ML20148Q430

Proposed Tech Specs Reflecting Improvements Made to Control Room Ventilation Air Intake Radiation Monitoring Sys Logic to Reduce Spurious Actuation of Emergency Filtration Mode of Operation & Unnecessary Challenges to ESF

ML20148Q430
Person / Time
Site:	LaSalle
Issue date:	07/01/1997
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20148Q421	List: ML20148Q416 ML20148Q430
References
NUDOCS 9707070088
Download: ML20148Q430 (25)

Text

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TABLE 3.3.7.1-1

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l RADIATION MONITORING INSTRUMENTATION ,

i MININUM CHANNELS APPLICABLE ALARM / TRIP NEASUREMENT INSTRUMENTATION OPERABLE _ CONDITIONS __SETPOINT RANGE ACTION l

a. Main Control Room Atmospheric Control 2/'htch ? 1,2,3,5 and

• 3.5 mR/hr 0.1 to 10,000 mR/hr' 70 l' r System Radiation Monitoring Subsystem PER TPJP SYSTED1 pig {brtW<6[# .

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• When irradiated fuel is being handled in the secondary containment. [

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• • A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance testing without placing the Trip System in the tripped condition, provided at least one other operable channel i

in the same Trip System is monitoring that Trip runction.

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9707070088 970701 m i DR ADOCK 05000373 l PDR l I.A SAI.T.E - IINIT 1 3/4 3-57 Amendment No . 104 l

l TABLE 3.3.7.1-1 (Continued)

RADIATION MONITORING INSTRUMENTATION ACTION l

ACTION 70 -

l a. With one of the required monitors inoperable, place the l inoperable channel in the downscale tripped condition i within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />; restore the inoperable channel to l OPERABLE status within 7 days, or, within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain operation of the control room emergency filtration system in the pressurization mode of operation.

b. With both of the required monitors inoperable. initiate and maintain operation of the control room emergency filtration system in the pressurization mode of operation within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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LA SALLE - UNIT 1 3/4 3-58 Amendment No. 18 o .. . . . . . . . . . . . . _ . . . _ _ . . _ . _ .. . . .

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l ATTACHMENT B PROPOSED CHANGES TO THE LICENSE / TECHNICAL SPECIFICATIONS Insert A l ACTION 70 -

a. With the number of OPERABLE channels per trip system one less than the minimum required, place the inoperable channel in the '

tripped condition within one hour.

b. With both channels in a trip system inoperable, declare the trip system inoperable. Restore the inoperable trip system to OPERABLE status within 7 days, or, within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain operation of the control room emergency filtration system in the pressurization mode of operation.

c. Otherwise, initiate and maintain operation of the control room emergency filtration system in tne pressurization mode of operration within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. r l

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9NSTRUMENTATION

. BASEE

3/4.3.5 REAC?OR CORE ISOLATION COOLING EYETEM ACTUATION INETRUMrnTATION 1

The reactor core isolation cooling system actuation instrumentation is l provided to initiate actions to assure adequate core cooling in the event of 4 reactor isolation from its primary heat sink and the loss of feedwater flow to

! the reactor vessel without providing actuation of any of the emergency core a cooling equipment.

specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-2-A, ' Addendum To q

Bases for Changes to surveillance Test Intervals and Allowed Out-of-service Times for Selected Instrumentation Technical Specifications ('d.;R RCIC l Instrumentation)*, December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains RCIC initiation capability.

If4.3.6 CONTROL ROD WTTHDRAWAL BLocr TNETRUMENTATTON The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Control Rod Program Controls. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

Specified surveillance intervals and surveillance and. maintenance outage times have been determined in accordance with NEDC-30851P-A, Supplement 1,

• Technical Specification 17tprovement Analysis for BWR Control Rod Block Instrumentation *, October 1988, and GENE-770-06-1-A,

• Bases for Changes to surveillance Test Intervals and Allowed Out-of-service Times for Selected Instrumentation Technical Specifications *, December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains Control Rod Block capability.

3/4.3.7 MONTTORTNC TNETRUMENTATION 3/4.3.7.1 RADIATION MONITORING TNETRUMENTATION The OPERABILITY of the radiation monitoring instrumentation ensures that:

(1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or automatic action is initiated when the radiation level trip setpoint is exceeded.

Specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-1-A, " Bases for Changes to surveillance Test Intervals and Allowed Out-of-Service Times for i Selected Instrumentation Technical Specifications *, December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains initiation capability.

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1.A SALLE - UNIT 1 B 3/4 3-4 Amendment No. 104

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ATTACHMENT B PROPOSED CHANGET TO THE 4

LICENSE / TECHNICAL SPECIFICATIONS i

INSERT B The Control Room and Auxiliary Electric Equipment Room (AEER) Emergency

Filtration System (CREFS) consists of two trains. Each train has one outside air

intake. The Main Control Room Atmospheric Control System (MCRACS) l Radiation Monitoring System consists of two trains, one for each train of I

CREFS. Each MCRACS train contains four radiation monitors arranged in two

, trip systems. Each trip system contains two radiation monitors. Both radiation l monitors in each trip system are required to be OPERABLE for that trip system

to be OPERABLE.

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TAR.LE 3 3 7 1-1 BADIATION MONITORING INSTRUMENTATION MINIMUM CHANNELS APPLICABLE ALARM / TRIP INSTRUMENTATION OPERABLE MEASUREMENT CONDITIONS SETPOINT RANCE ACTION

a. Main control Room 2 /b;;- E; - 1,2,3,5 and

• Atmospheric Control 3.5 mR/hr 0.1 to 10,000'mR/hr 70 Systea Radiation Monitoring Subsystem PER TR.T 5%Tt TEAwJhNuk9 ^

TABLE MpTATIONS When irradiated fuel is being handled in the secondary containment.

