ML20039E776

To Suppression Pool Leakage Return. Response to NUREG-0737,Item II.B.3 Re post-accident Sampling Encl

ML20039E776
Person / Time
Site:	Shoreham File:Long Island Lighting Company icon.png
Issue date:	12/30/1981
From:	LONG ISLAND LIGHTING CO.
To:
Shared Package
ML20039E768	List: ML20039E767 ML20039E776
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.3, TASK-TM SP-23.702.04, NUDOCS 8201110399
Download: ML20039E776 (16)

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SP Number 23.702.f,4 Revision 0 Date Eff.'12/307iil TPC TPC TPC SUPIRESSION POOL LEAIM E PETURN O

1.0 PURPOSE To provide detailed instructions to the Sectice Operating Parednnel for'thr.

operation of the Suppression Pool Leakage Return Systca.

2.0 RES PO.';S iBILITY

• ihe Operating Engineer chall b[ responsible for ensuring the proper inplete:ttation of this procedare.

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• 3'.1 The Suppression Pool Ledkage Retufn Systen Returns Leakage from Energency Core Cooling Systen (ECCS) passive failures (pump seal, instrument line, or "W

valve"p':idkin'g failufc7 'in"t' lid *R'c2ct'or 'Buhding' bdcli

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pool.

3.2 The Suppression Pool Leakage Return Pump (1Gil*P-2700) takes suction from Reactor Building floor drain sump IGil-TK-056C and discharges through a notor operated valve (lGil*MOV-639C) to the suppression pool via the core spray test.line.

3.3 The Suppression Pool Leakage Return Systen is designed for a maximum leakage of 100 gpn (Core Spray Pucp Discharge Pressure Instrument Line Break) and is manually initiated as required to retura post-LOCA ECCS leakage back to the suppression pool.

NOTE:

1.

All equipmant cnd component identification numbers arc preceded by the Systen Number IG11 unless specified otherwise.

2.

All control switches for recote operated valves and pumps are located r.n the rna cor. trol roca panel 1111l AP H.--PC:t unless specified otherwise.

3.4 The following procedures are provided for the operation of the suppression pool leakage return systen:

8.1 Normal Perf ormance Page 8.1.1 Standby Status 3

8.1.2 Systen Initiation 3

8.2 Abnornal Performance 8.2.1 Loss of offsite power 4

Appendix 12.1 Prerequisite Checklist Appendix 12.2 Valve Lineup Chechlist Append i:: 12.3 Systen Component Porer Supplies Checklint 4.0 PRECAUTIO::S 4.1 Before initial startup of sump pumps, inspect sumps for oil and debric accuuulatious.

After initial startup, sumps shculd be periodically inspected.

4.2 Observ2 all radiological precautions when operating this synten.

5.0 PitEREOUISITES 5.1 Prerequisites are delineated on the PREREQUISITE CHECKLIST, SPF23.702.04-1.

SP 23.702.04 Rev. 0 12/30/81 Page 2

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6.'O,, LIMITATIONS AND ACTIONS t.

,, '... ~6'.1... Technical Sp6cifications 6.1.1 During Power Operation, Startup and Hot Standby (CONDITIONS 1, 2, and 3) the Limiting Conditions for Operation of References 11.1 and 11.2 shall apply.

7.0 MATERIALS OR TEST EQUIPMENT N/A 8.0 PROCEDURE 8.1 Nornal Perfor.mance 8.1.1 Standby Status S.1.1.1 The Suppression Pool Leakage Return Systen wi)) not norrally be operating, but will be maintained in a standby condition.

The following steps place the Suppression Pool Ler:w s Return Sycten in standby readiness:

.1 Co.mplete Prcrequisite Checklist, SPF23.772.04-1

.2 The Systen is ncw in standby 8.1.2 System Initiation 8.1.2.1 The Suppression Pool Leakage Return System must be manually initiated to return leakage fron ECCS systems in the Reactor Building to the Suppression Pool.

The following steps place the Suppression Pool Leakage Return Systen in operation:

CAUTION:

To ensure all leakage is returned to the suppression pool, the Reactor Building Floor Drain and Equipment Drain Sump Pumpc should be secured.

.1 At the Prinary Contain cut Monitoring Panel (lHil*PNL-PCM), OPEN Suppression Pool Leakage Return containr7nt isolation valve IC11*MOV-639C.

NOTE:

If a contvincent isolation signal due to high drywell pressure or low reactor uater level in present, place the override switch in override prior to opening the valve.

