ML19344D665

Forwards Response to NUREG-0578 & Design Criteria & Conceptual Design Description for Technical Support Ctrs for Facilities for Review.Util Emergency Plan Will Be Submitted W/Lasalle County site-specific Emergency Plan

ML19344D665
Person / Time
Site:	Dresden, Quad Cities, Zion, LaSalle
Issue date:	04/15/1980
From:	Peoples D COMMONWEALTH EDISON CO.
To:	Eisenhut D Office of Nuclear Reactor Regulation
Shared Package
ML19344D666	List: ML19344D665 ML19344D667
References
NUDOCS 8004250427
Download: ML19344D665 (60)

Text

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f N Commonwealth Edison

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~ "J Address Reply to: Post Office Box 767 (j Chicago, liknois 60690 April 15, 1980 Mr. Darrell G. Eisenhut, Director Division of Project Management Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555

Subject:

LaSalle County Station Unit 1 NTOL Action Plan -

Short Term Lessons Learned Response NRC Docket No. 50-373 Reference (a):

D. B.

Vassallo letter to "All Pending License Applicants," dated November 9, 1979 (b): D. B.

Vassallo letter to "All Pending Operating License Applicants," dated September 27, 1979 (c): D. L.

Peoples letter to T. A.

Ippolito dated January 3, 1980 (d): B. K. Grimes letter to "All Power Reactor Licensees and Applicants for a License to Operate," dated March 10, 1980

Dear Mr. Eisenhut:

Enclosed is Commonwealth Edison's commitment to meet the requirements outlined in Reference (a) for "Pending License Applicants" for our LaSalle County Station Unit 1.

The materials presented in the enclosure will be incorporated into the Final Safety Analysis Report as suggested in Reference (b).

In addition, the Commonwealth Edison Generating Station Emergency Plan (GSEP) submitted.in Reference (c) is being revised to address the guidance outlined in NUREG-0654 (Reference (d)) and is expected to be resubmitted along with the LaSalle County Site Specific (mergency Plan Annex within the next few weeks.

Although we have and will continue to monitor the development of the NUREG-0660 NTOL Action Plan, with the intention of addressing it on the LaSalle County docket when finalized, it is requested that the enclosed materials supplemented oy the emergency preparedness program revisions discussed above De reviewed immediately and a.ly comments you may nave communicatec as soon as ggyg 3

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Commonwe:lth Edison Mr. Darrell G. Eisenhut April 15, 1980 Page 2 possible to facilitate closure of the operating license review for LaSalle County Unit 1.

One (1) signedLoriginal and thirty-nine (39) copies of the subject report are enclosed.

Very truly.yours, j7 _.

D.

L.

Peoples Director of Nuclear Licensing Enclosure 3065A i

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i Unit 1 Near Term Operatino License l

Action Plan

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Short Term " Lessons Learned" Response

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Reference:

NUREG-0578) l s

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April, 1980 TABLE OF CONTENTS (REFERENCE NUREG-0578)

Section Title Paae 2.1.1 Emergency Power Supp1y.............................

1 2.1.2 Performance Testing for BWR and PWR Relief.........

3 and Safety Valves 2.1.3.a Direct Indication of Power-Operated Relief......... 4 Valve, Valve and Safety Valve Position for PWRs and BWRS 2.1.3.b Instrumentation for Detection of Inadequate Core Cooling......................................

5 2.1.4 Diverse Containment Isolation..................... 6 2.1.5.a Dedicated H2 Control Penetrations

......y.......

30 2.1.5.c Capability to Install Hydrogen Recombine; at......

31 each Light Water Nuclear Power Plant 2.1.6.a Integrity o f Systems Outside Containme'at.....

.... 32 Likely to Contain Radioactive Materials for PWRs and BWRs 2.1.6.b Design Review of Plant Shielding and............. 35 Environmental Qualification of Equipment for Spaces / Systems Which May Be Used In Post Accident Operations 2.1.7.a Auto Initiation of the Auxiliary Feedwater........ 37 Systems (AFSW) 2.1.7.b Auxiliary Feedwater Flow Indication to........... 38 Steam Generators 2.1.8.a Post-Accident Sampling Capability.................

39 2.1.8.b Increased Range o f Radiation Monitors............. 41

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2.1.8.c Improved In-Plant Iodine Instrumentation.......... 43 i

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Section Title Page 2.1.9 Transient and Accident Analysis...................

44 Containment Pressure Indication (ACRS)............

45 Containment Water Level Indication (ACRS).........

46

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Containment Hydrogen Indication (ACRS)............

47 Reactor Coolant System Venting (NRR)..............

48 2.2.1.a Shif t Supervisor Responsibilities................. 50 2.1.a.b Shift Technical Advisor...........................

51 2.2.1.c Shif t and Relie f Turnover Procedures.............. 52 2.2.2.a Control Room Access...............................

53 2.2.2.b Onsite Technical Support Center...................

54 2.2.2.c Onsite Operational Support Center.................

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APRIL, 1960 LSCS - NTOL ACTION PLAN 2.1.1 Emeroency Power Suoply for Relief Valves to Reactor LaSalle Units 1 & 2 have four main steam lines on which are mounted eighteen dual function Crosby safety / relief valves (SRV's).

These valves are physically separated (quadrants) into groups of 4-5-4-5 around the reactor vessel.

Each SRV is separately piped to a floor-mounted T-quencher in the suppression pool.

The blow-down piping, SRV's, and quenchers are all seismically qualified ASME Class III hardware.

These CrosLy valves function as direct-acting safety valves t

where the valve is lifted from the seat when the increasing steam

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pressure (force) as the valve inlet exceeds the spring force.

The pop-open-to-capacity flow results from two-stages of reaction working together to produce a continuous pop action:

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initial lift from the valve face, and i

b.

subsequent augmenting lift from the flow of escaping steam such that the valve attains a smooth lift equal j

to or greater that the capacity lift.

The Crosby valve has internal balancing bellows and balancing piston to eliminate internal seat back-pressure which could affect the i

valve set point.

The same valve seat can be control-operated by an air cylinder mounted on the valve as an integral part of the lifting gear linkage.

This relief mode of operation is user-commanded by

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electrical activation of'the solenoid valves that control the air piston output.

Other features' include setable blowdown control, pressure assisted reclosing action. positive reseating, and integral valve-position indication that is remotely indicated in the control room.

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The Crosby SRV requires no external power to operate in the I

safety mode.

In the relief mode, the Crosby SRV requires only DC power and instrument air to actuate one of two solenoid valves thus j

dumping the line pressure at some point less than the self actuating set pressure.

At LaSalle, the four physically separated valve I

groups are powered by two electrical divisions of 125-volt ESS power.

Physically adjacent quadrants have separate electrical power divisions.

i The relief mode of valve operation can be excerised both i

manually from the control room or automatically by either.the Automatic Depressurization Sysem (ADS) or by the pressurizer (1120 psia).

This control circuit simply actuates the solenoid. valve which opens the pneumatic air supply to the driving piston.

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P The fully isolated capacity of the air accumulators is sufficient for multiple actuations.

there is no limit on the number of actuations.With its normally

supply, Over-pressure protection,There are no block valves in the main steam rel independent of external electric or pneumatic supplies, i s provided by a minimum of one valve from the available eighteen SRV's.

groups can provide lcng-term manual blow-down when its controlA solenoid is actuated from the safety-grade ESS power supply.

safety system activation and control are also powered by emerger.cy power at LaSalle.