A channel may be placed in an inoperable status for up to 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> for required surveillance testing without placing in the same Tripthe Trip System System is monitoring in the that tripped Tripcondition, Function,provided at least one other operable channel i

ACTION STATEMENT ACT N 70 -

a. W1 one of the requir ipped condition wit n 1monitors hours restore inoperabl , place the inoper e channel in e downscale ithin 7 days, or, thin the next E ho s, einitiate inoperable change to OPERABLE atus and ma htain operatio of the contro

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room emergency fit atton system in t With both of t requiredsystem monitors pressurisation so noperable, initiat and maintain o ration of t contro of operation.

room emergen 11tration n the pressurizati mode of opera on within I ifour, g t

6 SEE I N S ERT A i

i LA SALLE - UNIT 2 3/4 3-58 Amendment No. 90

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j ATTACHMENT B

PROPOSED CHANGES TO THE LICENSE / TECHNICAL SPECIFICATIONS J

4 insert A l ACTION 70 -

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a. With the number of OPERABLE channels per trip system one less

than the minimum required, place the inoperable channelin the j tripped condition within one hour.

l b. With both channels in a trip system inoperable, declare the trip j system inoperable. Restore the inoperable trip system to

, OPERABLE status within 7 days, or, within the next 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />, initiate and maintain operation of the control room emergency filtration

system in the pressurization mode of operation. t s

h j c. Otherwise, initiate and maintain operation of the control room j emergency filtration system in the pressurization mode of operation  :

within 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />.

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.. INSTRUMENTATION 1-

• nists 3rd.3.5 arAcTOR corr TsOLATION cOOLTNO EYETEM ACTUATION TNETRUMENTATION lt The reactor core isolation cooling system actuation instrumentation is {

provided to initiate actions to assure adequate core cooling in the event of '

reactor isolation from its primary heat sink ar.d the loss of feedwater flow to l j

the reactor vessel without providing actuation of any of the emergency core cooling equipment. '

I specified surveillance intervals and surveillance and maintenance outage times have been determined in accordance with GENE-770-06-2-A, ' Addendum to  !

4 Bases for Changes to surveillance Test Intervals and Allowed out-of-service Times for Selected Instrumentation Technical Specifications (BWR RCIC ,

Instrumentation)*, December 1992. When a channel is placed in an inoperable  !

l status solely for performance of required surveillances, entry into LCO and  !

required ACTIONS may be delayed, provided the associated function maintains i RCIC initiation capability.

3/d.3.6 CONTROL ROD WITHDRAMAL ELocr TMETRUMPNTATTON

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i The control rod block functions are provided consistent with the requirements of the specifications in Section 3/4.1.4, Control Rod Program '

Controls. The trip logic is arranged so that a trip in any one of the inputs will result in a control rod block.

1

specified surveillance intervals and surveillance and maintenance outage  ;

4 times have been determined in accordance with NEDC-30851P-A, Supplement 1.  !

j

• Technical Specification Improventrc Analysis fer BWR Control Rod Block

• . Instrumentation *, October 1988 Fa GENE-770-06-i-A, " Bases for Changes to surveillance Test Intervals and A) . owed out-of-Service Times for S' elected Instrumentation Technical specifications",' December 1992. When a channel is  ;

placed in an inoperable status solely for performance of required  !

surveillances, entry into LCO and required ACTIONS may be delayed, provided the associated function maintains control Rod Block capability. ]

}/4.3.7 MONTTORTNO TNCTRUMPNTATTON 3 /d 3. 7.1 RADTATION MONTTORTNC TNRTRUMENTATION 4

7 The OPERABILITY of the radiation monitoring instrumentation ensures

' thatt (1) the radiation levels are continually measured in the areas served by the individual channels, and (2) the alarm or aut action is initiated 3

hen the radiation level trip setpoint is exceed ecified surveillance ntervals and surveillance and maintenance outage t es have been determined i

in accordance with GENE-770-06-1-A,

• Bases for Changes to surveillance Test Intervals and Allowed out-of-service Times f or selected Instrumentation Technical specifications", December 1992. When a channel is placed in an inoperable status solely for performance of required surveillances, entry into  !

3 LCO and required ACTIONS may be delayed, provided the associated function 4

maintains initiation capability. l '

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LA SALLE - UNIT 2 B 3/4 3-4 Amendment No. 90 l

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I ATTACHMENT B

! PROPOSED CHANGES TO THE l LICENSE / TECHNICAL SPECIFICATIONS 1

INSERT B

The Control Room and Auxiliary Electric Equipment Room (AEER) Emergency i Filtration System (CREFS) consists of two trains. Each train has one outside air j intake. The Main Control Room Atmospheric Control System (MCRACS) l Radiation Monitoring System coaststs cf two trains, one for each train of CREFS. Each MCRACS train contains four radiation monitors arranged in two trip systems. Each trip system contains two radiation monitors. Both radiation

monitors in each trip system are required to be OPERABLE for that trip system

! to be OPERABLE.

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ATTACHMENT C l LASALLE COUNTY STATION

' EVALUATION OF SIGNIFICANT HAZARD CONSIDERATIONS FOR PROPOSED CHANGES TO t

^

FACILITY OPERATING LICENSES

, NPF-11 AND NPF-18 1

l APPENDIX A TECHNICAL SPECIFICATIONS l

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MAIN CONTROL ROOM ATMOSPHERIC CONTROL SYSTEM RADIATION MONITORING INSTRUMENTATION TECHNICAL SPECIFICATION 1

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! ATTACHMENT C i

SIGNIFICANT HAZARDS CONSIDERATION i j Summary of the Proposed Technical Specification Chanaes: i This proposed license amendment will change Technical Specification (TS)  !

l Table 3.3.7.1-1, Radiation Monitoring Instrumentation minimum number of I channels required to be operable and Action 70. The existing wording of Action

, 70 is confusing and is not consistent with other sections of the TS. A brief i

! discussion of the initial design, the existing design and the proposed design '

l change associated with this TS change is noted below.

