System (RCIC) was manually started to assist in pressure control of the reactor vessel.

This event is being reported in accordance with the requirements of 10 CFR 50.73 (a) (2) (iv) due to an automatic actuation of an Engineered Safety Feature (ESF).

3 NRC FORM 346 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0!04 (5 92) EXPIRES 05/31/95 ESTIM ATED BURDEN PER RESPONSE TO COMPLY WITH MIS l lNFORMATION COLLECTION REQUEST: 50.0 HRS. FORWAAD COMMENTS REGARDING BURDEN ESTIMATE TO THE INFORd ATION j LICENSEE EVENT REPORT (LER) AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUGEAR '

111XT CONTINUA 110N REGULATORY COMMISSION, WASHINGTON. DC 20555-000) ANDTO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFF.CE OF M ANAGEMENT AND BUDGET, WASHINGTON. DC 20503.

FACILITY NAME (1) DOCKET NUMBER (2 > LER NUMBER (6 i PAGE (3) ,

YLAR ELQULNTIAL RLvi$10N l wuMarp NUMSER LeSalle County Station Unit One 05000373 96 007 00 3 of 5 (if rnore space is required, use additional copies of NRC Fonn 366A)(17)

c. cAusE OF EVEFF The cause of the Scram was a Group 1 Isolation signal initiated by the spurious trip of Main Steam Line (MSL) High Flow switch IB21-N010D caused by a pressure spike on the instrument sensing line during the return to service of the IB21-N010C switch.

The instrument sensing lines of the 1B21-N010C switch (PCIS Group 1 A2 channel) are shared with the 1B21-N010D switch (PCIS Group 1 B2 channel) and with a Feedwater-Reactor level control system main steam line flow transmitter (1C34-N004C). The pressure spike occurred because the main steam high flow switch did not get properly prepressurized prior to returning the instrument to service.

The PCIS Group 1 Isolation trip of the A2 channel had not been reset and the subsequent B2 Channel trip of the PCIS Group 1 logic resulted in a full Group 1 isolation.

The manifold and vent valves of the main stean hi flow switch, 1B21-N010C, which was being returned to service were subsequently replaced and tested. It was suspected that one of the valves may have been leaking or sticking in a manner to result in an abrupt valve position change. It was determined by testing the manifold and vent valves as they responded to maintaining pressure, that none of the valve seats were leaking. However, the hi side vent valve required a greater amount of rotation than normal before valve seat contact was broken as evidenced by dropping of pressure being held. The high side vent had been the valve used by the IM to prepressurize the flow switch.

During the prepressurization of the flow switch, the IM throttled the high side vent valve. This was done in order to minimize the effect on the final switch pressure from the closure of the vent valve which causes further pressurization of the switch due to the relative incompressibility of the water. Because tne valve was being throttled near the fully closed position, the valve was not open off its seat. As a result, the IM was not actually pressurizing the instrument volume, but only the gage which was attached upstream of the high side vent. This method of prepressurizing instruments was verified to be used by other Instrument Mechanics.

The root cause is a work practice deficiency in the proper technique for prepressurizing instruments.

A contributing factor was a procedural weakness. The procedure did not include steps I to reset the main steam high flow isolation trip channel prior to returnirg the instrument to service. This action would have reduced the probability of receiving a full PCIS Group 1 isolation from a pressure spike induced while valving in the instrument.

l l

l

.D NRC FORM 366

• U.S. NUCLE.AR'REGUIATIMW COMMISSION APPROVED BY OMB NO. 3150-0104 0 42)

EXPIRES 05/3I/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITH THIS l INFORMATIONCOLIICTIONREQUEST: 50.0 HRS. FORWARD i LICENSEE EVENT REPORT (LER) COMMMS REGARDING BURDEN ESTIMATE TO THE INFORMATION l TEXT CONTINUATION AN RECORDS MANAGEMENT BRANCH (MNBB MM NWAR  ;

REGULATOR Y COMMISSION, WASHINGTON. DC 205554)001. AND TO -

THE PAPERWORK REDUCTION PROJECT (31504)l04).0FFICE OF l MANAGEMENT AND BUDGET. WASHINGTON. DC 20503. .

i FACIIJTY NAME (Il DOCKET NUMBER (2) LER NUMBER 46) PAC E Y&AR &LQUENTIAL REViklON Nt!Mgra Nt'MBER LsSalle County Station Unit One 05000373 96 007 00 4 of 5 (If more space is required, use additional copies of NRC Form 366A)(17)

