Text

Forwards Responses to Request for Addl Info Re Plant FSAR Concerning Radiation Protection Issues
	ML20245E010
Person / Time
Site:	Comanche Peak
Issue date:	04/28/1989
From:	William Cahill; TEXAS UTILITIES ELECTRIC CO. (TU ELECTRIC)
To:	; NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
	TXX-89226, NUDOCS 8905010265
	Download: ML20245E010 (57)
	v • d • e

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Log # TXX-89226

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File # 10010, 912.3, r

C 918 clo 1UELECTRIC April 28, 1989 W. J. Cahill Executwe Vice President U.-S. Nuclear Regulatory Commission.

Attn: Document Control Desk Washington, D. C.

20555

SUBJECT:

' COMANCHE' PEAK STEAM ELECTRIC STATION (CPSES)

DOCKET NOS. 50-445 AND 50-446 RESPONSES T0 NRC REQUEST FOR ADDITIONAL INFORMATION ON RADIATION PROTECTION ISSUES Gentlemen:

The attached responses address the NRC questioris on the CPSES FSAR, sections

.12.3.4 and'II.B.2.

FSAR changes which may-be needed as a result of these responses will be incorporated in a future FSAR amendment.

Si nce rely,-

/

1 fe:^y William J. Cahill, Jr.

GLB/daj Attachment c - Mr. R. D. Martin, Region IV

. Mr..J.11. Wilson, NRR-0SP Ms. M._Malloy, NRR-OSP Resident Inspectors, CPSES (3) n yd4

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Attachment l'to TXX-89226 l

April 28, 1989 l

Page 1 of 4 i

1 Response to Staff Questions on CPSES Radiation Protection Issues Question 1:

For FSAR Section 12.3.4, identify all accessible areas that could reach 100 rem /hr dose rate, for any period of time, and describe the design considerations that will prevent an overexposure of personnel.

Particular attention should be focused on areas with transient high dose rates, such as around the fuel transfer tube and the reactor cavity sump, as well as areas with intense sources of radiation continuously present.

Response to Question 1:

Fourteen accessible areas are identified within which expected dose rates could reach 100 rem /hr or greater during certain normal operational conditions. These areas are described below, as well as necessary conditions for occurrence and principal means of design protection against overexposure:

1.

Reactor Cavity Building: Containment Room No:

1 Elevation:

781 ft-7 in.

Conditions:

During power operation or during shutdown with incore instrumentation withdrawn Protection:

Primary shielding, locked door 2.

Incore Instrumentation Rooms Building: Containment Room Nos:

2, 10, 11 Elevations:

808 ft, 832 ft-6 in., 849 ft Condition:

During removal of incore instrumentation Protection:

Labyrinth shielding, locked doors 3.

Excess Letdown Heat Exchanger Room Building: Containment Roon No:

3 Elevation: 808 ft Condition:

During power operation Protection:

Labyrinth shielding, locked door

_- to TXX-89226 April 28, 1989 Page 2 of 4 4.

Regenerative Heat Exchanger Room I

Building: Containment Room No:

9 Elevation: 832 ft-6 in.

Condition:

During power operation Protection:

Labyrinth shielding, locked door 5.

Steam Generator Compartments Building: Co ainment Room Nos: 25, 24, 25, 26, 27, 28, 29, 30 Elevations: 812 ft, 841 ft Condition:

During power operation Protection:

Labyrinth shielding, locked doors 6.

Fuel Transfer Tube Area Building: Containment Room Nos: 33, 34 (connecting area)

Elevation: 808 ft Condition:

During spent fuel transfer Protection:

Labyrinth shielding, locked gates 7.

Pressurizer Room Building: Containment Room No:

16 Elevations: 861 ft-6 in., 905 ft-9 in.

Condition:

During power operation Protection:

Labyrinth shielding, locked doors 8.

Spent Resin Sluice Pumo Room Building:

Auxiliary Room No:

191 Elevation: 810 ft-6 in.

Condition: During sluice pump operation Protection:

Labyrinth shielding, locked door 9.