There are a total of four monitors (channels) assigned for each radiation monitoring system train. Two monitors are assigned to a trip system.

Therefore, there are two trip systems for each radiation monitoring system train.

In the original installed design, an actuation of either one of tia two monitors in a trip system was sufficient to initiate the isolation actuation. Tnis design

. resulted in frequent and unnecessary actuation due to spurious operations of a single monitor. This resulted in unnecessary challenges to an ESF system.

A design change was made to the Radiation Monitoring System in 1993 to eliminate these challenges. In this design, a two monitors were required for the ESF ac'.uation. This design change electrically connected both trip systems in an attempt to maximize the number of possible two-out-of four combinations.  !

However, the common electrical connection between the two trip systems violated the trip system redundancy requirements of the design basis.

The design associated with this TS change requires the two monitors of a trip system to initiate the Control Room Emergency Filtration System (CREFS).

However, it removes the electrical connection between the redundant trip systems. In addition, the proposed Technical Specification Changes clarify the existing wording relative to the words " monitors", " channels" , " trip systems"  !

and " trains" in order to clearly define the system logic and the specific actions required to operate the system.

Comed has evaluated the proposed Technical Specification Amendment and determined that it does not represent a significant hazards consideration.

Based on the criteria for defining a significant hazards consideration established in 10 CFR 50.92, operation of LaSalle County Station Units 1 and 2 in accordance with the proposed amendment will not:

C-2

ATTACHMENT C SIGNIFICANT HAZARDS CONSIDERATION

1) Involve a significant increase in the probability or consequences of an accident previously evaluated because:

The proposed Technical Specification (TS) change clearly defines the system logic and the specific actions requireci for system operability. It

! will not change the probability of occurrence of any accidents, because the affected radiation monitoring instrumentation is not an accident i.,itior. UFSAR Section 15.9.3.4 analyzed the effects of the loss of ventilation from the Main Control Room in the event of a Station Black Out (SBO). The scope of work for the design change associated with this TS change does not affect this analysis or any of its assumptions. The consequences of an accident will not increase, because the trip system redundancy is being restored to meet design basis requirements. The proposed design change will eliminate the potential of exposing main i

control room personnel to radiation doses that exceed the limits specified in General Design Criteria (GDC) 19. The design change associated with this TS change will comply with the redundancy due to two trip systems, either of which will actuate the control room emergency makeup train as

! required and the potential for spurious actuation's will be reduced due to l the logic change to require two channels of one trip system to cause l actuation. The overall control logic for the remaining portions of the CREFS is not changed by the design change.

The changes proposed to the actions are intended to clarify system logic l wording. The actions assure that automatic trip capability is maintained and if not, then the Control Room Emergency Filtration System is placed in the pressurization mode as in the current TS. This is consistent with the current TS.

Based upon tne above, the proposed amendment will not increase the probability or consequences of any accident previously evaluated.

2) Create the possibility of a new or different kind of accident from any accident previously evaluated because:

l The elimination of the electrical connection between the redundant trip l systems in a given CREFS subsystem will restore trip system independence and eliminate the potential of a single failure disabling the radiation monitoring instrumentation trip function. Specifically, a single  ;

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ATTACHMENT C l SIGNIFICANT HAZARDS CONSIDERATION i

l l failure, resulting from a blown fuse caused by a fault in the affected

existing circuit, could remove the control power to the isolation logic relays in both trip systems. These relays require power in order to

actuate and perform their safety function. A loss of control power to both

! trip systems due to the fault could result in exposing main control room

j. personnel to radiation doses that exceed GDC 19 limits.

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in addition, the changes to Action Statement 70 of the specification j assure that trip capability is maintained.

h Based upon the above, the proposed change will not create the possibility of a new or different kind of accident or transient previously evaluated.

3) Involve a significant reduction in the margin of safety because:

! The proposed TS change will not prevent the isolation logic relays from i performing their function or cause false trips. The alarm / trip setpoints for j the affected monitors (including their measurement ranges) remain i unchanged. The changes proposed to the actions are intended to clarify i system logic wording. The actions assure that automatic trip capability is  !

l maintained and if not, then the Control Room Emergency Filtration j System is placed in the pressurization mode as in the current TS. This is ]

consistent with the current TS.

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j Based on the above, the proposed TS change does not involve a significant reduction in the margin of safety.

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ATTACHMENT C

! SIGNIFICANT HAZARDS CONSIDERATION i

1 Guidance has been provided in " Final Procedures and Standards No Significant l Hazards Considerations, " Final Rule,51 FR 7744, for the application of standards to license change requests for determination of the existence of

significant hazards considerations. This document provides examples of

amendments which are and are not considered likely to involve significant j hazards considerations. These proposed amendments most closely fit the  !

example of a change which either result in some increase to the probability or l consequences of previously analyzed accident or may in some way reduce a j safety margin, but where the results of the change are clearly within the j acceptable criteria with respect to the system or component specified in '

i Standard Review Plan.