.2 Place the -Suppression Pool Leakage Return pump control switch in START and observe normal flow SP 23.702.04 Rev. 0 12/30/31 Page 3 u

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(l'00-150 gen) on ICll-FI-647C.

'yyyesystAmIs'nowEnope. ration and will-return

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,3 leakage collected in floor drain sunp IG11-TK-056C to the Suppression Pool.

NOTE:

If the. floor drcin sump level decreases to less than 36 inches (1c11-LE-642C), the pump will stop automatically but nust be manually restarted when level increases.

8.2 Abnornal Performance 8.2.1 Loss of Offsite Power 8.2.1.1 The Suppression Pool Leakage Return Pump and Containment Isolation MOV nre powered from an emergency bus and will e

remain operable durir.g a less of r.ormal AC po.:er.

9.0

,1CCEPTANCE CP.ITEPtIA N/A 10.0 FI:A1. CONDITIONS 10.1 Insure that all checklists have be5n ec picted and cisned.

11.0 REFERENCES

11.1 Technical Specifications, Section 3/4.4.3 11.2 Technical Specifications, Section 3/4.6.3 11.3 SP 23.404.01, Area I.cakage Detection

,11.4 SP 23.702.01, Equipment Drainage and Floor Drains 11.5 FM-46B-8, (S&W) Radwaste Equip. & Floor Drains - Rx Bldq, M-10149-8 11.6 ESK-6Cll68, Rev. 2, Eler.entary Diagram Succ. Pool Pro Back Is ol VV 11.7 ESK-601169, Rev. 2, Elementarv Diagram Suop, oel Pnp Pach Pmp n

11.8 LSK-31-6.2, Issue 4, Reactor Euilding Sur.p Pumps 12.0 APPENDICES 12.1 SPF 23.702.04-1, Prerequisite Checklist 12.1 SPF 23.702.04-4, Valve Lineup Checklist 12.3 SPF 23.702.04-5, System Co.nponent Power Supplies Checklist SP 23.702.04 Rev. 0 12/30/81 Page 4

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Appendix 12.1 EYSTE!!

SUPPRESSIO:4 POOL LEAKAGE RETUIC:

. PftBRPQUIS*iTE.CitEGibbIEE -

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Signature Initinic h

Date Autharization for Start (Watch Engineer)

Initiated by Completed by Reviewed by (Watch Engineer)

S cp No.

Procedure Initials 5.1 Taggint; Log and lif ted Lead & Jumper Log Revicued.

5.2 Sys t em Va h, Lineup Chacilint SPF23.702. M -4 complete.

5.3 sump inspected for oil and debris accunulations.

NOTE:

Prior to any breaker oparations, place the control switch in STOP, CLOSE, or PUIL-TO-LOCK position.

5.4 Equipment is energized per Synten Component Pouer Supply Checklist SPF23.702.04-5.

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SPF 23.702.04-1 SP 23.702.04 Rev. O 12/30/81 Page 3 I

Anaendix 12.2 I'Arge 1 of 1

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. V/J.VE LINE-UP

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'SI'P'NIISSS'IGITOOL Li'AR.'.GE ' nS'UI'i Si$ YEW

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Ah* A second qualified operator should verify proper aliganent Y1LVE REQUIRED INITIALS

'; UMBER DESCRIPTION POSITION REACTOR BUILDING EL. 8' - 0" 1G11*

LEAKAGE RETURN PU:i? "P-270 LOCKED 33V-2103 DISCilAPGE ISOLATION -

OPEN IG11*

OlV-9334 FE-647C I::LET ROOT VAL'lE OPEN Wi l*

J1V-9535 FE-647C OUTLET ROOT VALVE OPEN IG1If J1V-9536 PI-64C ROOT VALVE OPEN

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. CI.OSE D LG11*

l ICTITJ3V-211WC LOC (.ED

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LEAKAGE TEST CO:WECTION CLOSED 1G 1 O ;i3V-211 L C LOCKED

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i Olv-3536 l

LEAL'#JE TEST CONNECTIO::

CLOSED IGilk i

1G11*03V-211dC LOCKED OlV-3537 l

LEAEAGE TEST CO::NCCTT03 Cl.OSED IGl1*

IG11*p3V-21IfC LOCKED JlV-3538 l

LEAKAGE TEST CON: ECTION CLOSED 1E21*

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SUPPRESSION POOL LEAKAGE RETURN ISOL.