LIS-MS-102, Unit 1 Main Stean Line High Flow MSIV Isolation Calibration, instructs  !

the IM to have the Operator reset the PCIS isolation trip logic after the switch i being tested has been valved back in to service. It is however possible to reset j the isolation logic prior to valving the instrument into service because the flow switch is reset with the manifold equalization valve open. Had the isolation trip been reset prior to instrument being valved back in service, it is possible that only a B2 channel trip would have occurred as a result of the spike from the 1B21-

~

N010D switch. The IB21-N0 LOC switch may not have tripped causing a channel A2 trip, {

because the sensed differential pressure across the instrument with the equalizing valve open would still have been lower than the spike on the adjoining switch.

D. ASSESSMENT OF SAFETY CONSEQUENCES The safety significance of this event was minimal. The Primary Containment Isolation System functioned as designed when the high flow isolation signal was received. The positive reactivity due to pressure increase resulting from the ,

i

' closure of the MSIVs is reduced by the initiation of the reactor scram upon start of {

MSIV closure. A f ailure c,f the scram from MSIVs closing would result in a scram {

from either the Reactor Pressure High scram trip or from the Average Power Range Monitor high flux trip logic. However, the scram and Group 1 isolation do represent i a rignificant challenge to safety related equipment which should be minimized. All PCIS and RPS actions were initiated and completed as designed.

D. CORRECTIVE ACTIONS Instrument Maintenance Department personnel will receive additional training on  !

proper techniques to be used in valving instruments back into service. The training will emphasize the risks involved when instruments are not properly prepressurized and appropriate precautions which must be taken. The Control System Technicians  ;

will be trained by October 1, 1996. The *A* IMs will be trained by '

February 1,1997.

A procedure change to LIP-GM-909,

• Opening Process Instrument Lines and Valve Manipulation,* is being implemented to incorporate the information provided during the training sessions described above. The procedure will be revised by December 31, 1996.

A procedure change is being implemented to the Unit 1 and Unit 2 calibration procedures. LIS-MS-102 (202) to reset the isolation trip prior to valving the l instrument flow switch back in service. Electrically defeating the trip from the instrument being returned to service is also being evaluated as a part of this procedure revision. Completion of this procedure change is planned prior to the performance of the procedure while at power. LIS-MS-102 and LIS-MS-202, will be revised before August 15, 1996.

4

- - . ._. _ , _ _ - _ ~ . - - . . . ,- . - ._

1 Nl 1

NRC FORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 31500104 (5-92) EXPIRES 05/3 t/95

" ES11 MATED BURDEN PER RESPONSE TO COMPLY WITH THIS INFukMATION COLLICTION REQUEST: 500 HRS. FORWARD COMMENTS REGARDING BURDEN ESTIMA1T TO THE INFORM ATION LICENSEE EVENT REPORT (LER) AND RECORDS MANAGEMENT BRANCH (MNBB 7714) U.S. NUCLEAR '

TEXT CONTINUATION REGULA1 DRY COMMIS$10N, WASHINGTON DC 20555-0001. ANDTO THE PAPERWORK REDUCTION PROJECT (31504104). OFFICE OF MANAGEMENT AND BUDGET. WASHINGTON.DC 20503. l LER NUMBER (6i PAGE (3)

FACILITY NAME (1) DOCKET NUMBER (2)

YLAk St.QULNwd. kE VISION I Nt'MBF R NUMBER LaSalle County Station Unit One 05000373 96 007 00 5 of 5  !

(If more space is required, use additional copies of NRC Form 366A)(17)

F. PREVIOUS OCCURREDICES LER 3fDNBER TITLE 373/94-015-00 Unit 1 Primary Containment Isolation and SCRAM Due to Switch 4

Failure In the referenced LER, a PCIS Group 1 isolation of the main steam hi flow trip logic

occurred. At the time of the investigation, the tripping of one of the high flow switches was believed to have spuriously occurred due to a contact resistance i problem on the microswitch of the Static 0 Ring (SOR) switch. Foreign material found on the switch contact caused the flow switch to have erratic calibration settings. This was not believed to be the problem in the recent event because the calibration settings of the B2 channel switches were not erratic or abnormal. It is however possible that the cause of the previous event was due to inadequate prepressurization of the flow switch, and not from the erratic operation of the switch.