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The proposed amendment does not involve a significant relaxation of the f criteria used to established safety limits, a significant relaxation of the bases for

! the limiting safety system settings or a significant relaxation of the bases for j- the limiting conditions for operations. Therefore, based on the guidance

provided in the Federal Register and criteria established in 10 CFR 50.92(c), the

! proposed change does not constitute a significant hazards consideration. '

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i i l ATTACHMENT D

LASALLE COUNTY STATION 1

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ENVIRONMENTAL ASSESSMENT STATEMENT FOR PROPOSED CHANGES TO l a

FACILITY OPERATING LICENSES l NPF-11 AND NPF-18 i

APPENDIX A TECHNICAL SPECIFICATIONS I MAIN CONTROL ROOM ATMOSPHERIC CONTROL SYSTEM RADIATION j' MONITORING INSTRUMENTATION TECHNICAL SPECIFICATION i

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, ATTACHMENT D l ENVIRONMENTAL ASSESSMENT STATEMENT

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i Commonwealth Edison has evaluated the proposed amendment against the 1 criteria for the identification of licensing and regulatory actions requiring l l environmental assessment in accordance with 10 CFR 51.21. It has been

] determined that the proposed changes meet the criteria for a categorical

exclusion as provided under 10 CFR 51.22(c)(9). This conclusion has been i determined because the proposed changes do not pose a significant hazards

} consideration or do not involve a significant increase in the amounts, and no 1 significant changes in the types, of any effluents that may be released Offsite.

This request does not involve a significant increase in the individual or cumulative occupational radiation exposure. Therefore, the Environmental Assessment Statement is not applicable for these changes.

._ -. -_ _. -. _ _ _ _ _ _-._m - _ . _ _ . . _ . _ _ _ _ ____ _ m__. _ _ .

NRC FORM 366 UA NUCLEAR REGULATORY COMMISSION

($-92) , APPROVED BY OMB WO. 3150-0104 EXPIRES 05/3I/95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITil Tills INFORMATION COLLECTION REQUEST: 50.0 !!RS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REG ARDING BURDEN ESTIMATE TO Tile INFORM ATION AND RECORDS MANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR REGULATORY COMMISSION, WASillNGTON, DC 20555 000). AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OFFICE OF MANAGEMENT AND BUDGET. WASIIINGTON, DC 20503 FACILITY NAME(1):

DOCKET NUMBER (2) PAGE (3)

LaSalle County Station Unit one 05000373 1 of 9 TITLE (4)

Inadequate Review of Modification of Main Control Room Atmospheric Control System Radiation Monitoring Logic Results in an Unreviewed Safety Question EVENT DATE (5) LER NUMHER (6) REPORT DATE(7) mom OTilER FACILITIES INVOtXED (N) oAv vuk vuR segy agsjoN Mom uAv van FACIUTY NAME DOCKET NUMBER LaSalle County Station 05000374 Unit Two 12 17 96 96 021 01 02 14 97 FACRJTY NAME DOCKET NUMBER oglpf 4 rona THIS REPORT IS SUBMITTED PURSUANT TO THE REQUIREMENTS OF 10 CFR l: (Check one 000 t.rvri no)

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LICENSEE CONTACT FOR Tills LER (12)

NAME TELEPl!ONE NUMBER (include Area Code)

M. A. Whelan, Engineer (815) 357-6761 Extension 3280 CAUbE ErmM COMPI ETE ONE I INE FOR EACil COMPONENT FAlt.UHF. DESCRIHED IN TilIN REPORT (13)

COMPONENT MANUI ACIURiR REPORIABLE < CAUhE Sv5 TEM s COMPONEN1 MANUI ACTURER REPOR TABLE TO NPRDS < >j;; ,y>] TO NPRDS

> < jf

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Sl*PPLEMENTAL REPORT EXPECTED (14) F G L(, j g,N MOMnt DAv vEAR O YES (ifvez. complete EXPECTED SUBMISSION DATE)

@ NO SUBMISSION DATE (15)

ABSTRACT (Lumi to 1400 spaces, i c approximately filleen smgle-space :)pewn ten imes 16) i On December 17, 1996, an independent review of selected Technical Specification (TS) Clarifications, identified an apparent inconsistency between the TS l Clarification related to the Main Control Room Atmospheric Control System (MCRACS) Radiation Monitoring Svstem and TS 3.4.7.1. On January 13, 1997, the subsequent investigation determined that;

1. There was a functional inconsistency between the Design Basis as described in the FSAR text and the FSAR Logic Diagram for the MCRACS. The original installed design matched the logic diagram.

2.

The modified design installed in 1993 was not consistent with either Design Basis as described in the FSAR text or as shown on the FSAR Logic Diagram. )

l 3. SER Section 9.4.1 indicates that no single failure within the control circuit for the isolation dampers will result in a failed open Control Room Ventilation System. The modified design installed in 1993 does not meet this Single Failure Criteria and consequently introduced an Unreviewed i Safety Question. A 4 hour4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> Emergency Notification System (ENS) phone i

notification was made due to the plant being outside the design bases and i

in an unanalyzed condition.

i Subsequent engineering review has determined that a postulated single failure in the modified circuitry combined with a Design Basis Accident could have resulted

{ in a failure of the Control Room Ventilation System Isolation dampers to isolate resulting in a radiation exposure to Control Room Personnel in excess of 10 CFR 50_ Appendix A General Design Criteria 19 limits.

l

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l NRC FORM 366 U.S. NUCLEAR RFEULATORY COMMISSION APPROVED BY OMB NO. 3150-0104

,(M2) EXPlRES 05/31'95 ESTIMATED UURDEN PER RESPONSE TO COMPLY WITil THIS INFORMATION COLLECTION REQUEST: 50.0 IIRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIM ATE TO Tl!E !NFORM ATION TEXT CONTINUATION AND RECORDS MAN AGEMENT BRANCH (MNBB 7714) U.S. NUCLEAR REGULATORY COMMISSION, WASHINGTON. DC 20$$$-0001. ANDTO THE PAPERWORK REDUCTION PROJECT (3150-0104). OFFICE OF 1 MANAGEMENT ANDBUDGET WASHINGTON,DC20$03.