Spent Resin Storage Tank Room Building:

Auxiliary Room No:

192 Elevation: 810 ft-6 in.

Condition:

During spent resin storage Protection:

Labyrinth shielding, locked door

' Attachment 1 to TXX-89226 April 28, 1989 Page 3 of 4

10. Demineralized Piping Area Building:

Auxiliary Room No: 213 Elevation: 831 ft-6 in.

Condition: During spent resin ' transfer Protection:

Locked doors, posting

11. Filter Pit Monorail Corridor Building: Auxiliary Room No:

234 Elevation: 852 ft-6 in.

Condition:

During waste process filter removal Protection:

Locked gates, posting

12. Maste Process Module System Area Building:

Fuel Room No:

261 i

Elevation: 822 ft-4 in.

Condit!.on:

During spent resin transfer Protection:

Locked door, posting

13. Waste Conditioning Tank Room Building:

Fuel Room No:

270A Elevation: 841 ft Condition:

During spent resin conditioning Protection:

Labyrinth shielding, locked door

14. Calibration Source Storage Area Building:

Turbine Building Annex Area location:

Building roof Condition:

During high-range instrument calibration Protection:

Fence, locked gate, posting, flashing light The method of protection for these areas may vary with time, depending on actual radiological conditions and plant modifications, but adequate design considerations will be provided commensurate with the radiological hazard to allow the physical access controls required by 10CFR20.203.

Athchment 1. to TXX-89226 April 28, 1989 Page 4 of 4 Question 2:

FSAR Section II.B.2 discusses the plant design review to ensure access to vital areas following an accident, but does not elaborate on the results of the review.

Provide an analysis which demonstrates that the requirements of NUREG-0737 Item II.B.2 are met anJ revise the FSAR accordingly.

Response to Question 2:

The plant design review to ensure accessibility to vital areas following an accident has been completed.

This analysis demonstrates that the requirements of NVREG-0737, Item II.B.2 are met.

The enclosed FSAR pages have been marked up with. proposed changes to indicate the results of this review.

(NOTE:

Figures II.B.2-1 through II.B.2-24 are scheduled to be deleted.) 7Iiis review also includes an updated analysis which demonstrates post-accident sampling capability as prescribed in NVREG-0737 Item II.B.3.

Tables II.B.2-4 and II.B.2-5 with Figures II.B.2-25 through II.B.2-40 are indicative of this portion of the study.

These reviews show that post-accident manual actions can be achieved and the operator can access the locations where these actions must be performed.

Actual paths, times, and required actions will vary depending on the specific accident scenarios and actual plant configuration at the time of the accident but the reviews demonstrate that the design criteria at CPSES will allow personnel to perform post-accident required actions.

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Enclosure to.TXX-89226 Page 1 of 52 CPSES/FSAR RESPONSE TO NRC ACTION PLAN L

29 l 11.B.2.2 System Review 66 l After establishing the source terms noted above, the systems which may l contain highly radioactive materials in a post-accident situation have lbeenevaluated.

The specific systems reviewed are described in the 29 l following subsections.

29 l11.8.2.2.1 Containment 66 lDirectradiationpenetratingthroughthecontainmentwallsand l radiation streaming through penetrations was based on the source term l discussed in Section 12.2.1.3.1.

Radiationstreampgcausesa l significant dose rate [n the vicinity innediatelyJoutside the _A 53 Q ontainment equipment hatchf At ground level, a Zone IV region "f extends to a distance of 80 feet from the hatch, and a Zone III l condition exists for the region between 80 and 300 feet from the l hatch.

56 l

.29 l11.B.2.2.2 Emergency Core Cooling System Amendment 66 January 15, 1988 11.B-6

Enclosure to TXX-89226 Page 2 of 52 Insert I l

i in a field within ! 25' from the hatch center line.

The dose i

rate changes with time and distance.