In addition, as described in section 9.3.1.2.2 of the FSAR each c f the seven ADS valves has two accumulators for manual re actuations as well as,seprate nitrogen bottle banks that remain on line to supply the AD3 valves in the ev or a loss of the pneumatic compressors.ent of either a loss of power For long term cooling operation, comtressors from Diesel Generator Buses. procedural steps can be taken to f

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APRIL, 1980 LSCS - NTOL ACTION PLAN 2.1.2 Performance Testing for BWR Relief and Safety Valves Although the BWR design basis includes no transients or accidents in which two-phase flow or subcooled liquid flow at high pressure is calculated or expected, liquid flow is expected for an alternate shutdown mode at LaSalle at low pressure under operator control and is possible at high pressure if control. system failures are assumed.

Therefore, in view of the NRC's desire for demonstration tests of all BWR relief and combined safety-relief valves, the BWR owner's group is developing a specific test program.

includes performance tests on the Crosby SRV. Edison will particip Prior calculations of this SRV liquid capacity at saturated conditions during low pressure operation in the alternate cooling configuration indicate a wide margin beyond the flow needed for alternate cooling capacity.

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APRIL, 1980 LSCS - NTOL ACTION PLAN 2.1.3a Direct Indication of Power Operated Relief Valve and Safety valve Position LaSalle Units 1 & 2 utilize the Crosby combined safety / relief valve (SRV).

This valve has a central spindle rod which moves vertically through the spring which provides seating force to hold the valve closed against the steam-line pressure (force) at the SRV inlet.

A mechanical crank and cam assembly is mounted at the valve top outside the bonnet which encloses the spring.

The spindle rod extends through the bonnet so that it can be lif ted mechanically by the crank for relief-mode operation.

It is lifted mechanically by steam pressure acting on the valve face for safety-mode operation.

This dual-mode capability is the reason for Crosby valves are called safety / relief valves.

At LaSalle, an electromechanical lift indicating assembly is directly mounted atop the SRV.

It has its own housing,which mechanically mates to the valve bonnet.

A reverse-spring-loaded actuator rod rides the end of the valve spindle rod to directly transmit valve motion relative to the valve seating surface.

Actuator-rod positions (fully open or fully closed) are sensed by dual microswitches through mechanical contact arms that ride the actuator rod.

Electrical outputs from the microswitches are fed to the control room to remotely indicate SRV position there.

Event annunciation is also provided in the control room.

These redundant, single channel data circuits are separated physically and electrically consistent with IEE 279 criteria and in a manner consistent with the ESS divisions to which control solenoid (for relief function) is assigned for electrical power.

A confirmatory indication of SRV popping or long trend leakage is provided via temperature elements mounted in thermowells on each of the SRV blowdown pipes to the suppression pool.

These indications are for back-up confirmation of the direct-indicating l

SRV position read-outs.

Environmental and seismic qualification of the i

electro-mechanical position sensore and control room indicators is j

currently underway.

The LaSalle equipment qualification programs is

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scheduled for completion during calendar year 1980.

Cabling, l

circuitry, and mounting panels are qualified to IEEE 344 (1975) and IEEE~323 (1974) standards.

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ACTION PLAN r

2.1.3b Instrumentation for Detection of Inadeouate-Core Coolino in PWR's and BWR's t

A subcooling meter is not needed in the BWR because the BWR operates in'all power modes with liquid and' steam in the reactor pressure vessel; thus saturation conditions are always maintained irrespective of system pressure.

The BWR Owners Group of which CECO is a member submitted GE document NEDO 24708 (Aug. 1979) to the NRC staff in response to requests for information or vessel level, transient analyses, etc.

Section 2.3.2, Reactcr Water Level Instrumentation, contains an indepth analysis of EWR level instrumentation.

The conclusion of that analysis is that no additional hardware is needed to identify inadequate core cooling in the BWR.

Further analyses and operator guidelines for the detection and mitigation of inadequate core cooling are being developed per TM1 requirement 2.1.9 and responses to B&O Task Force questions.

These studies include an evaluation of currently installed reactor vessel water level instrumentation, and possible use of other instruments, to detect inadequate core cooling.

The need for further measures, if any, will be addressed after these analyses and operator guidelines are completed.

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April, 1980 LSCS - NTOL ACTION PLAN 2.1.4 Containment Isolation The containment isolation system for LaSalle has been reviewed as required by NUREG 0578 to assure that:

1.

Diverse containment isolation signals that satisfy safety-grade requirements exist.

A summary of primary containment isolation signals given in table A.

2.

That essential and non-essentisl systems are identified.

Essential and non-essential systems for the purpose of isolation are identified by penetration in Table B.

Essential systems are those that may be needed within 10 minutes of a LOCA, a normal reactor scram or a scram system failure.

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That non-essential system are automatically isclated by containment isolation signals.

All non-essential systems that provide a possible open path out of the primary containment were found to be either isolated by isolation signals, by check valves that would prevent flow out of the containment, by manual valves that are normally closed during reactor operation, or as in the case of instrument lines by closed piping systems.

In the case of small diameter instrument lines which penetrate the containment the LaSalle design meet Reg. Guide 1.11 using excess flow check valves on instruments connected to the primary system and automatic isolation valves on non-essential containment instrumentation.

4 Resetting of containment isolation signals shall not result in the automatic loss of containment isolation.

This review revealed only one case in which primary containment isolation is removed by resetting of a j

containment isolation signal.

This was on the t

recirculation loop hydraulic lines to the flow control valves.

These valves will be modified to j

prevent automatic reopening upon reset.

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TABLE A ISOLATION SIGNAL CODES SIGNAL DESCRIPTION A

Reactor vessel low water level leve.' 3 - (A scram occurs at this level also.

This is the higher of the two low water level signals.)

B Reactor vessel low low water level level 2 - (The RCIC and HPCS systems are initiated at this level also.

This is the lower of the two low water level signals.)

C High radiation - Main steam D

Line break - High area temperature or very high system flow.

E Main condenser low vacuum.

F High drywell pressure.

G Reactor vessel very low water level (Level 1) or high drywell pIessure (Emergency Core Cooling System are i

started).

J Line break in cleanup system - high space temperature.

M Line break in RHR shutdown and head cooling (high space temperature).

P Low main steamline pressure at inlet turbine (RUN mode only).

U High reactor vessel pressure - close RHR shutdown cooling valves and head cooling valves.

Y High radiation, fuel pool ventilation exhaust.

Z High radiation, reactor building ventilation exhaust.

i RM Remote manual switch from control room.

(All regular Class A and Class B isolation valves are capable of remote manual opert tion from the control room.)