OPEN 9

SPF 23.702.04-4 SP 23.702.04 Rev. 0 12/30/81 Page 6

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1,aendix 12.3 ca r.c 1c. 1 SYSTEM CC:iPONENT PO;'ER SUPPLIE9 SUPP:U:SS ION l' DOL LSAKAG: IETURN SYS Ti'.M

• • • A second cualified onerntor should verify proper niinnment iCO:.UO:;iN "

Po'J c:R S'Ti>LY/

REQUIRED i INITIAL

!* :!3ER CCMPC:'E::T DESCRIPTION PE'? tP.ER " !M3ER POSITION 1011*MOV-6390 SL'PPRESSIO:! POCL LEAUCE RETURN ISOLATIO:! MOV IP.24 *MCC-1 I li'/1 AD ON IC11*P-27CC SUPPRESSIO:: PCOL LEAP. ACE P.CTUPL PU:!P 1 ". 2 ': !!C C-I l li.'h. AC ON i

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y, II.B.3 Post-Accident-Sampling, Staff Position 2 Atmospheric Sample Mixing in the Suppression Chamber Sample lines-from both of the suppression chamber access hatches are tied into existing. sample lines for the gas analyzers for the primary containment atmospheric control system.

These tap points.were selected because mixing studies had insured these hatches had ample mixing for representative sampling.

To ensure that the* post accident sample would be representative as well, the noble gas mixing has been confirmed by calculation.

The study proceeded as.follows:

There are two possible means of-sample mixing in the suppression chamber _ atmosphere:

p 1.

' steady state, where the gases diffuse via Brownian motion and 2.

motion through induced velocity caused by system turbulence.

I As pure diffusion is. a very slow process and not relevant under accident conditions, the second approach was applied..The

-justification for this approach is that water motion, due to-safety relief 'ralves lifting and steam condensation, causes.

4 differential pressure and (some amount of) atmospheric turbulence.

i The noble gases transported by bubbles, are-released into the suppression chamber atmosphere which, because of-the turbulence, imparts an initial velocity to the sample medium.

Equilibrium sample mixing is reached within 10 minutes.

Therefore, it is concluded that adequate mixing will occur and the samples extracted are representative of the mixture.

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_p Item II.B'.3' --Post-Accident Sampling dustification of Jet Pump Discharae Sampling Location The-following provides justification for the jet pump discharge sampling locations for the post-accident sampling system'(PASS) during small break or non-break accidents:

In order.to assure that the jet _ pump discharge provides a repre-sentative sample, two conditions should exist:

A. -Enough core flow to allow circulation _of water from inside the shroud'to the jet pump intake.

B.

No significant dilution of makeup water.

Two assumptions were made for this study:

1.

Reactor water -level can be maintained at or near normal water level after the accident.

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2.

Reactor power level is greater than 1% rated, up to approximately 10% rated, when the water sample-is being taken.

Regarding condition A, after a small break or non-break accident, the reactor water level will be maintained at or near normal water level by the operator using Emergency Procedures.

For decay power-above 1% of rated power, the core flow is estimated to be greater than 10% rated recirculation flow due to natural circulation.

This amount of core flow assures the existence of a flow route from the t

core to the sampling points; it takes about 3 to 4 minutes-to circu--

late the entire reactor water inventory through the jet pumps.

Therefore, a representative sample of the cc.re water will be avail-able at the jet pumps.

Regarding condition B, for small steam line breaks or non-break accidents, makeup water is pumped in to remove decay heat and to make up for steam loss through the break.

This makeup water _ amounts to approximately 2% of the core flow present.

Even for small liquid line breaks, the makeup water flow rate is estimated to be less than 18% of-the core flow present.

Therefore, it can be concluded that no significant dilution would occur; the bulk of the water going through the jet pump comes from the reactor core.

In conclusion, the jet pump discharge can supply a representative i

sample of the reactor core water for the PASS under conditions of small break or non-break accidents.

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II.F.1

' ATTACHMENT-1 NOBLE GAS' EFFLUENT MONITOR

'NRC Positions

Noble gas effluent monitors;shall be installed with an extended range

~ designed to function during accident conditions as:well as during

' normal operating-conditions.

Multiple monitors are considered necessary tol cover the ranges of interest.

1.

Noble gas; effluent monitors with an upper; range capacity of 105 /( Ci/cc (Xc-133) are considered to be practical

. and should lx3 installed in all operating plants.