Immediately following a LOCA the dose rate is calculated to be about 44 rem per hour at a i

point 30 feet from the containment surface and about 4 rem per i

)

hour at 240 feet.

These dose rates decrease to less than 2 rem per hour an'd 0.06 rem per hour respectively at 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months after the i

accident.

1 O

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' Enclosure to.TXX-89226 CPSES/FSAR Page 3 of 52 RESPONSE TO NRC ACTION PLAN Plant shielding has been evaluated considering the Emergency Core l 66 Cooling System (ECCS) to be operating in the recirculation mode.

The l residual heat removal pumps take suction from the containment sump, l

and the safety injection pumps and centrifugal charging pumps take j

suction from the residual heat removal pump discharge.

l Prior to the start of recirculation, the ECCS will contain water from j66 he refueling water storage tank. Although recirculation operation l

g[is not initiated untU f5) minutes to several hours af ter the start of l N

an accident the shielding review has assumed that at the beginning of l

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the accident the ECCS contains the source terms described in Section l

12.2.1.3.4.

l-II.B.2.2.3 Resiffual Heat Removal System l29 Plant shielding has been evaluated considering the Residual Heat l 46 Removal System (RHRS) contains the source terms specified in Section l~

12.2.1.3.3 for nondepressurized LOCA conditions, or the source terms l

specified in Section 12.2.1.3.4 for depressurized LOCA conditions as j

appropriate.

l II.B.2.2.4 Containment Spray System l29 Plant shielding has been evaluated considering the Containment Spray l29 System (CSS) to be operating in the recirculation mode.

Prior to the l

start of recirculation, the CSS will contain water from the refueling l

water storage tank.

Although CSS recirculation, operation is not l66 initiated unt inutes to several hours after an accident (if l

required), the shielding evaluation has assumed that at the beginning l

of the accident the CSS contains the source terms described in Section l 12.2.1.3.4.

l II.B.2.2.5 Chemical And Volume Control System l29 a[fr0Krrnafoly/0 Amendment 66 II.B-7 anuary 15, 1988

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Enclosure to TXX-89226-CPSES/FSAR Page.4 of 52 RESPONSE 10 NRC ACTION PLAN II.B.2.2.7 Post-Accident Sampling System l29 L

As a result of the shielding review, the existing Primary Sample l29 System described in FSAR Section 9.3.2 was found to be incapable of l

obtaining and analyzing highly radioactive post accident samples l

without excessive personnel exposures.

Therefore a new Post-Accident l66 Sampling System (PASS) has been installed at CPSES as described in l

Section II.B.3.

I Plant shielding has been evaluated considering PASS components and l66

. sample lines containing either liquid sources as described in Section l

12.2.1.3.3 or 12.2.1.3.4, or the containment atmosphere source as l

described in Section 12.2.1.3.2 as appropriate.

l II.B.2.3 Shielding Methods l29 Evaluation of plant shielding for equipment qualification dose l66 evaluations following an accident includes direct radiation from the l

containment, and radiation from piping and components of systems l

discussed in Section II.B.2.2, as appropriate.

Direct radiation l

)

through shield walls and streaming and scattered radiation through l

)

penetrations were considered.

FSAR Section 12.3.2.2 discusses the l

. analytical methods used in the shielding analysis.

l II.B.2.4 Design Review l66 The plant shielding design review has identified the radiation l 66 exposure rates in vital areas requiring personnel access following an K

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56g accident.[Basedontheplantshieldingdesignreview, post-accident 3l

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9 radiation zone drawings have been developed to illustrate post-

'l h

accident radiation levels within the plant.

These drawings are l

presented as Figures II.B.2-1 through II.B.2-24.

The post-accident l

H radiation zone drawings describe the radiation levels in the Unit 1 l

l and comon structures resulting from a postulated accident in Unit 1. ' l

)

Due to symmetry, an accident in Unit 2 would cause similar radiation l

l levels in Unit 2 and common structures.