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.- I, TABLE B

0:4TAIN'4ENT ISOLATION ENETRATIOff VALVE NUMBER SYSTEM NUMBER FIGURE CLASSIFICATION SIGNAL COMMEtiTS 4-1 to M-4 Main Steam (includes 1821-F022A,B,C,0 Detail (a)

E B,C,D,E,P,RM drain line and 1821-F028A,B,C,0 8,C,0,E,P,RM Note 1 MSIV - LCS line) 1821-F067A,B,C,0 B,C,0,E,P,RM RM j

1E32-F001A,B,C,0 45 & M-6 Reactor Feed 1821-F010A,B Detail (b)

E Rev. Flow (includes connection 1821-F032A,B B,F, Rev. Flow Note 2 to RWCU) 1821-F065A,B RM (Note 6) 1G33-F040 RM (Note 6) 1-7 RHRS/ Shutdown lE12-F009 Detail (t)

NE A,0,U,RM Suction lE12-F006 A,D,U,RM 9-8 & M-9 RHRS/ Shutdown lE12-F050A,B Detail (d)

NE Rev. Flow Return lE12-F053A,B A,D,U,RM Note 3 1E12-F099A,B A,0,F,0,RM M-10 LPCS lE21-F006 Detail (d)

E Rev. Flow Injection RM (Note 7) lE21-F005 lE21-F354 B,F,RM i

M-ll HPCS lE22-F005 Detail (d)

E Rev. Flow Injection RM (note 7) lE22-F004 lE22-F354 B,F,RM N,

M-12 to M-14 RHR/LPCI lE12-F041A,B,#.C" Detail (d)

E Rev. Flow Injection f'

lE12,F042A,B,C RM (Note 7) 1E12-F327A,B,C G, c,.RM M-15 Steam to RCIC lE51-F063 Detail (e)

E 0,RM System (includes lE51-F076 D,RM RHR Supply) lE51-F064 0,RM lE51-F008 0,RM _

1 JONTAINMENT

'ENETRATION VALVE ISOLATION

, NUMBER-SYSTEM NUMBER FIGURE CLASSIFICATION SIGNAL-COMMENTS M-16 Cooling LWR 029 Detail (f)

NE B,F,RM Water Supply LWR 179 B,F,RM W-17 Cooling 1WR040 Detail (f)

NE B,F,RM Water Return 1WR180 8,F,RM M-18 & M-19 RHRS/ Containment lE12-F017A,B Detail (g)

NE G,RM Note 4)

Spray 1E12-F016A,8 G,RM Note 4)

M-20 Vent to IVQO30 Detail (s)

NE 8,F,Y,Z,RM Drywell IVQO29 B,F,Y,Z,RM M-21 Vent from IVQ034 Detail (h)

NE F,B,Y,Z,RM Drywell IVQO35 F,B,Y,Z,RM lVQ036 F,B,Y,Z,RM IVQ068 F,B,Y,Z,RM M-22 Main Steam IB21-F016 Detail (c)

NE B,C,D,E,P,RM Drains IB21-F019 B,C,D,E,P,RM M-23 & M-24 Spare lM-25 & M-26 Chilled 1VP063A,B Detail (f)

NE B,F,RM Water Sapply IVP113A,B B,F,RM M-27 & M-28 Chilled lVP053A,B Detail (f)

NE B,F,RM Water Return IVPil4A,B B,F,RM

.M-29 RCIC RPV IE51-F066 Detail (i)

NE Rev. Flow Rev. Flow Head Spray IE51-F065 (Includes RHR RM Head Spray)

IE51-F013 IE12-F023 A,0,U,Rf1 IE51-F355 B,F,RM IE51-F354 B,F,PM

~M-30 Reactor IG33-F001 Detail (c)

NE 8,J,RM Cleanup IG33-F004 B,J,RM

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-NUMBER SYSTEM NUMBER FIGURE CLASSIFICATION SIGNAL COMMENTS M-31 & M-32 Spare M-3; & M-53 Combustible IHG001A,B Detail (g)

NE RM (Note 7)

Gas Control RM (Note 7)

'Drywell Suction IHG002A,8 M-34 Standby IC41-F007 Detail (u)

E Rev. Flow Rev. Flow Liquid IC41-F006 NA Control IC41-F004A M-35 M-36 Recirc.

IB33-F019 Detail (t)

NE B,0,RM 8,C,Rif Loop IB33-F020 Sampling M-37 Clean IMC033 Detail (v)

NE NA Locked Closed NA Condensate IMC027 M-38 Service Air ISA046 Detail (v)

NE NA Locked Closed NA ISA042 M-39 M-40 A,B,C,0 CRD ICll-0001-120 Note (9)

E A,RM Typical of 185 Insert A,RM Typical ICll-0001-123 of 185 M-41 A,B,C,D CRD ICll-0001-121 Note (9)

E A,RM Typical of 185 Withdrawal ICll-0001-122 A,RM Typical of 185 M-42 to M-46 TIP Drive IC51-J004 Note (10)

NE A,F,RM M-47 Air Supply IIN-031 NE B,F,RM M-55 ADS Pneumatic IIN101 Detail (j)

E NA Supply.

E CONTAINMENT PENETRATION VALVE ISOLATION NUMBER-SYSTEM NUMBER FIGURE CLASSIFICATION SIGNAL-COMMENTS M-49 & M-50 Recirc. flow IB33-F338A,8 Detail (c)

NE B,F,RM Control Valve IB33-F339A,B Detail (c)

B, F,Ri1 Hydraulic IB33-F340A,8 Detail (c)

B,F,RM Note (5)

Piping IB33-F341A,8 Detail (c)

B,F,RM IB33-F342A,B Detail (c)

B,F,RM IB33-F343A,8 Detail (c)

B,F,RM IB33-F344A,B Detail (c)

B,F,RM IB33-F345A,B Detail (c)

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-2 Spare M-52 Spare M-54 Spare M-48 Spare M-56 Spare M-57 Spare M-58 Spare M-59 Clean Condensate 1FCll3 Detail (v)

. NE NA Locked closed NA to Refueling Bellows 1FCll4

• M-60A Drywell Pneumatic IIN018 Detail (9)

NE Rev. Flow Compressor 11N017 B,F,RM Discharge M-60B Air Dryer IIN074 Detail (g)

NE B,F,RM Blowdown lIN075 B,F,RM M-61 ADS Pneumatic IIN100 Detail (j)

E NA Supply M-62 Drywell Pneumatic IIN001A Detail (g)

NE B,F,RM Compressor Suction lIN001B B,F,RM f

CONTAINMEriT ISOLATI0tl VALVE NUMBER-SYSTEM

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• PEtlETRATION M-63 & M-64 Recirc. Pump Seal IB33-F013A,B Detail (h)

NE Reverse Flow Injection Supply 1833-F017A,8 flote (25)

Reverse Flow M-65 Reactor Well 1FCll5 Detail (v) tlE '

NA Locked Closed NA Bulkhead Drain 1FC086 M-66 Suppression IVQO27 Detail (s)

NE F,A,Y,Z,RM F,A,Y,7.RM Chamber IV0026 Purge Line M-67 Suppression IVQ031 Detail (h)

NE F,A,Y,Z,RM F,A,Y,Z,9M Chamber Vent IVQ040 F,A,Y,Z,RM Line IVQO32 M-68 LPCA Suction lE21,F001 Detail (m)

E RM from Suppression Pool M-69 HPCS Suction From lE22-F015 Detail (m)

E RM Suppression Pool M-70 to M-72 RHR (LPCI) Suction IE12-F004A,B,C Detail (m)

E RM From Suppression Pool M-73 & M-74 RHR to Suppression IE12-F027A,8 Detail (z)

NE G,RM Pool Spray Header M-75' RCIC Pump Suction lE51-F031 Detail (m)

E RM From Suppression Pool M-76 RCIC Turbine IE51-F068 Detail (o)

E RM Reverse Flow Exhaust lE51-F040 M-77 LPCS Min.

lE21-F012 Detail (1)

E RM (Note 7)

RM (note 7)

Flow Line IE21-F011 LPCS Test Line,

o' CONTAINMENT PENETRATION VALVE ISOLATION NUMBER-SYSTEM NUMBER FIGURE CLASSIFICATION SIGNAL COMMEf!TS M-78 M-79 & M-84 RHR Min. Flow lE12-F024A,B Detail (q)

E G,RM Line RHR Test Line lE12-F021 G,RM lE12-F302 lE12-F064A,B,C RM (Note 7) lE12-F0llA,8 G,Rf1 (flote 7) lE12-F088A,B M-80 RCIC Pump Min.