{

FACILITY NAME 0) DOCKET Nt!MHER (2) IIM NUM RER (6) PAGE (3)

YEAR EtQULNTIAL ILEVIMON NUMRFR NUMHTR j LaSalle County Station Unit One 05000373 96 021 01 2 of 9 l (If more space is required, use additional copics of NRC Form 366A)(17) l ]

( PLANT AND SYSTEM IDENTIFICATION  !

General Electric - Boiling Water Reactor Energy Industry Identification System (EIIS) codes are identified in the text as [XX]. ,

A. CONDITION PRIOR TO EVENT i

l Unit (s): 1/2 Event Date: 12/17/96 Event Time: 1350 Hours Reactor Mode (s): 4/N Mode (s) Name: Cold Shutdown / Power Level (s): Ob/06 Defueled B. DESCRIPTION OF EVENT On July 13, 1993, and July 26, 1993, modification of Unit 1 and Unit 2 Main '

Control Room Atmosphere Control System (MCRACS) Radiation Monitoring System L (PR/VC) [IL) logic was made to prevent spurious Engineered Safety Features (ESP) ,

l actuations. Prior to installation of the modifications, the logic for initiation i of the Emergency Makeup Mode due to High Radiation levels in the ventilation i intakes consisted of four radiation monitors per intake divided into two channels  !

l to provide a one out of two taken once trip logic. See Attachment A. Due to this trip logic, a spurious trip of any one of the radiation monitors caused an  ;

ESF actuation. To eliminate spurious ESF actuations the modifications changed l f the trip logic for each Unit to require actuation of two monitors per channel to  ;

initiate the ESF actuation. See Attachment B.

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In anticipation of the above modification, on May 13, 1993, Technical  !

Specification / License Clarification 03-93, Revision 0, was developed to provide '

guidance to plant reactor operators for compliance with Technical Specification 3.3.7.1 and Table 3.3.7.1-1. The clarification provided a definition of which radiation monitors constitute a trip channel and provided actions for the operator to take in the event a monitor became inoperable.

On August 30, 1996, all Technical Specifications / License Clarifications were reviewed by an Independent Review Group as part of an investigation under LER 374/96-010-00, Inadequate Standards for Technical Specification Clarifications resulted in violations of Technical Specifications and Design Basis. A total of 43 Clarifications were reviewed. Of the 43, seventeen could not be confirmed through engineering judgment that the interpretations met Design -

and Licensing Basis. Sixteen of the 17 were promptly deleted. One of the 17 was appropriately revised and approved August 30, 1996. Technical Specification Clarification 03-93 was reviewed and the determination made that the Clarification was acceptable as written.

On December 17, 1996, an independent review of Technical Specification / License {

Clarification 03-93, Revision 0, determined that there was an apparent

, inconsistency between the Technical Specification Clarification and Technical Specification 3.4.7.1.

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NRC PORM 366 U.S. NUCLEAR REGULATORY COMMisslON  !

APPROVED BY OMH NO. 3150-0104 (542) EXPIRES 05/31!95 i ESTIMATED BURDEN PER RESPONSE TO COMPLY WITil Tills INFORMATION COLLECTION REQUEST: 50.0 IIRS. FORWARD  !

LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIM ATE TO Ti!E INFORMATION i AND RECORDS MANAGEMENT BRANCil(MNBB 7714), U.S. NUCLEAR TEXT CONTINUATION REGULATORY COMMISSION WASilINGTON DC 20555-0001. AND TO ,

• lilE PAPERWORK REDUCTION PROJECT (3150 0104). OFFICE OF MANAGEMENT AND BUDGET WASil!NGTON DC 20503. l FACII ITY NAME (1) DOCKET NUM HER (2) l I.ER NUMHER (6)  !

PAGE (3)

LaSalle County Station Unit One 05000373 96 "2W 021 n= 01 l

3 of 9 (If snore space is required. use additional copies of NRC Forrn 366A)(17)

Based on this, a Problem Identification Form (PIF) was initiated to investigate l j

and resolve this issue. LaSalle Station took a conservative approach and '

declared the MCRACS Radiation Monitoring Systems INOPERABLE for LaSalle County Units 1 and 2 until this issue is resolved. The Control Room Ventilation (VC)  ;

l and Auxiliary Electric Equipment Room Ventilation (VE) Systems were already l

). INOPERABLE for other reasons. No additional actions were required because the ,

t MCRACS Radiation Monitoring Systems for LaSalle County Units 1 and 2 are not <

[ required to be OPERABLE with Unit 1 in Operational Condition 4, Cold Shutdown, '

l and Unit 2 defueled.

Review of the Technical Specifications, FSAR, UFSAR and the Design-Criteria for the MCRACS Radiation Monitoring System determined that there was an inconsistency  ;

between the Design Basis as described in the FSAR text and as shown on the FSAR  !

Logic Diagram, and that the original installed design was consistent with the i Design Basis as described in the FSAR Logic Diagram. However, there was an inconsistency between the original installed design and the Design Basis as i

described in the FSAR text. The FSAR Logic Diagram described a two channel j

}

system, either of which would initiate an ESF actuation, compared to the text  ;

which described a two channel system, which required both channels to +

concurrently actuate to initiate an ESF actuation. In addition, during the l investigation, a question arose concerning the Safety Evaluation performed for .

the modifications. The question raised the possibility of an Unreviewed Safety

' l Question related to the trip logic modifications. This issue was documented on -

PIF# 97-0167, i

On January 13, 1997, at 16:00 CST, an engineering review determined that a condition existed in the MCRACS that resulted in the plant being in an unanalyzed l {

condition. The engineering evaluation determined that the Modification to the ,

MCRACS Radiation Monitoring logic installed in 1993 did not meet the single failure criteria required by the Safety Evaluation Report, and increased the number of monitors required to initiate the trip logic which may have increased l the probability of failure of equipment important to safety. This was considered to constitute an Unreviewed Safety Question. This issue was documented in PIF# 97-0241. On January 13, 1997, at 17:55 CST, a four hour ENS phone notification was made to report the plant being in an unanalyzed condition per 10 CFR 50.72 (b) (2) (1) and 10 CFR 50.72 (b) (2) (iii) .