2.-

Noble gas effluent monitoring shall'be provided for the total range of concentration extending from normal condi-tion (a s low as reasonably achievable ( ALARA) ) concentra-tions to a maximum of 103 q Ci/cc (Xe-133).

Multiple monitors are considered to be necessary to cover the ranges of interest.

The range capacity of individual monitors should overlap by a factor of ten.

Licensecs shall provide continuous monitoring of high-level, post-accident releases of radioactive noble gases from the plant.

Gaseous effluent monitors shall meet the requirements specified in the attached Table II.P.1-1.

Typical plant effluent pathways to be monitored are also given in the table.

The monitors shall be capable of functioning both during and follow-ing an accident.

System designs shall accommodate a design-basis release and then be capable of following decreasing concentrations of noble gases..

Of fline monitors are not required for the PWR secondary side main steam safety. valve and dump valve discharge lines.

For this applica-tion, externally mounted monitors viewing the main steam line upstream of the valves are acceptable with procedures to correct for the low energy gammas the external monitors would not detect.

Isotopic iden-tification is not required.

Instrumentation ranges shall overlap to cover the entire range of effluents from normal (ALARA) through accident conditions.

The design description shall include the following information.

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System description, including:

a.

instrumentation to be used, including range or sensitivity, l

energy dependence or response, calibration frequency and technique, and vendor's model number, if applicable;

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b.

monitoring locations (or points of sampling), including des-l

~ cription of methods used to assure representative measurements and background. correction;

'II'F.1-2 1/5/82

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location o'f= instrument readout (s)'and method of recording,-

including description offthe method or procedure for trans-x-

=mitting or disseminating the information~or data;

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assuranceioffthe' capability-to obtain readings at least'

every 15' minutes during and following-an accident;' and 2

e.-

the' source of-power to be used.

'2.

Description of procedures or. calculational methods to be used for

converting instrument: readings to release. rates per unit time, based on exhaust air flow and considering radionucli6e spectrum-distribution as a function of time after. shutdown.

LILCO' Position'

-The effluent monitor categories in' Table II.F.1-1, which apply.to' the Shoreham Nuclear Power Station are: (a) "BWR reactor building exhaust air", - (b) ' "other release points", and (c) " buildings with systems-containing primary coolant or gases".=

See Fig. II.F.1-1 for a simplified diagram of Shoreham's gaseous effluent layout.

i The maximum anticipated primary containment leakage rate is 0.005 volumes per day (volume.of primary containment is 1.93x105 cu f t) into the secondary containment-which has a volume of 2x106 cu ft.

The primary containment leakage is highly diluted in-the secondary

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containment atmosphere.

This mixture will be discharged after.

passing through high efficiency particulate absolute filters and charcoal adsorber banks via the reactor building. standby' ventilation system (RBSVS) discharge pipe, at the top of the station vent exhaust.

Two Class lE radiation monitors (RE-021 and 022) serve this system downstream of the~ filters and adsorbers along with a post accident 4

+-

Class lE monitor (RE-134), which is added to the system for higher i

ranges.

i The RBSVS monitors are supplied with power from vital instrument huses.

These monitors read out in the control room and are located in the con--

trol building (RE-021 and 022) and in the turbine building (RE-134) to l'~~

permit access during an accident for collection of their radiciodine and particulate sample media for laboratory analysis.

The criteria in Table II.F.1-1 for other release points and buildings t

with systems containing primary coolant or gases are applicable to the station vent exhaust monitor. (RE-042) and the station vent post accident

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high range monitor (RE-126).

Normal ventilation discharges from the reactor building, the turbine building, and the radwaste building are mixed, thereby providing dilution prior to being exhausted through

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i the station vent exhaust.

When RBSVS is operating, and the reactor building normal ventilation system (RBNVS) is isolated, the loss of normal reactor building ventilation flow i s compensated by. opening r

L louvers at the station vent exhaust to permit 90,000 cu ft/ min of II.F.1-3 1/5/82

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SNPS-l'FSAR outside air--for dilution and to. maintain a constant air velocity through the~ station. vent.

This single discharge point.for the-combined; ventilation flow from all potentially contaminated buildings is monitored by a' noble gas radiation monitor. (RE-042) and~ post. accident high range effluent radiation monitor (RE-126).

The monitor -(RE-042) is : supplemented by in-line RE-069 with a high upper range.