Radiation levels arising l

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fromnormallyradioactivesourcessuchasthespentfuelpools,gasj i

II.B-9 Amendment 66 January 15, 1988 j

i

1" Enclosure to TXX-89226 Page 5 of 52 1

l' Insert II l

The vital areas of the plant which may need periodic access after a LOCA for 'erforming manual operations to mitigate the consequences of the accident are described in Table II.B.2-4.

This table also presents the operator path or route numbers, the time after LOCA when access to the area is required and the calculated radiation dose that may be received by the operator during the performance of the assigned task.

-}

Plant drawings have also been developed to illustrate the post LOCA access routes to the vital areas.

These drawings are presented in figures II.B.2-25 through II.B.2-40.

Radiation dose rates between various points of the operator path, the time spent between consecutive points and the path or route number are shown in Table II.B.2-5.

~' Enclosure to TXX-89226 CPSES/FSAR Page 6 of 52' RESPONSE TO NRC ACTION PLAN T

6

.l decay tanks, and demineralizers are also included in the post-accident l-radiation zone drawings.

These areas are identified by notes-f

. l-included in the drawings as appropriate.

66 l'

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Amendment 66

-January 15, 1988 II.B-10 i

w--______-__-___.

Enclocure to TXX-89226 CPSES/FSAR Fage 7 of.52 RESPONSE TO NRC ACTION PLAN The Control Room Complex and the Technical Support Center (TSC) are l29 vital areas requiring full-time occupancy during the course of an l

accident.

The integrated dose for these areas is less than 5 rem l

whole-body, or equivalent, for the duration of the accident in kl accordance with As can be seen from Figures II.B.2-I7 and l 66

[11.B.2-18thedoserateintheseareasislessthan15 mrem /hr.

l 1

NI

[l 6eW:

The post accident radiation zone drawings, Figures II.B.2-1 through l66 ec.tc M-II.B.2-24, provide a basis for assessing post accident exposures in l

4 g

various plant areas.

Areas that may require limited post accident

,l 0

access include the hot laboratory, counting room, post-accident sample l29 panel, motor control centers, ESF switchgear, emergency power supplies l

o.

p (diesel-generators and batteries), and instrument panels. Access to l

qMf these areas may be required on an irregular basis to perform l

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f"*

surveillance and inspection and post-accident radiochemical analyses.

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To meet the requirements for post-accident sampling, a new system, ll66 described in Section II.B.3, has been added.

This new Post-Accident l

Sampling System (PASS) incorporates remote operation capabilities and sufficient shielding to ensure that the dose criteria dclineated in Section II.B.3 are met.

l 4

ta p

i Amendment 66 11.B-11 January 15, 1988 i

Enclosure'to TXX-89226 Page 8 of 52 Insert IJJ l-Motor' control centers (MCC's) are not considered to be vital areas for purposes of this review, since single failure criteria were assumed, and a single failure would not prohibit the ability of the MCC's functions to be performed from the control room.

Access to radwaste control panels is not considered to'be vital, since the safety injection signal during postulated DBA conditions will isolate reactor coolant letdown.

Therefore, no highly radioactive post-accident fluids wf11 be present in the radwaste systems.

" Enclosure to TXX-89226

.CPSES/FSAR Page 9 of 52 RESPONSE TO NRC ACTION PLAN 291 l11.B.2.5~

Radiation Qualification of Class IE Equipment 66 l The post-accident radiation exposure to Class IE equipment located i

l outside and inside containment has been evaluated for post-accident l sources postulated to exist following a LOCA. The source terms for 1

l equipment located outside containment are described in Section l 12.2.1.3.3, 12.2.1.3.4, or 12.2.1.3.5, as applicable. For Class IE l equipment inside containment, the larger dose resulting from sources bll described in either Section 12.2.1.3.3 or Sections 12.2.1.3.1 and (g@go 12.2.1.3.4 were utilized.

Equipment dose evaluations for Class 1E f ll of NUREG 058g. These dose evaluations conside equipment inside and outside containment are based on the guidelines

(

(j [

l radiation, as appropriate for post-accident conditions, and incl 6 Y

l contributions from normal operations.