This event is reportable per 10 CFR 50.73 (a) (2) (ii) due to the plant being in an unanalyzed condition. This event is also reportable per 10 CFR 50.73 (a) (2) (1) because this condition could have prevented the fulfillment of a safety system j function needed to ndtigate the consequences of an accident.  ;

i l

I i  !

i I

l NRC FORAt 366 U.S. NUCLEAR REGl*LATORY COMMinlON APPROVED HY OMB NO 3150-0104

,(M2) EXPIRES 05/31'95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITil THIS INFORM ATION COLLECTION REQUEST. 50 0 llRS I ORWARD LICENSEE EVENT REPORT (LER) COMMENTS REG ARDING BURDEN ESTIMATE TO TiiE INFORMATION AND RECORDS MANAGEMENT BRANCH (MNDD 7714), U.S NUCEEAR TEXT CONTINUATION REGULATORY COMMISSION. WASHINGTON, DC 20555-0001 AND TO THE PAPERWORK REDUCTION PROJECT (3150-0104), OrrlCE OF .

MANAGEMENT AND DUDGET WAS!!!NGTON, DC 20503

]

FACII.lTY NAM E (1) IOCKET NUMHER (2) LER NUMHER (6) PAGE (3) I YEAR RQULh71AL RE VlMON NL IMRF R NUMHF R LaSalle County Station Unit one 05000373 96 021 01 4 of 9 (If more space is required, use additional copies of NRC Form 366A)(17)

C. CAUSE OF EVENT The root cause of the inadequate safety evaluation for the modifications performed in 1993 was that LaSalle County Station failed to recognize that damper isolation for the control room ventilation air intakes is not functionally redundant. Hence, the modified design would have allowed a single failure to result in the failure to isolate one train.

Contributing to this event was a failure to conduct an adequate in-depth review of the documentation which comprise the licensing basis. The evaluation of the modification for conformance to single failure criteria, focused on single ,

failure events described in the UFSAR and on the apparent functional redundancy '

of the two trains of ventilation. The safety evaluation determination that the change did not constitute an unreviewed safety question was based upon the text description in section 6.4.4 of the UFSAR. However, this description was in conflict with other design basis documentation as well as in conflict with the facility installed configuration. The root cause of this conflict is not known.

The discrepancy existed at the time of licensing the facility. LaSalle County Station failed to review key documents other than the UFSAR which would have identified these discrepancies. At the time the safety evaluation was performed, LaSalle Safety Evaluation Procedures focused on reviews of the UFSAR and the Technical Specifications and did not specifically direct the reviewers to other license bases such as the FSAR and the SER. Reviews at that time were mainly limited to hard copy searches with limited electronic search capabilities.

In Updating the FSAR, numerous drawings had been removed for simplification of the UFSAR. The system logic diagram in the FSAR which depicted a different design had been incorrectly deleted from the UFSAR and review of the FSAR was nct specifically called out in the Safety Evaluation Procedures at LaSalle County Station at the time.

HISTORICAL BACKGROUND For the purpose of identifying the planned corrective actions, the following historical background is provided. The MCRACS consists of two 100% capacity redundant HVAC trains. Each train is supplied through its' own ventilation i intake. Each ventilation intake is equipped with four radiation monitors j installed and oriented approximately 90 degrees apart around the perimeter of each intake. Normal operating mode is for one of the two trains to be in operation. In the original design, a high radiation signal from any one of the four radiation monitors in an intake resulted in an automatic isolation of the normal outside air and actuation of the Emergency Makeup Filtration mode of the operating train.

Due to the one-out-of-four logic and the very low setpoint (2.5 mR/hr), the system was subject to spurious actuations. In 1989, LaSalle County Station investigated several modifications to the system to reduce the occurrence of spurious actuations. One proposal was to revise the actuation logic to require a signal from minimum of two radiation monitors in order to actuate the emergency mode.

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i NRC Ft>RM M6 U.S. NUCLEAR REGULATORY COMMINSION

, (542) APPROVED in OMB No 3150-0104 EXPIRES 05/31 a95 ESTIMATED BURDEN PER RESPONSE TO COMPLY WITil Tills INFORMATIONCOLLECTION REQUEST: 50.0 llRS. FORWARD  !

LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO Tile INFORMATION AND RECORDS M ANAGEMENT BRANCH (MNBB 7714), U.S. NUCLEAR TEXT CONTINUATION REGULATORY COMMISSION, WASil!NGTON, DC 20555 0001 AND TO l Tl!E PAPERWORK REDUCTION PROJECT (3150-0104), OITICE OF '

MANAGEMENT AND BUDGET, WASilINGTON, DC 20503.

FACILITY NAME (1) DOCKET NUMHER (2) LER NUMHER (6)

( PAGE (3)

YLAR sEQUENMAL KLVISION l

NUMRT R NUMBER

! LaSalle County Station Unit One 05000373 96 021 '

01 5 of 9

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

In April, 1990, LaSalle County Station authorized its AE to proceed with design development of a modification to implement a two out of four actuation logic.

The modification was to be implemented in conjunction with another modification deleting a trip feature from the air intake chlorine detectors.

In Septenber, 1990, due to interface problems with the chlorine detector modification, Station.

further work on the modification was postponed by LaSalle County In January, 1992, LaSalle County Station again authorized the AE to resume development of the modification based on a two out of four logic.