O. COMPONENT FAILURE DATA Since no component failure occurred, this section is not applicable.

l i

• n i -

e ..nnn. su a alt h I eless een 4 ..uili.im

. 1.es .sli. t. in asing siatc n

.'s a e I % .e tis .* l .1 N. nl u s, - .n. . n e . t < e e ..- <.-

. i a n << . .I 4

l l

j July 28,1996 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20555 Licensee Event Report #96-008-00, Docket #050 373 is being submitted to your office in accordance with 10 CFR 50.73(a)(2)(i).

\

Respectfully, l l

~

D.J. Ray

. 7g Station Manager LaSalle County Station i

Enclosure cc: H. J. Miller, NRC Region ill Administrator M. P. Huber, NRC Senior Resident inspector - LaSalle C. H. Mathews, IDNS Resident inspector - LaSalle F. Niziolek,IDNS Senior Reactor Analyst INPO - Records Center DCD - Licensing (Hardcopy: Electronic: )

t N

r

/

-960009001S 960728 PDR ADOCK 05000373 g PM OP \ \

Q s .W '

_ _ - _ .. _ - -- - - - _ . ~ - - .. - - - -- --- ~

• g APPROVED HY OMH NO 11Wol44 Nar iORM M6 13. NtWI.LAR RfLl'I.ATOkt COMMis5 ION EXPIRES uvilN9

.4 v2s ESTIM ATED HURDt'N PEk RESPONSE TO LOMPLY WITH Tills INR)RMATION COLL 1CTN)N IGQUEST 46o HMS FORw ARD CwMcNTs REcARDmc HURDm ESTIM ATE TO THE ms >wu u n s 1.lCENSEE EVENT REPORT (1.ER) AND RECURDS MANAGEMENT HR ANCH tMNHH 7744L t 5 NL'Cil AH REGULATORY COMM155 TON W ASHINGTON.DC 204WR511 AND TO THE PAPERWORK REDUCTION PROJECT tilhuln4 t OR K T. 0F M ANAGEMENT AND BL'IMET. W A5HINGTON. DC 20911 j PAGEW DOCKET NUM8F.M 121 l 6 Aril.ITT N AME :In: 05000373 1 of 7 1

LaSalle County Station Unit one TITI ti4:

Foreign Maternal Injected Into ervice Water Tunnel Causes Dual Unit Shutdown Due Inadegaate Work Control

' ' attPo st DATY 476 OTHER FACILIT1F.% INVOI.VED 4RI I Fb Evr DATE on i 1.Em Nt:MsEn as:

wam im u ns uma www u us"* ***T" "" "^" FACILITY NAME DOCKhT NI'MHG

"" "* LaSalle County Station Unit Two 05000374 07 28 96 FACIUTY NAME DOCKET Nt'MH17.R 06 28 96 96 008 00 IChed one or mm'llit f Q"lmasa l 0 302 l THIS REPORT IS SUBMITIT.D PURSUANT TO THE REQUIREMFN!'S OF 10 CFR I:

L t'% BJ .1 7 % 7lebi 20 220 ka H % lll 30 7 h a el 2 N esil 20 2201:451 "

~

40 7 h a M 2 p ee p 7 4 ilis e

' 20 220h as l e 20 2tulhait lii nst ~

~

SO7han2NwI

• iTHl.H Ju 220 hah 2 nit 20 2U0h a u di

%O Se c a l l SO 7 h a n 2 pi vgll 2il 220 ha n 2 N ui

" iSpeoly m Aburau brium 20 220 ban 2 nnis 90 % m21 ~5 50 7 hen 2nwnin A vi7han [ So 7 kaN2xvnin Hi and en Test SRC herrn LMA 20 :20 ka n 2 nsil ,g 20 2(slhan 2nti %0 7 han 2nne ] 507kan2 Mal LICENSEE CONT ACT 7011 Titl5 iER 4121 TE11 PHONE NUMHER tinclude Area t de Denn:s "rtstave. System Engineering (815) 357-6761 Extens2on 10A1 lNAME COMPTETE ONE I.lNE FOR FA CH COMPONENT F Allt'RE DE%CRIBED IN T'HIS RFFORT (13) teAw1 Ai n me s se m e i en.)

sta s t asi t ( At w win.as rusouhENT epu mn w semm nt*1 es Aw Wn ium fu enigi ng to t %#91n EXPT,CTED ^T" ' "*"

NLtPLEMEVTAL RfrORT EXPMTED el48 96 M) NUBMISSION 12 02 l@ %th all ses .. weep 6ctr EXPf1'TE's SUtsMKilON DATTl D ATE (156 l