In' addition, the individual building ventila-tion.flowsLto the station vent exhaust are each analyzed by a high' range in-line radiation monitor (RE-066, 067, and 068).

.All these' monitors, except RE-042', are powered-from a vital instru-ment bus, however, it is powered from a. dependable backup power supply to normal ac.

Where practical, initial calibration includes detector response for a minimum'of three decades using standard sourcesLof three

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different energies and -intensities.

These calibration curves are initially generated using both gaseous and solid sources, where practical.

Routine calibration of these monitors is in accordance with technical specifications provisions using solid sources related-to the initial calibration.

Calibration sources used are Sr-90, Cs-137, and Co-60 for low range monitors end Cs-137'for high range-monitors.

The conversion of the instrument readings to release rates are determined using~the energy response of the detectors obtained during calibration.

Accident release rates are then calculated based on anticipated radionuclide inventories following a design basis loss of coolant accident.

Actual releases may be determined by analyzing a grab sample and correcting the release rata calculated.

Continuous strip chart recording and CRT display are provided in the control room.

Digital readout for the high range effluent monitors RE-126 and 134 will assure the availability of continuous reading in.

the control room during or after an accident.

The effect of background radiation on readings of RBSVS noble gas monitors (RE-021, 022, and 134) and station vent exhaust monitor (RE-126) will be minimized during an accident, due to their location in the control building or turbine building (RE-134) and (RE-126) and the detector's location in a 4 7/ lead shield.

For the station vent exhaust monitor (RE-042), background radiation in the vicinity of

.the monitor within the secondary containment will have minimal effect on the noble gas detector, due to its location in a 4 telead shield and the fact that the detector, is a thin beta scintillator.

This type.of detector is very inefficient for detecting gamma radiation which might penetrate the lead shield, while it is efficient for detecting the beta radiation associated with the sample stream's noble gases brought in close contact with the detector.

For a listing of the radiation monitors with the ranges provided, refer to Table II.F.1-4.

II.F.1-4 1/5/82-I

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SNPS-1rFSAR

' ATTACl!MENTf 2 SAMPLING AND ANALYCIS OF PLANT' EFFLUENTS-

-~NRC Position Because iodinergaseous effluent monitors for the accidentLcondition are not considered to be practical at.this time, capability for.

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effluent monitoring of'radiciodines for the. accident. condition shall be provided with sampling conducted by adsorption on ' charcoal or other. media,Efollowed by onsite laboratory analysis.

'.y Licensees.shall1 provide continuous sampling of plant-gaseous effluent.

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for;postaccident releases of' radioactive iodines and-particulates to meet the requirements of the enclosed ~ Table.II.F.1-2.

Licensees-shall also. provide onsite laboratory capabilities to analyze or measure' these samples.. This requirement should not be construed to prohibit design and development of radioiodine and particulate. monitors to

~

provide online sampling and analysis for the accident' condition.

If

-gross gamma radiation measurement techniques are used, then provisions shall be made to minimize noble gas interference.

The shielding design basis-is given in Table II.F.1-2.

The sampling system design shall be such -hat plant personnel could remove samples, replace smapling media and transport the samples to the onsite analysis facility with radiation e-

>sures that are not in excess of the cri-teria of GDC 19 of 5-rem whole-body exposare and 75 rem to the extre-mities during the duration of the accident.

The design.of:the systems for the sampling of particulates and iodines should provide for sample nozzle entry velociti~cs which are approxi-mately isokinetic (same velocity) with expected induct or instack air velocitics.

For accident conditions, sampling may be complicated by a reduction in stack or vent effluent velocities to below design levels, making it necessary to substantially reduce sampler intake flow rates to achieve the isokinetic condition.

Reductions in air flow may well be beyond the capability of available sampler flow controllers to maintain isokinetic conditions; therefore, the staff will accept flow control devices which have the capability of maintaining isokinetic conditions with' variations in stack or duct design flow velocity of i 20 percent.

Further departure from the isokinetic condition need not be considered in design.

Corrections for non-isokinetic sampling conditions, as provided in Appendix C of ANSI 13.1-1969 may be con-sidered on an ad hoc basis.

Effluent streams which may contain air with entrained water, e.g.,

air ejector discharge, shall have provisions to ensure that the adsorber is not degraded while providing a representative sample, e.g.,

heaters.

II.F.1-5

1/5/82_

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SNPS-1 FSAR' LILCO Position The: normal station. vent. exhaust. monitor (RE-042) is not powered

fromia~ vital instrument bus, however, it is powered from a depen-dable backup power supply.to' normal ac.