1E51-F019 Detail (r)

E RM (Note 7)

Flow Line M-81 RCIC Vacuum lE51-F069 Detail (r)

E RM (flote 7)

Pump Discharge lE51-F023 Reverse Flow M-82 HPCS Min. Flow 1E22-F023 Detail (1)

E G,RM Line HPCS Test Line 1E22-F012 G,RM M-83 & M-93 LPCS Safety / Relief lE21-F018 Detail (r)

E Process Valve Discharge lE21-Fo31 Process M-85, M-86, RHR Safety / Relief lE12-F025A,B,C Detail (r)

E Process

M-87, Valve Discharge Process M-90, 1E12-F030 Process M-91, lE12-F030 Process M-99 lE12-F005 Process M-88 & M-89 RHR Safety / Relief IE12-F073A,8 Detail (p)

E RM (flote 7)

Valve Discharge lE12-F074A,B RM (flote 7) and Hx Vent Line lE12-F055A,B Process lE12-F311A,B Process lE12-F103A,8 Process lE12-F104A,8 Process M-92 RCIC Safety /

lE12-F036 Detail (r)

E Process Relief Valve Discharge M-94 HPCS Safety / Relief 1E22-F014 Detail (r)

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~ NUMBER SYSTEM NUMBER FIGURE CLASSIFICATION SIGNAL-COMMENTS H-95 & M-102 Combustible Gas lHG005A,B Detail (g)

NE RM (Note 8)

Control Return lHG006A,B RM (Note 8)

M-96 Drywell Equip.

1RE025 Cetail (g)

NE B,F,RM Drains 1RE024 B,F,RM M-97 Drywell Equip.

1RE029 Detail (g)

NE B.F RM Drain Cooling 1RE026 B,F,RM M-98 Drywell Floor 1RF012 Detail (g)

NE B,F,RM Drains 1RF013 B,F,RM M-100 M-101 RCIC Turbine lE51-F080 Detail (c)

E F,RM Exhaust IE51-F086 F,RM Breaker Line

'M-103 to M-110 Vacuum Breaker 1PC003A,B,C,0 Detail (y)

E NA 1PC002A,B,C,0 NA 1PC001A,B,C,0 Pressure Differential --

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,_.~.n NOTES 1.

The MSIV-LCS valve is required to open following a LOCA.

The signals that close the valve are 1) high steamline pressure,

2) excessive leakage through the inboard MSIV, and 3) insufficient dilution air flow.

2.

The check valves on the feedwater returr. lines are provided wi.th an air operator for testing the valves to ensure that the disks are not frozen in the open position.

The actuator moves the disk partially into the flow stream, but is not capable of completely closing the valve against flow.

The feedwater valve actuator is used to apply seating force to the valve for ensuring leaktightness at low differential pressures.

The actuator will be exercised to assure operability prior to leak testing.

3.

Testable check valves are provided with an air operator for remote opening with zero differential pressure across the valve seat.

These valves will close on reverse flow even though the test switche; may be positioned for open.

The valves open when pump pressure exceeds reactor pressure even though the test switch may be closed.

4.

Containment spray and suppression cooling valves have interlocks that allow them to be manually reopened after automatic closure.

This setup permits containment spray, for high drywell pressure conditions, and/or suppression water cooling.

5.

To satisfy the requirements of General Design Criteria 55 and system functionality, these instrument lines which penetrate the primary containment from the reactor coolant pressure boundary conform to Regulatory Guide 1.11 in that they are equipped with a restricting orifice located inside the primary containment and a manual shutoff valve and an excess flow check valve located outside and as close as practicable to the primary containment.

The lines are Seismic Category I and terminate in Seismic Category I instrumentation.

Isolation is provided by the excess flow check valve.

In the event of a line rupture, downstream of the check valve, this valve would close to limit the amount of leakage.

The flow-restricting orifice is sized to assure that in the event of a postulated failure of the piping or component, the leakage is hydraulic fluid from the FCV hydraulic system which does not communicate with the containment atmosphere.

The function of these lines will be tested during reactor plant operation.

These lines and their associated instruments will normally be pressurized to reactor operating pressure.

l Surveillance inspections are performed to ensure the leaktight l

integrity of these lines and their associated instruments.

Additional inservice inspection is included in the Technical Specifications.

This inservice inspection verifies the function of the excess flow check valves and their leakage

! i rates.

i 6.

These valves are provided for long-term leaktightness only.

Feedwater check valves in each line provide immediate isolation.

These MO valves are remote manually closed from the control room upon indication of loss of feedwater flow.

Therefore, no additional isolation signcls are required.

7.

Although only one isolation valve signal is indicated for these valves, the valves also receive automatic signals from various system operational parameters.

For example, the ECCS pump minimum flow valves close automatically when adequate flow is achieved in the system; the ECCS test lines close automatically on receipt of an accident signal.

Although these signals are not considered isolation signals; and are therfore, excluded from this table, there are other system operation signals that control these valves to ensure their proper position for safe shutdown.

8.

These valves are required to open on signals B and F during the post-LOCA conditions.

They remain closed during all other plant operating states, except cold shutdown.

Therefore, there is no reason to provide them with any isolation signal other than remote manual.

9.

Criterion 55 concerns those lines of the recrtor coolant pressure boundary oenetrating the primary reactor containment.

The control rod drive (CRO) insert and withdraw lines are not part of the reactor coolant pressure boundary.

The basis to which the CRD lines are designed is commensurate with the safety importance of isolating these lines.

Since these lines are vital to the scram function, their operability is of utmost concern.

In the design of this system, it has been accepted practice to omit autonatic valves for isolation purposes, as this introduces a possible failure mechanism.

As a means of providing positive actuation, manual shutoff valves (1C110001-101 and -102) are used.

The charging water, drive water and cooling water headers are provided with a check valve (1C11D001-115, -138 and -137) within the hydraulic control unit (HCU), a Seismic Category I module, and'the normally closed solenoid valves (1C110001-120, -121,.-122 and

-123).

These valves will prevent any direct flow away from containment.

If an insert line fails, a ball check valve provided in each drive is designed to seal off the broken line by using reactor pressure to shift the ball check vt1ve to the upper seat.

This feature also prevents any direct flow away from the primary containment.

- 1

A piping integrity test is accomplished for leaks of the HCU's during daily inspection (HCU operating pressure above 1000 psi).

In addition, several indictors ir the control room, such as temperature and pressure of CRD cooling water or drywell sump pump operation, indicates whether leakage is excessive.

The maximum leakage expectec at this penetration is 3 gpm when the RPV is still pressurized (about 1000 psi).

Tnis leakage also assumes a single active failure of a Check valve inside the HCU.

After the reactor vessel is depressurized, the CRD leakage will decrease' to about 0.5 gpm.

It may also be said that leakage monitoring of the CRD insert and withdraw lines is provided by the overall type A leakage rate test.

Since the RPV and ncnseismic portions of the CRD system are vented during the performance of the Type A test, any leakage from these lines would be included in the total Type A test leakage.