AH%TN ACT itarrut in tem spact i e , appruaimaiety fificen smgic4paa typewnteen isnes IN At 23:5 hours on June 28, 1996, with Unit 1 at 14 power and Unit 2 at 100% power. the station declared all Core Standby Coolang Systems (CSCS). Emergency Core Cooling Systems (ErC8) and Diesel Generators (DC) anoperable due to fcreign material identified on the

10o: .> l the serv:ce water tunnel. The tunnel as the source for the Essential Service Wafen Syst em. Non-essentaal Servnce Water Systems and the Fire Protection systems. The foseavn ma ut vil had the potentaal to cause a common mode failure of the Essential Servare Water System Although the systems were declared inoperable. they were available. The foreagn maternal was an injectable sealant foam substance which had been used since May, 1996, in the Lake Screen House (LSH) to seal water seepage cracks in a portion of the floor of tne building (the ceiltng of the service water tunnel). After the units were shutdown. sealant material was removed from the tunnel. Systems and components which would have been affected by the foreign material were inspected. cleaned ano tested to verify operability prior to returning the units to service.

The cause of the event was a breakdown in the procedural and work control process. The LSH crack repair had been incorrectly classified as minor facility maintenance and had not been adequately reviewed as a nuclear work request. Weaknesses in the policies and procedures for control of work of minor maintenance actions were identified and corrective acr2ons scheduled.

l 4

U.S. N14' LEAR REGUI.ATORY COMMIMION APPkOVED BY OMB NO IIM0lud l N R C i O E M )*6 j EXPIRE.5 UA/11M d C8 i ESTIMATED BURDEN PER RESM)NSE TO COMPLY WITH THi%

INPORMATION n)LL10' TION REQUEST 90 0 HR5 R)R W AkD COMMENTS REGARDING BURDEN ESTIM ATE TO THE INH *M a flON LICENSEE EVENT REPORT (LER) AND RECORDS MANAGEMENT BR ANCH tMNBB 7714. U S NUrtL AR 111XT CONTINUATION REGULATORY COMMiss UN. WASHINGTON. DC 2nsmunit. AND TO THE PAPERWORK REDUCTION PitOJECT tll WOl(M). OFTlCE OF MANAGEMENT AND BUDGET.WA%HINGTON.DC 2nion PAGE 436 FArttJTY NAME til DOCKFT Nt!MBER (26 EE.R NUMsER 60 VLam suga.wuas as sism ,

wrwou w wou  !

96 008 00 2 of 7 LaSalle County Station Unit One 05000373 ilf annte spxc n required. uw additmaal wpics of NRC Forrn 46AN17)

PLANT AND SYSTEM IDENTIFICATION Gener al Elect ric - Boiling Water Reactor Energy Industry Identification System (EIIJ) codes are identified an the text as IXXJ.

A. ContDITION PRIOR TO EVEarT Unit t sl : 12 Event Date: 06/28/96 Event Time: 23:15 Hours Feactor Modets): 2/1 Models) Name: Startup/Run Pcwer Level (s): 001/1004 D. DESCRIPTIO8f OF EVE 3fT On May 23, 1996, workers began seeltr.g cracks in the walls and floors of the Lake Screen House with an an)ectable sealant. This work was being done to stop ground water inleakage and prepare the building for painting. The work continued through June 21. 1996. The actavaty, evaluated to be reinor maintenance work, was performed nder a t; act son request by the Station's Corn 211 dated Facility Maintenance (CFM) ,

Jrv p us:ng contractors expertenced with this type repair. J The process for performing the crack repair normally eequired drilling holes on each side of a crack along its length and injecting an expanod.Sa sealant into these holes to seal the crack. Cracks of 3 feet to 9 feet long were being repaired.

Normally, this process would not require any drilling through the wall or floor.

However, af a void was found, as indicated by excessive arounts of water, then the practtee was to drill through near the crack and inject to fill the void. The i

sealant used would expand and block further water intrusion.

Whtle doing floor repairs. the workers started fixing cracks on the top or cellang of a service water tunnel which runs the length of the building approximate 1'y 20 feet below lake level (see sketch). This tunnel supplies cooling water to both the non essenttal and essentaaltsafety related) cooling water pumps at the plant. As they repaared these cracks, the large amount of water at pressure indicated to the workers that a large void was present. The workers believed that they were working on a concrete floor laid over soil. They proceeded to drill five holes through the ceal:ng of the service water tunnel and an]ect sealant. Instead of being injected into a vond under the building floor, the material was injected into the tunnel 3

accumulating on the ceiling and floor or dispersing into the cooling water.