Due to its location in

the secondary containment, it may be. inaccessible during an acci-dent.

This would-preclude obtaining the radiciodine and-particu-

-late sample media'from.the monitor for analysis.

However, inability to obtain these samples is~ compensated.for by the fact that the turbine building and radwaste building ventilation flows are each sampled; for-radiciodine and particulates by the equipment associated

! with the normal range noble gas monitors for these flows :(RE-057 and 055).

These monitors are both located in the turbine building permitting access for collection of the sample media during an accident'in order that laboratory analysis-may be performed.

Adding the results obtained for radiciodine or particulates from the turbine building and radwaste-building ventilaticn flows will give the radio-iodine 1or particulate release at the station vent exhaust should the secondary containment be inaccessible.

Under these circumstances, RBSVS is operating and there is no reactor building ventilation contribution to the station vent exhaust.

As discussed above, the RDSVS. release is monitored separately for noble gases and_ continuous collection of samples for particulates and radiciodine releases (RE-021, 022, and 134).

These monitors are capable of representative monitoring and sampling for-all accident conditions except for pipe break outsido containment (refer to Appendix 3C).

The monitors associated with the reactor, radwaste and turbine buildings ventila-tion systems are not powered from a vital bus.

This ic consistent with the design of the monitored systems.

The post accident station vent exhaust monitor (RE-126), located in the turbine building, will be accessible during an accident.

The station vent exhaust monitor RE-042 radiciodine and particulate sample media can be obtained for analysis if the seconlary containment is accessible.

The addition of the high range station ventilation exhaust monitor (RE-126) assures continuous campling of radiciodine and particulates during accident conditions.

Continuous sampling is achieved with isokinetic sampling during normal operation and accident conditions.

Provisions have been made to comply with ANSI N13.1-1969 to the maximum extent practical to assure representative sampling.

The sampling collector will initiate an alarm in the control room when it reaches a concentration of 102 /c Ci/cc and 30 min collection time.

At this time the microcomputer associated with RE-126 transfers the flow to the next particulate and iodine assembly, isolates the alarmed assembly, and indicates to the operator the need to replace the collector assembly and transfer it to the laboratory for analysis.

The sampling media is paper with more than 90 percent collection efficiency for 0.3 micron particles and a charcoal cartridge witn more than 90 percent collection efficiency for methyl iodide.

II.F.1-6 1/5/82

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SP!PS-1 FSAR l

The radioi.odine and particulate sampling media is analyzed in the counting coom at Shoreham.. Charcoal cartridges are purged with nitrogen or air to remove entrapped noble gases.

A separate counting station is.provided'which serves as a backup for the counting facility in the radiochemistry laboratory.

At least one of these locations will remain a low-contamination, low-background area for all postulated accident conditions.. The above meets the requirements of Table II.F.'l-2.

Further, procedures will be prepared for conducting all aspects of the measurement and analyses c orrectly and in a manner to minimize personnel exposure.

II. F.1 -7

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SNPS-1 FSAR TABLE II.F.1-4 RADIOACTIVITY CONCENTRATION RANGES FOR SHOREHAM GASEOUS EFFLUENT RADIATION MONITORS.

GASEOUS EFFLUENT MONITOR RANGE Reactor Building Standby Ventilation RE-021, RE-022*

lx10-6 to lx10+2 Post Accident Reactor Building

-Standby Ventilation RE-134*

1x10-2 to 1x10+4 l

Reactor Building Normal Ventilation RE-029*

lx10-6 to lx10-1 Turbine Building Ventilation

-1 RE-057*

lx10-6 to lx10 Radwaste Building Ventilation

.RE-055*

lx10-6 to lx10-1 Station Vent Exhaust RE-042*

lx10-6 to lx10-1 Post Accident Station Vent Exhaust RE-126*

lx10-2 to lx10+4 l

Reactor Building Normal Ventilation RE-068*

1x10-2 to 1x10+3 l

Turbine Building Ventilation RE-067*

lx10-2 to lx10+3 Fadwaste Building Ventilation RE-066*

lx10-2 to lx10+3 l

l Station Vent Exhaust RE-069*

lx10-2 to lx10+3 l

Drywell Monitors RE-08 5A, B lx100 to 1x107 R/hr

• Ranges shown for these radiation monitors are for the noble gas portion of the monitor.

1 of 1 L

R/3/@2