The flowout of the CRD is restricted through the HCU performance test requirements to ensure that HCU leakage does not exceed 0.2 gpm.

The maximum leakage expected for these penetrations is 0.2 gpm per HCU.

If a single failure is assumed, the maximum leakage would be 3 gpm.

Seismic tests have demonstrated the seal integrity of the CRD system.

Maximum leakage following these tests did not exceed 3 gpm.

The system design criteria are as follows:

Quality Seismic Quality Group Assurance Cateoory Classification Classification Valves; insert and withdraw I

B I

Insert and withdraw line piping I

B I

The CRD insert and withdraw lines are compatible with the criteria intended by 10 CFR 50, Appendix J for Type C testing, since the acceptance criterion for Type C testing allows demonstration of fluid leakage rates by associated bases.

The maximum leakage expected has been factored in with the total allowable containment penetration leakage and determined to be acceptable.

10.

The TIP drive guide tubes provide a sealed path for the flexible drive cable of the TIP probes.

The TIP tubing seals the TIP system from the reactor coolant and forms a leak tight boundary designed for reactor coolant pressure boundary conditions.

The shear valve is provided to cut the cable in the event that the drive cable cannot be withdrawn, and the ball provides the guide tubes with shut-off capability.

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The LaSalle TIP system design specifications require that the maximum leakage rete of the ball and shear valves shall be in accordance with the Manufacturers Standardization Society (Hydrostatic Testing of Valves).

The ball valves are 100%

Icak tested to the following criteria by the manufacturer:

Pressure 0 - 62 psig Temperature 3400F Leak Rate 10-33 s.

/

A statistically chosen sample of the shear valves is tested by the manufacturer to the following criteria:

~

Pressure 0 - 125 psig Temperature 3400F 10-3cm3 sec STP.

/

Leak Rate The shear valves have explosive squibs and require testing to destruction.

They cannot therefore be 100% tested.

Since the traversing incore probe (TIP) system lines do not communicate freely with the containment atmosphere or the reactor coolant, General Design Criteria 55 and 56 are not directly applicable to this specific class of lines.

The basis to which these lines are designed is more closely described by General Design Criterion 54, which states in effect that isolation capability of a system.should be commensurate with the safety importance of that isolation.

Furthermore, even though the failure of the TIP systems lines presents no safety consideration, the TIP system has redundant isolation capabilities.

Equivalent safety features have been reviewed by the NRC for BWR/4 (Duane Arnold), BWR/5 (Nine Mile Point) and BWR/6 (CESSAR), and it was concluded that the design of the containment isolation system meets the objectives and intent of the General Design Criteria.

Isolation is accomplished by a seismically qualified solenoid-operated ball valve, which is normally closed.

To ensure isolation capability, an explosive shear valve is installed in each line.

Upon receipt of a signal (manually initiated by the operator), this explosive valve shears the TIP cable and seals the guide tube.

When the TIP system cable is inserted, the ball valve of the selected tube ~ opens automatically so that the probe and cable may advance.

A maximum of four valves may be opened at any one time to conduct calibration, and any one guide tube is used, at most, a few hours per year.

If closure of the line is required during calibration, a signal causes the caole to be retracted and the ball valve to close automatically after completion of cable withdrawal.

If

.a TIP cable fails to withdraw or a ball valve falls to close, the explosive shear valve is actuated.

The ball valve position is indicated in the control room.

As stated above, the penetration is normally closed (open an average of 15 hours1.736111e-4 days <br />0.00417 hours <br />2.480159e-5 weeks <br />5.7075e-6 months <br /> per month), and the design leak rate-is cm3 sec.

If a failure occurred, such as not being 10-2

/

able to withdraw the TIP cable, the shear valve would isolate the penetration, and the resulting maximum leakage would be 10-3 cm3 sec.

The shear valves are shop tested by

/

statistical sampling methods to ensure that the leakage limits conform to the design specification limits of 10-3 cm>/sec.

Testing of the ball valve is not recommended, since a very small amount of leakage is expected, and any testing would need to be performed from inside the drywell, exposing the operator to radiation dose estimate to be about 50 mR.

These lines should therefore be exempted from Type C tests of 10 CFR 50 Appendix J.

April, 1980 LSCS - NTOL ACTION PLAN 2.1.5a Dedicated Penetrations for External Recombiner LaSalle County Station has two permanently installed post-LOCA combustible gas recombiners each taking suction and discharging through deditated penetrations and safety grade piping and valves.

The containment purge system also uses dedicated penetrations and safety-grade piping and valves.

All penetrations and piping are sized to meet system flow requirements.

In addition, both systems meet the redundancy and single-failure requirements of General Design Criteria 54 and 56 of Appendix A to 10CFR50.

Since the LaSalle County Station design already meets the requirements of Item 2.1.Sa, no design changes are necessary.

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i April, 1980 LSCS - NTOL i

ACTION PLAN l

2.1.5c Capability to Install Hydrogen Recombiners at LaSalle

.The LaSalle County Station hydrogen recombiners are permanently installed and are capable of. remote operation from the control room.

Therefore, no additional action 11s required to satisfy Item 2.1.5c.

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April, 1980 LSCS - NTOL ACTION PLAN 2.1.6a Inteority of Radioactivity Barriers in Systems outside Containment Commonwealth Edison has one difference with the recommendations of the November 9, 1979, NRC letter.

This concerns the use of an integrated leakage test on systems to determine leakage.

It is Commonwealth Edison's position that an integrated leak rate test is not effective for the specific concern of out-leakage and is not practical, given LaSalle plant design considerations.

In addition, integrated leak rate testing does not lend itself to an on-going maintenance program which could have the greatest benefit in a leak reduction program.

As an alterrative, Commonwealth Edison proposes the following program:

1.

Liquid systems will be visually inspected for leakage while systems are at approximate operating pressures.

Gas systems will be evaluated using helium leak tests, pressure decay tests for specific tanks and metered make-up pressure tests.

2.

An aggressive maintenance program will be used to assign high priorities to leakage related work requests.

Essentially all leakage on concerned systems will be covered.

3.

Systems lists will be available for review detailing specific methods used to test systems, the systems involved, frequency of testing and individuals responsible for testing.

4.

The LaSalle Technical Staff will review leakage-related work requests to evaluate possible modifications to keep leakage "as low as practical".

5.

An annual report will be prepared for LaSalle Station and submitted to the Nuclear Stations Division Manager.

This report will include past-year performance, current leakage rates, and status of leakage work requests and modifications.

The advantage of this program is that specific components l

requiring maintenance are identified during the actual surveillance.

This facilitates maintenance action.

Visual surveillances are supplemented by walkdowns during valve alignment procedures or other Inspections.

Leaks identified during these frequent inspections can be repaired rapidly thus facilitating the leak reduction effort.

In addition to the above, a water inventory program will be developed at LaSalle to allow trend-analysis cf leakage by monitoring sump levels, pump run times and tank inventories.

In this manner, knowledge of current leakage rates is available and performance of the overall program may be evaluated.

The following table identifies the systems which will be monitored, the monitoring methods, and inspection frequencies.

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SYSTEM & LEAK TEST METHODS IDENTIFICATION System / Component Method Used Frequency

1. Low Pressure Visual (1)

During operability Core Spray (LPCS)

Detailed inspection surveillanccs leakage measurement Once per refueling cycle 2.