' On June 19. 1996 with both unats at approximately full power, high differential pressure occurred on the on-line non essential service water strainers (WS)(KG).

Operators also observed that service water header pressure had decreased below normal. Upon inspection, two of the three strainers were found in automatic backwash but failurer on the backwash valve actuators and/or binding of the strainer basket Jtverters prevented proper flushing of accumulated material. Power reductions

" m e done on both units to approximately 850 MWe to reduce the service water heat loads and asolate each strainer, one at a time, to repair the valve actuators and free the diverter. Following this, the operators were able to manually backwash The initial investigation into what caused the high dif f erent tal pressures identified that

• corn cob' material used for se.ndblasting the exterior of the Lake Screen House was the potential problem. A large amount o: this Ieachstrainersuccessfully.

material was along the outside walls of the building and it was postulated that the material could have gotten into the water and been pulled into the strainers.

Operators w. e stationed to periodically backwash the strainers and a contingency

i

• at ,om is Nen.zAm a> w i. Atom m >w MawsON

^ FF""$"g"y' ,",," *"*

ESTIMATED 8URDEN PER RE5 PUN 5E TO COMPLY %

INR*M ATRIN RMMrTk7N REQUEST M10 HR5 M*w AND GH&ENTs ktG ARDING BURDEN ESUM AR TO THEl 1,1CENSEE EVENT REPORT (1.ER) AND REn)RD5 M AN AGEMENT HM ANCH sMNHH 771  !

TEXT CONT 1NUATION REGULATOR Y(TJMMt55 TON. 4I98WA5HINGTUN.

otm of7K't ut Ir 2054!

THE PAPERWORK REDt'CTION PetOJECT Sit M AN AGEMENT AND tlUDGET. W ASHINGTON. PAMf re2fi il IJN NtIMsER t.)

IM M UT % t 'M 88< 8t e 2 e uvi exTW stsw

>ACIIIT1%AME O veas wwpe u wous 00 3 of 1 96 008 LaSalle County Station Unit one l 05000373 (li rimre T. ice as requued. use additumal wpies of NRC Forrn WA g17) be estabitsbed to trip the units 2f service water heades pressure could not I maintained.

f On June 24. high strainer differential presrure Prior occurred to this, again on the on-liaeOperating had beeni non-essentral service water strainers. (KP). These pumps normal surveallance tests on the OA and 08 Diesel Fire pumps (FP)After the OA pump was tested take their suction from the service water tunnel.was unsuccessful when the pump had to be stopped satisfactoraly, the OB pumpdue test to high cooling water temperature, an indication of a f ave rninutes anto it's rur'..e OD and the OA D2esel Fire Pumps were declared posssble flow blockage. I and 2 were reduced in power to 850 Mwe to reduce

noperable at 0927 hours0.0107 days <br />0.258 hours <br />0.00153 weeks <br />3.527235e-4 months <br /> and Unit coul

ng water demand, i material in the trash basket which collects Based upon the observation of sealantbackwash water and through discussions with the workers performi the strainer rack repaar in the Lake Screen House. at was now concluded that sealant matestal the was the probable cause of the problems with the strainers and fire pumps notbecame clear fro-corn cob mater 2al. When the strainers were again backwashed, it flush water that sealant material was the predominarst the maternal collecteo in theC!nemical analysis of samples collected f rom several locations fore;gn materaal These samples did not reveal any sneluding the OB Diesel Fire Pump confirmed this.

corn cob

• material.

l Along with this evaluation, further inspections of the tunnel and strainers were scheduled. In addition. the Residual Heat Removal (RHR) 180) and Dis:sel results. GeneratorAn Service Water Systems were run to verify operability with satisfactory t

Operab21 sty Evaluation was performed. Information was obtained from the vendor as l

f when injected into the service water tunne!

to the expected behavior of the sealantthe sealant would expand in the tunnel and that the l This information indicated that reruitang mass would float. Based on thas, the evaluation concluded that there was r:o r a sk to the Essential Service Water System. This was determined by the fact that the mat er ial floated. that theand suction poants from the service water tunnel were given the velocity of water in the tunnel. It was reletavely low in t he tunnel .

not probable that floating matertal would be drawn into the pumps suction supply l

lJnes, A compensatory action taken was to bring two diesel fare pumper trucks on l ote to back up the non-essential service water pumps as the source of water for j

fare suppressson. This equipment was manned on a round the clock basis. This action l

I eas taken because the dsesel fire pumps had previously been declared inoperable.