In February, 1992, a draf t plan for the modification was issued for review and  ;

comment. The design proposed incorporated two out of four logic with sufficient i redundancy to satisfy the design basis for the system.

l In March, 1992, the associated Engineering Change Notices (ECN) for the modifications were issued for review and comment.

In June, 1992, the AE was directed by LaSalle County Station to revise the design i to incorporate a " limited two-out-of-four" logic. The reason given for the change was to allow the removal of two monitors at a time from service for ease of maintenance. The change also reduced the complexity and expense of the design change.

The AE incorporated the request and the revised design was issued in  !

September, 1992. This design no longer met the channel redundancy requirements l in the design basis. This was not recognized during the review and approval process as the new design appeared to be a basic simplification of the design change which had been proposed and reviewed numerous times since 1990.

The modification was installed in the facility in July, 1993. i l

D. ASSESSMENT OF SAFETY CONSEQUENCES General Design Criteria 19 of 10 CFR 50, Appendix A, requires a maximum allowable calculated control room dose of 5 rem to the whole body or its equivalent to any organ as a result of a Design Basis Accident. Equivalents to the 5 rem whole body dose are 30 rem to the skin and 30 rem to the thyroid. MCRACS is designed to limit the exposure of the Control Room Personnel to less than these allowable exposure limits. Upon detection of high radiation at the outside air intake, the System is designed to automatically isolate the MCR from the normal outside air, starting the Emergency Makeup Filtration, and initiate a high radiation alarm in the MCR. The modifications introduced an electrical cross connection between radiation monitoring channels that could have resulted in the loss of the i

l automatic isolation feature by a single fault in one radiation monitor. This would have resulted in the operating train continuing to introduce contaminated outside air into the control room during a Design Basis Event prior to. manual action by the MCR operator. The alarm function of the radiation monitors was not affected. ,

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N8 C FDRM 366 U.h. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104 I

, (3(92) EXPIRES 05'3195  !

' ESTIM ATED BURDEN PER RESPONSE TO COMPLY WITH Tills INFORMATION COLLECTION REQUEST: 50 0 }!RS. FORWARD j LICENSEE EVENT REPORT (LER) C MMENTS REGARDING BURDEN ESTIMATE TO Ti!E INFORMATION AND RECORDS MAN AGEMENT BRANCH (MNBil 7714), U.S. NUCLEAR l TEXT CONTINUATION REGULATORY COMMISSION. WASil!NOTON, DC 20555-000) AND TO i

TIIE PAPERWORK REDUCTION PROJECT (3150-0104). OFTICE OF MANAGEMENT ANT) BUDGET. WASHINGTON, DC 20503. '

FACII.ITY NAME (1) DOCKET NUMHER (2) 1 ER NUMBER (6) PAGE (3)  !

YLAR ELQUWRAL RE VISION I NUMRFR NUMRFR 6 l LaSalle County Station Unit One 05000373 96 021 01 6 of 9 (

l (If more space is required, uSe additional copies of NRC Form 366A)(17) l In the event of a design basis accident requiring the isolation of the MCR HVAC system combined with a single failure which disabled the automatic isolation of 3

l the operating MRACS train, the Control Room Personnel would still have received a I high radiation alarm from the operating radiation monitors. The alarm response I for a high radiation alarm is for the operator to confirm that the Emergency l Makeup Train has actuated as required. If the single failure had prevented the i automatic actuation, the control room operator would have manually activated the  !

isolation mode as required by alarm response procedures. t i

However, in accordance with the accident analyses, no manual operator action can be assumed for the first 10 minutes of the accident. In 1993, an analysis was performed of the effect of a delay in MCR isolation due to an increase in the response time of the radiation monitors as a result of a modification. The analysis determined that during normal outside makeup air operation, the maximum increase in delay of isolation without exceeding GDC 19 dose limits is 30 seconds. Based on this analysis, the single failure of the MCRACS isolation combined with a minimum 10 minute operator response time could have resulted in exposure of the MCR personnel exceeding the GDC 19 limits.

E. CORRECTIVE ACTIONS The Safety Evaluation for the modification was performed in September, 1992. The procedures and culture of the Facility at that time was such that sufficient rigor was not always exercised in the Safety Evaluation process. Numerous programmatic weaknesses were subsequently identified with LaSalle County Station's Safety Evaluation process. Corrective actions were taken which included increased training and programmatic changes resulting in greater rigor and thoroughness in the Safety Evaluation and modification review and approval process. The safety Evaluation Procedures were revised to specifically require reviews of the additional documentation which make up the license basis other than the UFSAR and Tech Specs. The documents comprising the license bases are now available for electronic searches which assist in a more thorough review and i evaluation process. 1

, 1. The potential extent of condition for this type problem will be evaluated I

as part of the System Functional Performance Review Program. This program is being conducted for all systems important to safe and reliable operation 1

and includes 1) determining the required system functions derived from the

design bases, 2) identifying materiel condition problems that affect j achieving these functions and 3) ensuring the periodic' testing requirements  ;

adequately confirm system functions. Corrective actions including design i i

' changes and maintenance activities will be implemented when required to I ensure the system functions are achieved. If substantive functional  !'

I problems are encountered, a detailed design review will be performed to

!- confirm whether supporting detailed analyses are available and identify necessary design changes.

I

2. Additionally, LaSalle County Station is implementing a long-term plan for preparation of a major scope of design bases documents, verification of f

these design bases documents with other documentation, and plant system l verification. This effort will include reconstitution of selected analyses I

and calculations, improvements in calculation control and UFSAR validation.