High Pressure Visual (1)

During operability Core Spray (HPCS)

Detailed inspection surveillances and leakage require-Once per refueling ment cycle

3. Residual Heat Visual (1)

During operability Removal (RHR)

Detailed inspection surveillances and leakage require-Once per refueling ment cycle 4.

Rec tor Core Visual (1)

During operability Isolation Cooling Detailed inspection surveillances (RCIC) and leakage require-Once per refueling ment cycle

5. Hydrogen Recombiners Visual inspection During Primary (HG)

(soap bubble)

Containment Leak Rate Testing

6. Reactor Water Leakage inspection Once per refueling Cleanup (RT) and measurement Cycle

7. Coolant Sampling Leakage inspection Once per refueling and measurement cycle when samples Visual (1) are taken

8. Reactor Building Level Recorder Daily Equipment Drain Tank 9.

Instrumentation Leakage inspection Once per refueling Lines and measurement cycle Visual Routine

10. Standby Gas DOP/ Freon Testing Once per refueling Treatment (UG) cycle 5

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0 April, 1980 LSOS - NTOL ACTION PLAN 2.1.6b Desinn Review of Plant Shieldino and Environmental Qualification of Equipment for Systems Wnich May Be Usea in Post-Accident Operations A radiation and shielding design review was made for LaSalle using the NRC-prescribed post-accident distribution of radioactivity.

For a BWR, this distribution carries considerable radioactivity (concentrations) throughout the plant via piping which contains reactor water and by airb"rne particulate concentrations in various spaces.

However, based upon the fact that no operator actions other than those which take plact in the control room or at the remote shutdown panel are critI7a1 for plant shutdown, only these areas and the sampling stations and Technical Support Center (TSC) are considered to be vital for personnel access fcr post accident cases.

Dose rate maps were drawn to indicate the spacial-and time-dependent characteristics of the radiation emanating from the two NRC accident-release prescriptions.

In general, those preliminary maps show that the control room, the Auxiliary Electric Equipment Room, where the Remote Shutdown Panels are located, and the Technical Support Center have dose rates which allow continuous occupancy given the accident scenerios.

The preliminary results also show that accessibility to the control room is not a problem during such accidents.

Application of the GDC 19 accident limit of 5 Rem whole body (or equivalent) for areas requiring infrequent access indicates that adequate occupancy times are available for typical operator actions.

The significant radiological conclusion is that the "less than 15 mr/hr" criteria is met at LaSalle for plant areas requiring continuous occupancy.

See the table of preliminary results' on the next page.

The evaluation of environmental qualifications for essential equipment is also a part of the on-going LaSalle design assessment.

IEEE 323-1971 is the current licensing basis for LaSalle, however, the on-going assessment is being pursued as described in paragraph 6 of IEEE 323-1974.

New post-accident instrumentation to monitor the 3 ACRS-identified containment parameters and containment sampling and reactor water sampling are being designed to IEEE 324-1974 standards including radiation phenomena to the extent practicable. L.

Post Accident Dose-Rates for LaSalle 1 & 2 Dose rate in R/hr at the listed locations for the NRC-prescribed accident cases (TM1 2.1.6b).

Type of Control Remote Technical Sample Diesel Pia Accident Room Shutdown Support Station Gen.

Prct:

Panel Center Rooms cti Fen Unit 1 Unit 2 1 hr

< 0.001 4 0.001

<0.001 0.015-0.1 1.0-10 0.1-1 day

< 0.001

< 0.001

<0.001

< 0.015

( 0.015 0..'. -

1 week

( 0.001

< 0.001

< 0. 001 Shutdown 0.1-1 0.01 level 0..

G April, 1980 LSCS - NTOL ACTION PLAN 2.1.7a Auto Initiation of the Auxiliary Feeflyter Systern (AFWS)

N/A to BWR.

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April, 1980 LSCS - NTOL ACTION PLAN 2.1.7b Auxiliary Feedwater Flow Indication to Steam Generators N/A to BWR, April, 1980 LSCS - NTOL ACTION PLAN 2.1.8a Improved Post Accident Samplino Capability Working models of a reactor coolant liquid-sample system and a containment air-sample system for post accident samples are under evaluation at this time.

1.

Post-Accident Reactor Water Sampling The present liquid sample system at LaSalle takes water from the discharge side of the recirculation pump in the B recirculation loop.

It is adequate for routine, non-accident sampling.

The liquid sample system discussed here is for the post accident situation.

It is an entirely new system to be installed at elevation 687 feet (upper basement level) of the Auxiliary Building which has shi'elded access independent from the reactor building proper.

Adequate space and geometry for significant shielding are also features of this location.

The initial plan for this sampling capability includes the following:

- Draw a 15 milliliter sample of raw reactor water

- Dilute to a 10-3 aliquot

- Run a dissolved-gas sample for isotopic gammas

- Run a dissolved hydrogen determination

- The system will allow sample collection and analysis within the exposure guidelines of NUREG-0578 The containment air sampler will utilize automatic grab-samples at the time of the accident or as immediately thereafter as practical, at 20 minute post accident, and at 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> later.

The l'aproved chemical analysis of water will include a capability to analyze for boron using a 10-3 diluted sample.

The analysis technique employs a spectrophotometer to read color.

Chloride Analysis is also provided via lon chromatography.

Dissolved oxygen in the reactor coolant will be determined by use of an on-line dissolved oxygen meter.

The solution ph and conductivity readouts are available from miniaturized in-line equipment similar to plant process equjpment.

The radiation analysis will be performed on the 10-aliquot.

=

2.

Post-Accident Primary Containment Atmosnhere and Gross Gamma Monitorino The purpose of the containment atmosphere and gross gamma monitoring systen is to provide the signals necessary to indicate and alarm high hydrogen concentration or high gross-gamma radiation in the drywell following a loss-of-coolant accident (LOCA).

The containment atmosphere monitoring subsystem monitors hydrogen concentration in the drywell resulting from radiolytic and chemical phenomena associated with accident condition.

The gross gamma monitoring subsystem monitors gamma radiation resulting from the gross release of fission products from the fuel.

Each subsystem has two redundant channels of instrumentation which are physically separated and electrically independent.

Separate electric sources (120 Vac) on separate divisional buses power each channel.

Each channel provides a local measurement and transmits the signa;. to the control room where a permanent record is made on seismically qualified recorders.

This system is designed in accordance with Seismic Category I requirements.

The piping for this subsystem up to and including the outboard isolation valves is designed in accordance with ASME III (1974) Class 2 requirements.

The hydrogen monitoring subsystem is automatically activated on the occurrence of a LOCA and it remains in operation after initiation until turned off with a handswitch.

The hydrogen concentration is recorded up to 10 percent (volumetric) with an accdracy of + 5% of the readout.

An alarm is activated.on high concentration.

April, 1980 LSCS - NTOL ACTION PLAN 2.1.8b Increased Ranoe of Radiation Monitors 1.

Radioloolcal Noble Gas Effluent Monitorino.

The General Atomic wide-range monitor will be installed in the ef fluent stream widch enters the LaSalle Station vent stack.

A separate monitor will be installed for the Standby Gas vent stack which is wholly contained inside the station vent stack.

This monitor has a range for radioactive gas concentration of 1 X 10-7 uCi/cc to 1 X 10+5 uCi/cc.

The monitor is designed to meet Class IE requirements and is in the process of being qualified to IEEE 323-74.