On June 25. anspections in the Service Water Tunnel were started Jsing divers end.

later, robotte inspection equipment. Due to diver safety considerations, the first inspection was restricted to the area around the bottom of the ladder in the Serv 2ce Water Tunnel. The diver inspection was delayed one day because of a unit 1 SCRAM on

' June 26. The SCRAM was not related to the sealant material in the service water I tunnel. Divers wath robotic equipment reentered the Service Water Tunnel on June 27 They noticed sealant material attached at the top of the Service Water Tuntiel but did not fand sealant on the tunnel floor. On June 27 a Unit 2 serv 2ce water strainer was inspected. This strainer is supplied by pumps close to where the sea; ant was injected. There were no signs of plugging and no signs of damage.

Us:ng this snformation together with the successful surveillance tests, t he Operability Evaluat son was completed on June 28, 1996 indicating that Essentra' service Water systems were operable. Later that day divers encountered sealant materaal on the floor of the tunnel. The discovery that the sealant did not float

nvalidated the Operability Evaluation and as a result the Essential Service Water I

=

l I

.1 ,

AryROVED HY OMH NU llWilW gm poRu ~wa t .s %tCfJAR MUt I. ATOM COMMI% ION  !

EXPING uVilM 6 vf.

ESTIMATED tit.,RDEN PER RESPONSE TO COMPLY WITH THi%

INR)RM ATION CULLECTION REQUEST WlHRS R*w AND COMMEMS REGARDING BURDEN ESTIM ATL TO THE INf4

• M ATioN I,ICENSEE EVENT REPORT (LER) AND RECURDS MANAGEMENT BR ANCH tMNisil 7714 l' % Nt Ct.LAN MIXT CONTINUATION REGULATORY COMMIS$10N WASHINGTON.DC 205Wssil oD TO THE PAPERWORK REDUCTION PROJECT t H 401%.OFTIO. OF M ANAGEMENT AND BUDGET. WASHINGTON. DC 20%t 11RNtlMflER W Paid th

> AC8tJTT hAME Ill (M)CM"Y NUMsER (26 i t.Aa savtn a m at s sum 4t'Maf9 WMMS 96 008 00 4 of 7 l L.a Sa l i County Station Unit One 0500.373 i l til mine 9mc n required, uw addite ml copses of NRC Form M6A H17) syr.iems Core Standt. Ing System [VF). Emergency Core Cooling System, and Diesel Generator Cooling Wat. ./ stem were declared inoperable it 2315 hours0.0268 days <br />0.643 hours <br />0.00383 weeks <br />8.808575e-4 months <br /> on June 2R.

1996. 'Jnit I was a ppr x2mately 19 core thermal power in the startup mode at the

! time and Unat 2 was a 10us core thermal power. To comply with Technical l Speesticataon 1.0.3. unit I was manually scransned at 2338 and Unit 2 was reduced it.

i power t o 9 and manually scraPuned at 0528 on June 29, 1996.

The sta*.aon requested a Notice of Enforcement Discretion (NOED) to allow both Units

' o r ema a n in Hot Shutdown until July 9. 1996, to maximize decay heat removal l

capaba lat y and minimize the probability of the material entering the essential Thas service water system while cleanup and removal of the material were performed.

request was granted.

l l Extensave :nspection and testing of plant equipment were performed to identify the pre.ence of sealant material. Approximately eighty cubic feet of the material was removed, mostly from the tunnel. Some material was found in the essential service w.s t e n equipment sneluding large pieces in one of the resadual heat removal serv 2ce l

water strainers tunat 2. Div2ston I). Foll6 wing successful cleanup and testing of l

a?Iected equipment, the unats were returned to service This even* as reportable in accordance with 10 CFR 50.73(a)(2)(il because of entry

.a*o Technical Specifications 3.0.3 C. CAUSE OF EVEnff The r c,o t cause of this event is t hat work affecting plant safety related structures was assigned and performed outside the controls of the Nuclear Work Request Proces8 Thas occurred when the work was approved without identifying a potential impact on l the Seismac Category 1 Service Water Tunnel or the Service Water System from the

[

' sealant tracetion process used in concrete crack repairs and resulted in the work not being reviewed by Engineering. The sealant work being performed was conssdered "Materael Condition