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, NRC PORM 366 U.S. NUCLEAR REGULATORY COMMISSION APPROVED BY OMB NO. 3150-0104

, (>02) EXPIRE $ 05 31a93 i ESTIMATED BURDEN PER RESPONSE TO COMPLY WITil TlitS INFORMATION COLLECTION REQUEST. 50 0 liRS FORWARD i LICENSEE EVENT REPORT (LER) C MMENTS REGARDING BURDEN ESTIMATE TO TIIEINFORMATION AND RECORDS MANAGEMENT BRANCil(MNBB 7714), U.S. NUCLEAR TEXT CONTINUATION REGULATORY COMMISSION, WASil!NOTON, DC 20$$5-0001, AND TO Tile PAPERWORK REDUCTION PROJECT (3150-0104). OFFICE OF MANAGEMENT AND BUDGET. WASil!NGTON, DC 20503.

I FACILITY NAM E (1) DOCKET NUMHER (2) LER NUMHER (6) PAGE (3) J YLAR &bQULt(FIAL ILE VihlON NUMBFR NUMRFR LaSalle County Station Unit One 05000373 96 021 01 (If more space is rcquired, use additional copies of NRC Form 366A)(17) 7 of E )

t All the following corrective actions will be taken prior to restart of either i Unit 1 or 2. )

1. A modification will be developed to correct the installed design. The original channel separation and redundancy will be restored. The revised design will meet the single failure criterion of the design basis.

2. The Technical Specifications will be revised to conform to the design and to eliminate confusion in the wording of the action statement.

3. An ongoing detailed system design review of the MCRACS is already in progress as a result of previous concerns as identifjed in LER 96017.

4. A review of other modifications to the MCR Ventilation System which could affect the single failure criteria as it applies to the MCR Isolation Dampers and Control Circuitry will be conducted.

5. A review of the changes which removed FSAR drawings from the UFSAR to ensure that the information contained on the drawings is adequately referenced in the UFSAR for subsequent Safety Reviews.

F. PREVIOUS OCCURRENCES LER NUMBER TITLE None.

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

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l NRCCORM 366 U.S. N UCLEAR REGULATORY CON 1 MISSION (Srf2) APPROVED BY OMI) NO. 3150-0104 i EXPIRES 05?31.95 l

i ESTlhiATED BURDEN PER RESPONSE TO COhlPL'f WITil Tills 3 INFORA1ATION COLLECTION REQUEST: 50.0 llRS. FORWARD LICENSEE EVENT REPORT (LER) C MMENTS REGARDING BURDEN ESTlh1 ATE TO Tile INFONATION AND RECORDS MANAGEMENT BRANCil(MNBB 7714), U.S. CLEAR TEXT CONTINUATION REGl'LATORY COMMISSION. WASillNOTON. DC 20555 000). ANDTO Tile PAPERWORK REDUCTION PROJECT (3150-0104). 0FFICE OF MANAGEMENT AND BUDGET, WASillNGTON, DC 20503 FACll,ITY NAM E (1) IXWKET NUMHER (2) I.ER NUMHER (O PAGE (3)

YEAR hl.QUENTIAL RE VlkluN NL MBF R NUMBFR LaSalle county station Unit one 05000373 96 021 01 8 of 9  ;

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

! ATTACHMENT A l

UNIT 1(2) MCRACS RADIATION MONITOR TRIP LOGIC PRIOR 20 INSTAT.TATION OF MODIFICATION M01-88-003A(B)

==\ 1(2)Dl8-K751A = = 1(2) Dis-K7518 , R SET

== 1(2}D18 K751C ,= = 1(2)O18-K75tO

== SEAL-IN (close on N rec)/ \(close on N red) -= SEAL-IN I l

k ,

SENL IN DAMPERS DAMMERS DAhAPERS DAMPERS SEAblN a a EMUFAN l EMUFAN CHANNEL 1 CHANNEL 2 (Representative Configuration for Illustration Only) l

. - .___._m_ _ _ = _ _ _ - . _ - _ . . . _ m . - __ . _ _ _

l NHC NDRM 366 U.S. N UCLEAR REGULATORY COMMISSION APPROVED BY OMll NO. 3150 0104 (g-92) EXPIRES 03/31.95 l ESTIMATED DURDEN PER RESPONSE TO COMPLY WITil Tills INFORMATION COLLECTION REQUEST: $0.O llRS. FORWARD LICENSEE EVENT REPORT (LER) COMMENTS REGARDING BURDEN ESTIMATE TO TIIE INFORMATION AND RECORDS MANAGEMENT BRANCil(MNBB 7714), U.S. NUCLEAR TEXT CONTINUATION REOULATORY COMMISSION. WASillNGTON. DC 205$5 0001. AND TO Tile PAPERWORK REDUCTION PROJECT (3150 0104), OFFICE OF MANAGEMENT AND BUDGET. WASillNGTON DC 20503.

FACII.lTY N AM E (1) DOCKET NUhlitER (2) 1.ER NUMitER (6) PAGE (3)

YEAR bEQULPmA1. RE VihlON NUMRT R NUMRFR LaSalle County Station Unit One 05000373 96 021 01 9 of 9 (if more space iS required, use additional copicS of NRC Form 366A)(17)

ATTACHMENT B UNIT 1 (2) MCRACS RADIATION MONITOR TRIP LOGIC AFTER INSTAT TATION OF MODIFICATION M01-0-88-003A(B) f I

JClose on hi red)

= = l(2)O 8-K[5[A 1(2)DlS-K[5[C PB

=- SEAL-w = ='1(2)D18-K751B == 1(2)Dl8-K75tD SEAL.W I

(close on hi rad)/ I SEL-IN A DAMPERS DAM ERS DA OERS DAMPERS SEA'L'-IN a a EMU FAN EMU FAN CHANNEL 1 CIIANNEL 2 (Representative Configuration for Illustration Only) 1 l

l