The monitor meets Table 2.1.8.b.2 of NUREG 0578.

The energy dependence will be determined during calibration.

The monitor, therefore, requires only one level of radioactive gas for each detector.

Kr-85 and Xe-133 at concentrations of 10-4 uCi/cc, 10 uCi/cc, and 1000 uCi/cc, will be injected into the monitor for calibration purposes.

Then each decade response will be verified using a set of Cs 137 sources.

At the time of purchase and/or af t er the replacement of any detector, an energy response curve will be~run using at least five solid sources of different gamma energy levels.

The calibration will be conducted at least once every quarter for the first year of operation and then once every six months.

The method for converting instrument readings to release rates will be determined after the energy responses of the detector are obtained from actual tests, even then the monitor response can only give a very rough estimate of the release.

Actual releases will be determined by using En automatic grab sampler, counting the samples collected, and calculating the release.

Studies will be conducted relating the ventilation air flows, monitor reading, monitor energy response, and time after shut down to improve estimates of release.

Monitor readouts and other technical information will be provided in the Technical Support Center and the control room.

2.

Radiciodine and Particulate Effluent Monitoring The sampling media will be analyzed in the counting room at LaSalle.

Charcoal cartridges will be reverse-blown with air to purge interferring noble gases.

The eetectors at LaSalle are currently Ge(L1) crystals.

In addition silver-zeolite cartridges will be used to further reduce noble gas interference.

.=

The monitoring and sampling locations are the same as those used for noble gas detectors, above.

Sample retrieval procedures are to be developed by LaSalle health physicists in consultation with the equipment vendor.

An alternate power supply for the counting room will be provided from an essential power bus.

3.

Hioh-Ranoe Containment Radiation Monitors Two higa range containment radiation monitors will be installed on each of LaSalle's units.

The monitors will be mounted in steel sleeve: which protrude into the containment this installation will result in attenuation of the 67 Kev photons of Xel33 CECO. fe?ls that this system is adequate to monitor containment radihtior when combined with the grab sample system which we are installing in response to item 2.1.8a.

The monitors will provide on scale containment radiation reading during an accidentandgrabsampleanalysiswillgrovidedetailedisotopic information including to presents of Xe 33

.. y; The General Atomic monitors to be installed were designed ;3 meet all the requirements of NUREG-0578.

Following is a list of the specifications:

1.

Radiation Lifetime:

10"R(integrated dose) for detector 2.

Range:

100 to 108 R/HR 3.

Self-Test:

continuous detector signal corresponding to 1 R/HR from radiation source inside detector 4.

Accuracy:

+3% of equivalent linear full-scale output 5.

Electronics Temperature Coeffficient:

0.1% of full scale per DC J

6.

Checkcurrent:

electronics calibration via internal 5 R/HR.

current source corresponding to 10 7.

Time Constant (RC):

25 MSEC 8.

Maximum Temperature:

3500F detector, 1300F electronics-9.

Maximum Pressure:

65 psi detector-10.

Seismic:

tested per IEEE 344-1975 using random biaxial inputs i

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April,-1980 i

i LSCS - NTOL ACTION PLAN j

2.1.8.c.

Improved In-Plant Iodine Instrumentetion-LaSalle County Station will purchase a portable SAM-2 radiolodine measurement system from Eberline Instrument Corporation.

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April, 1980 LSCS - NTOL ACTION PLAN 2.1.9 Analysis of Deslan and Of f-Normal Transients and Accidents A generic analysis of the following transients and accidents for the BWR-5 wa, included in NEDO-24708, GE's response to the Bulletins and Orders Task Force on TMI-2 Short Term Lessons Learned:

1.

Loss of Flow (LOF)

LOF plus SORV LOF plus SORV with RCIC only LOF plus SORV with Manual HPCS and RCIC LOP plus SORV with neither HPCS or RCIC 2.

Small Break LOCA (SLA)

Analyses were made for the following Break sizes:

0.0001 ft.2, 0.005 ft.2, 0.08 ft.2 These sizes cover the cases described below.

a.

Cases where core power generation exceeds the energy loss rate through the small break hence system pressurization occurs, b.

Cases where core power generation roughly matches the energy release rate through the break hence reactor coolant steady state essentially exists.

c.

Cases where the core power generation lags the energy loss rate through the break hence vessel depressurization occurs.

This August 1979 NE00 Report will be followed by'an additional set of generic analyses selected by NRC for other transients and accidents including the Inadequate Core Cooling' hypothesis.

This generic analysis activity and its accompanying definintion of emergency procedure and operator retraining will be done (for LaSalle and other BWR-5's) on a schedule consistent with those established by the Bulletin's and Others Task Force.

Revised procedures reflecting the results of these studies will be available prior to fuel load.

_ _ _. ~.

.o August, 1980 LSCS - NTOL ACTION PLAN 2.1.9 Containment Monitorino - Pressure Indication The ACRS recommendation to have a continuous indication of containment pressure available in the control room is met with presently installed Rosemount hardware which senses containment pressue via two measurement channels with two sensors in each channel.

The present wide-range pressure measurement covers 0-60 l

psig and the narrow-range pressure measurement covers -5 to +5 psig.

These will be changed to replicate wide-range instruments with the range -5 to 135 psig, which is three times the concrete containment design pressure of 45 psig.

The scaled ranges of the recorders in the control room will also be changed accordingly.

In addition to the drywell pressure instrumentation described above, there are two channels of pressure monitoring equipment for the air volume above the suppression pool.

Their purpose is to indicate the pressure of this air space as it may be affected by canoensation phenomena in the suppression pool or by suppression pool bypass leakage.

Each channel is powered from

,. redundant Class IE emergency instrument buses.

Each reads out on a seismically qualified indicator in the main control room.

The indicators are mounted on the front of the panel for operator visibility.

The LaSalle pressure transmitters are qualified to IEEE 323-1971.

Components which satisfy all the requirements of IEEE 323-1974 have not been lccatad.

It is judged, therefore, that the LaSalle equipment represents the current state-of-the art.

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April, 1980 LSCS - NTOL ACTION PLAN 2.1.9 Containment Mcnitoring - Water Level Indication The present LaSalle suppression pool water-level measurement system vill be modified to obtain the capabilitv to measure water-level over a 38-foot range from the vacu_.a breaker valve return line connection (25 feet above normal level) down to one foot below the suppression pool downcomers (approximately 13 feet below normal level).

This range is consistent with the range suggested in proposed Revision 2 to Regulatory Guide 1.97 (Top of vent to top of weir well).

The top of weir wall of a Mark III containment being equivalent to the bottom of the suppression pool downcomers of aMark II containment.

The presently installed differential p ressure transmitters will be replaced by Rosemount hardware of the same type to cover the new range.

The scale and chart range of the control room recorders will also be changed.

Continuous indication and recording in the control room is met with the seismically qualified Class IC recording equipment.

The LaSalle pressure transmitters are qualified to IEEE 323-1971.

Components which satisfy all the requirements of IEEE 323-1974 have not been located.

It is jedged, therefore, that the LaSalle equipment represents the current 5 tate-of-the-art.

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April, 1980 LSCS - NTOL ACTION PLAN 2.1.9 Containment Monitorina - Hydrocen Indication The ACRS recommendation to have a continuous indication of containment hydrogen concentration available in the control room is met with redundant Delphi monitoring units.

Their capability covers the range of 0 to 10 percent hydrogen concentration by volume over a pressure regime of minus 2 psig (12.7 psia) to plus 60 psig.