• repaar and ancorrectly assumed to be non-intrusive. The workers were using

• craft capability

• to perform the sealant work, no Work Package was generated for the work. Two Action Requests (ARs) were used as the authorazation for the repairs. Crack repair work was performed on walls as well as l the floor at the Lake Screen 6-ouse but also on the Seismic Category 1 Service Wate:

l Tunne1-D. ASSESSMEhr? OF SAFETY CONSEQUENCES This event resulted in degradar2on of station Non-Essential and Essential Service Water Systems and the Fire Protection Systems. The degradation occurred due to the presence of in3ectable sealant material which was free to move within the Service water Tunnel. Since the material was free to move, in the event that Essential l

Service Water Pumps were required and had started, a loss of Essential Service Wa'.er could have occurred. In this case, the event is Safety Significant, and resulted an 2ncreased risks to the facility. The injected material created the potential for a conrnon mode f ailure of Essential Service Water and Fire Protection Systems.

However, the actual :nnsequences of the event wera minimal because the both units were maintained in a hot shutdown condition which did not require the use of l

essential service water pumps. By not operating these pumps, significant amounts of foreign material were not pulled into these systems. In addition, the residual heat removal and diesel generator service water systems were tested during the event with satisfactory results.

I r

a

.C rx Ntu.EAa arcus.A rony coMMiniON

• erhuvED ay (Pul* NO *i*ai'd NRrtunM w FAPIRESUtilN' a4 O ESTIM ATED ftl klitNPf* N ESM)NSE TO G)MPLY w ITH 1 Hl4 INM*MATM)N COLLE('rlON REQUEST 900lik5 Ft*% AND

'""MfATS *MARDING BURDEN ESTnt ATE TO THE INH

• M AT N 8%

LICENSEE EVENT kEPORT (LER) AND REGmD$ M ANAGWENT UN ANCH eMNbB WL M W11Ak TEXT CONENUATION REGULATORY COMMt5$K)N.WA1HINGTON.DC 3)ittGEll AND fif TH'i PAPERWORK REDUCTION PRDJECT all%0 Hut OFTICE Ol' MAN AGEME*'T AND BUDGET.W ASHtNGTON. DC 3501 PAGF 836 e ar:IJTV % AME a e torswT Nt Marm v2 LEn NUMsElt se i

• eu we exna ee mm w we,e s uste

")

05000M3 96 006 00 5 of

• aAs!'e County Stetton Unit One i!! more 9.m.c a required. unic additumal copses of NRC Forni 3t> A g171 D. CORASCTIVE ACTIO0f8

1. Floor and wall repair work an the Lake Screen House was stopped.

J. The serv;ce water tunnel was cleaned and inspected.

1 Essent tal Service wat er Systems and balance of plant equipment that could have been impacted by the intrusion of sealant were inspected, cleaned and +ested

?o ensure that the syst ems and equipment would function as designed.

4. The structural integrity of the .*eismac-01 ass ! service water tunnel was evalant ed to ensure t hat repaar activities had not degraded the structure.

'. An Engineering Policy was publishei which outlined repair and controls assontated with building structura, repaars and use of sealants as a temporary l

ir pe r manent s epair requa r sng engar sering t'eview and approval .

l t Pr eedures and guidance documents were revised to ensure consistent direction as provided to personnel when assigning. preparing, and supervising work of i this nature (sealingi The fo. lowing documents were revised: LAP-1300-1. Work l Eequest Procedure; LAP-240-6. Temporary Alteration Procedure: and Maintenance Memo fir 4. Use of Furmanste Spec 2facally, all

• sealant

• act2vities are now r equi red to go t hrough t he Nuclear work Request Process. ,

Appropriate Maintenance, sperat2ons. and CFM personnel were coached on the expectataons regarding sealant type work and the new engineering policy.

fl . The Action Request (ARI screening review process has been changed to include multi-discapline involvement.

9. The St ation has prepared and implemented a Consolidated Facili'ies Maintenance ICFM) Responsibilities Administrative Controls document to clearly delineate work scope boundaries and process lamatations of the CFM organization.

l Guadance was provided to clearly outlines supervisory responsibilities for l

' acceptance of work assignments withan the area. This document al.o provided clarafica?lon of the expectation that Maintenance LAPS and Memo's application to the CFM organazation.