These monitors are currently undergoing IEEE-323-1974 qualification testing on a type basis.

That program is not expected to be completed until June 1980.

LaSalle currently meets the requirements for continuous indication in the control room via seismically qualified Westronics pen recorders which have an accuracy of +10% of span as required in the draft ANS 4.5.

This equipment is testable on-line from the control room.

(See the second part of 2.1.8.a above)

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April, 1980 LSCS - NTOL ACTION PLAN 2.1.9 Reactor Coolant System Vent The LaSalle EWR's are provided with eighteen power-operated relief valves which can be manually actuated from the control room to depressurize (vent) the reactor pressure vessel.

These vent lines are piped to the suppression pool where water submergence of the quencher discharge device assures condensation and pressure relief.

Water quenching also plates out solubles from the reactor steam and localizes fission products in the suppression pool.

The relief-valve mode of operation of the dual function Crosby SRV's satisfies the need for vessel venting because the vessel vent is the mainsteam lines themselves.

They exist at the top of the vessel cylindrical section.

Accumulation of gases above the steamline exit elevation of the RPV will not af fect. venting of the four 26-inch steamlines.

Three of the eighteen relief valves can be manually actuated from the remote shutdown panel outside the control room.

1 Seven of the eighteen relief valves are integrated into the Automatic Depressurization System (ADS) which automatically depressurizes the reactor vessel via the piped blowdown pathway.to the suppression pool.

The main steam piping, the SRV's, the folw restrictors and supports from the RPV out to and including the outboard isolation valve are all Class I seismic qualified.

The SRV's also meet ASME Section III Code requirements. ADS logic is redundant and safety grade equipment.

The BWR Owner's position is that imposition of the single-failure criteria to prevent inadvertent actuation of these valves, and the requirement that power be removed from these valves during normal operation are not applicaale to BWR's.

These valves serve an important function in mitigating transient effects and in ASME code overpressure protection.

Therefore, the addition of a second " block" valve on the vent lines would actually decrease the safety margin and violate the ASME code in some cases.

Also, inadvertent opening of a relief valve in a BWR is a design basis event with a controllable transient of no safety significance.

In addition to these power-operated relief valves, the LaSalle units include other means for high-point venting.

LaSalle reactors use a normally open reactor head-vent line which discharges to mainsteam line A.

This vents the volume inside the RPV head that is above the cylindrical part where the mainsteam nozzles attach.

Additionally, a main-steam-driven RCIC system accepts RPV steam (gases), expands them through a turbine, and exhausts the condensate

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Although the controllable relief valves fully satisfy the intent of this TMI requirement, the alternative methods described herein also provide protection against the accumulation of non-condensible gases in the GPV.

r The result of a break in the SRV discharge pipe or any other systems pipe enumerated above (all inside containment) would be the same as a small steam line break.

The SBA is a part of the LaSalle design basis and has been shown to be of lesser severity than_the Chapter 15 accident cases reported.

A number of reactor system blowdowns due to SORV's have confirmed this in practice.

Thus, no addit.onal analyses are required to show conformance to 10 CFR 50.46.

Likewise, based upon the fact that the SRV's, the head vent, and the RCIC system will vent the reactor pressure vessel continuously when needed, and based upon the fact that continuous venting 13 a part of the containment hydrogen calculations for normal safety analyses, no special analyses are required to-demonstrate that the direct venting of non-condensible gases with I

(perhaps) high hydrogen concentrations does not result in violation r

of combustible gas concentration limits in containment.

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April, 1980 LSCS - NTOL ACTION PLAN 2.2.1.a Shift Supervisor Responsibilities The directives and procedures necessary to meet the requirements of NUREG-0578 and clarification contained in the November 9, 1979 V.

B. Vassalo, letter will be prepared and implemented prior to fuel load.

The directives and procedures will be similar to those outlined in the January 1, 1980 D. L. Peoples letter to H. R. Denton with applicable _ references.

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April, 1980 LSCS - NTOL ACTION PLAN 6

2.2.1.b.

Shift Technical Adviser The Shift' Technical Advisor will be on shift prior to fuel load in a manner similar to that outlined in the January 1, 1980 D.

L.

Peoples letter to H.

R. Denton.

Long Term Aoproach LaSalle's ultimate goal is to provide on each shif t, a

technical graduate licensed at the Senior Reactor Operator (SRO) level.

This individual will have the training necessary to perform the accident assessment function and will be in excess of the minimum shift SRO requirement identified in the August, 1979 BWR Standard Technical Specifications.

This position is expected to be filled no later than June, 1981.

Fuel Load Approach 1.

If an excess of supervisors on shift, i.e. more than are requjred by the August, 1979 3WR Standard Technical Specification, obtain a license at the SRO level, the accident assessment function will be fulfilled either by one of these shift SRO's who has in addition completed an augmented training program equivalent to'that outlined in the C. Reed letter to H.

R. Denton dated November 30, 1979 or by one of these shift SR0's who is a technical graduate.

2.

If an excess of Supervisors licensed at the SRO level are not available, one SRO Supervisor on each shift will receive the augmented training program discussed in part I and the accident assessment function will be provided by a Technical graduate assigned to each shift.

Either Fuel Load Approach will provide an individual on 4

shift who is able to report to the control room within 10 minutes to advise the Shift Supervisor during an accident. T 1

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- August, 1980 4

LSCS - NTOL ACTION PLAN 2.2.1.c.

Shift and Relief Turnover Procedures The directives and procedures necessary to meet the j-requirements of NUREG-0578 and the clarification contained in the November 9, 1979 V.

B. Vassalo letter will be prepared and Implemented prior to fuel load.

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April, 1980 LSCS - NTOL ACTION PLAN 2.2.2.a.

Contrcl Room Access The directives and procedures necessary to meet the tcquirements of NUREG-0578 and the clarification contained in the November 9, 1979 V.

B. Vassalc letter will be prepared and ir:plemented prior to fuel load.

April, 1980 LSCS - NTOL ACTION PLAN 2.2.2.b Onsite Technical Support Center Commonwealth Edison will establish an on-site technical support center at LaSalle prior to fuel loacing.

Communications with the control room and the NRC will be completed by this time-frame.

Communications with the near-site emergency operations center will be established on a time-frame consistent with the requirements of the September 17, 1979 letter for OL plants.

Procedures will be written to cover the accident assessment function in the Technical Support Center (TSC) and the control room.

Procedures for prevention or reduction of radiation exposure to personnel will be revised or written as required.

The direct display of plant parameters in the TSC may not be possible,

.given the short time interval between now and the end of the year.

However, LaSalle response procedures and direct communications between knowledgeable individuals in both cont rol room and TSC will ensure reliable and timely transmittal of plant information to the TSC.

By January 1, 1981, within the limits of equipment availability and scope of construction, the TSC will be upgraded to meet the recommendations of the Atomic Industrial Forum Subcommittee on Emergency Response Planning.

Design criteria and a conceptual design description of the upgraded onsite technical center is attached.

April, 1980.

LSCS - NTOL

-ACTION PLAN E

2.2.2.c-Onsite Ooerational Support Center An Onsite-Operational.Suppori. Center will be established at the LaSalle' Station.

Communication will be provided between the Operational Support Center and the Control Room.

Procedures will be i

prepared and implemented reflecting the existence of the Center and i

establishing the method and lines o' communication of management.

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