10. Clear identification (by signs or other measures) of the operatsonal safety l

rignificance of the Service Water Tunnel and Lake Screen House floor has been l

completed. i i

11. A Corporate investigation team has been established to independently rev2ew thas event and the Station's response to the event including the investigation and evaluation of conservative decision makirg, unexpected conditions and the resolution of problems. Corrective actions will be issued to institutionalize r I

the lessons learned and apply them to all Comed sites. This action will be completed by October 1. 1996. The details of this investigation will be

! provided in a supplemental report. l l

l 12. An EPN assigned coding or equivalent system to accurately and easily determine structures and non-system specific components which may be safety, seismae, or regulutory related will be developed. This system will be part of the 0-1.ist I

m o'

• e e APPROVED HY OMH NO tl% tipu  ;

, %kt'tOkM h6 t% NL OJAR RFUt't.ATORY COMMIMION i E APIRi$ uV 41/vt l i 4**.', '

J ES IURDEN FER RESPONSE TO COMPLY %ITH Itt!% l' INI aON COL 1E7 TON REQUEST 40ft HRN FON% Akie

^ .15 REGARDING BURDEN f3 TIM Al t. TO THE INH )N M A llos 1,1CENSEE EVENT REPORT (LER) aND RECORDS MANAGEMENT HR ANCH IMNBB 7714) L' $ %t Cll.AN i

l TliXT CONTINUATION REGULATORY COMMISSION.W ASHINGTON.DC 20954 issil AND it ,

l THE PAPERWORK REDUCTION PROJECT til%0 Ol(Ms.OITICL Of MANAGEMENT AND HUDGET. WASHINGTON. DC 201411 DOCKIT NUMBER (2 p 11R Nt?MBER M) FA(.F s.ts D' ACILl1T N AME 8 YLAR iMA t.wTML RH unm 6t WSF R ht'MS$2

'aSalle County Station Unit One 05000373 96 008 00 6 of ~

sll enore spac is required. uw additional copics of NRC Form 366A Hl7) j (cr othe: engineering controlled document) and EWCS data base. This action will enable sample and accurate decision making when performing screening of l new work through the EWCS Syrterr. This will be completed by October 1. 1996.

F. PaEVIOUS OCCURREnfCEs i

l LER aftmIBER TITLE i

l None I

G. CoetPontzarr FAILURE DATA Since no component faalure occurred, this section is not applicable.

l l

l t

I l

l l

t l

[ ,

A NE S E .W U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO 31SolW 4492, EXPIRES 05/31/95 l

. . ESTIMATED BURDEN PER RESPONSE TO COMPLY WTTH THis INFORMATlON COLLECTION REQUEST SC0 HR5 FORWARD .

COMMEm REGARDING BURDEN ESTIM ATE TO THE INFOR M A T80% '

LICENSEE EVENT REPORT (LER) AND RERDS MAN AGEMENT BR ANCH (MNBB 7714L U 5 NUCLE Ak TEXT CONTINUATlON REGULATORY COMM11580N WASHINGTON.DC 20$5%0U01 AND TO THE PAPERWORK REDUCTION PROJEC"r a 31SolWt OFTICE OF M AN AGEMENT AND BUDGET. W ASHINGTON, DC 20W LE.R NUMBER tel PAGE I h d '.11JTY N AML (Il DOCKET NUMBER (2) vs.na sentwTw aa viseaa wenseas wehaaEA 96 000 00 7 of 7 LaSa'le

. County Station Unit One 05000373 lif nuare spae as required, use additional copies of NRC Form 366AX17i Lake Screen House Cross-section View l

l I I I J

s e

N d

s 1

M pocasse sesvts 9

• - M o f

h

~ ~rT: ---f m

i 5

[ ' ^

i ri,......, , .

i

,*.ys;,;

-- ~'

yy%

E

• e.

Water Leve.

• , maam costWLavies == ,.

, matae manateetes 3 ' _

g j. ,

ss - ---- E  ;*

N -_W"*,,e

' w l

{Ti[,

js.

• 710 ft- elev samvecs

• festse ow-* , ,

. -py, ','e g*a

/ t-6-

l '. . ; 1 .e..* .

_ tas ea'sa

',*, M g- . ./. e. -

A,<* ,, . .; . m. . .s.. . .A:d t

T 1

.m. , c e p -

l .w

• _I . 4.s I

l water runae: --  ;

• pt,6. ,. , . . . . 'OA /

p,{' .

7 7 :..'& ..,-:-x *:!., *3 IX, f G%M cscs v X - - == >

COOLeme matta SCaffW gypagg gypPt1 M Pt#8 4 omTge tePPtf l

i gm,..a

,, esist evePet tai test s s at L

I i

4 1

i I

l