The calculated post-accident l radiation dose values for Class 1E equipment are included in FSAR l Appendix 3A.

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66 l

II.B.3

. POST-ACCIDENT SAMPLING 4

Action Plan Requirements:

"A design and operational review of the reactor coolant and containment atmosphere sampling line systems shall be performed to determine the capability of personnel to promptly obtain (less than I hour) a sample under accident conditions without incurring a radiation exposure to any individual in excess of 3 and 18-3/4 rem to the whole body or extremities, respectively. Accident conditions should assumt.

a Regulatory Guide 1.3 or 1.4 release of fission products.

If the review indicates that personnel could not promptly and safely obtain the samples, additional design features or shielding should be provided to meet criteria.

"A design and operationa~, review of the radiological spectrum analysis facilities shall be performed to determine the capability to promptly quantify (in less than 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months ) certain radionuclides that are indicators of the degree of core damage. Such radionuclides are noble gases (which indicate cladding failure), iodines and cesiums (which Amendment 66 11.B-12 January 15, 1988

p' Enclosure to'Txx-89226-CPSES/FSAR Page 10 of'52 RESPONSE TO NRC ACTION PLAN correspond to a Regulatory Guide 1.3 or 1.4 release.

The review should also consider the effects of direct radiation from piping and components in the auxiliary building and possible contamination and direct radiation from airborne effluents.

if the review indicates that the analyses required cannot be performed in a prompt manner with existing equipment, then design modifications or equipment procurement shall be undertaken to meet the criteria.

"In addition to the radiological analyses, certain chemical analyses are necessary for monitoring reactor conditions. Procedures shall be provided to perform boron and chloride chemical analyses assuming a highly radioactive initial sample (Regulatory Guide 1.3 or 1.4 source tenn).

Both analyses shall be capable of being completed promptly (i.e., the boron sample analysis within an hour and the chloride-sample analysis within a shift)."

CPSES Response

7 lII.B.3.1 General Description 27 l FSAR Section_9.3.2 describes the cess sampling system l

stalled in CPSES.

This system was designed for the lcapabilityofobtainingsamplesduringnormaloperatingconditions.

l During postulated accident conditions, radiation levels would be lsignificantlyincreased.

Therefore, additional sampling capability dded to CPSES to specifically address the requirements for ge[ ] post accident sampling. This new sampling system will be installed l and operational prior to op stion above 5% power.

27 l A review of the reactor coolant, containment atmosphere and lcontainmentsumpsamplingsystemsandtheradiologicalspectrumand }

lchemicalanalysisfacilitieshasbeenconducted.@ntmodifications) 36 lhe being implemented to permit personne1No obtain and analyze (isamples within three hours after a decision is made to take a sample 1

( % e eesul4S 08 -\\h'CSCeVAeod bd'

-\\he colpbilify January 15, 1988 11.B.14 L_ c_ ___ _ _

s

Enclosure to TXX-89226 Page.11 of 52 CPSES/FSAR l'

RESPONSE TO NRC ACTION PLAN e

Containment at a different elevation than the sampling l 40 locations.

l-L 4.

An area radiation monitor p included within each sample module l36 tomonitorlocal[ activity 1evelswithinthemodule.

l J

p d iG **0 0 l

5.

A dedicated sample cooler is provided for the RC PASS function.

l 36 6.

Post accident sampling does not require an isolated auxiliary l 36 system to be placed in operation in order to use either PASS l

subsystem.

l 7.

In order to permit post-accident sampling, the capability to l 36 override the containment isolation signal for the appropriate l

sample isolation valves is provided.

l-

~

- 8.

PASS samples can be obtained from the following sample points in-l 36 each unit.

l Reactor Coolant System hot leg No. I and hot leg No. 4 l 36 a.

normal sample lines.

l b.

Containment sump via a connection to the ECCS l 36 recirculation (RHR) system which can be used when in the l

recirculation mode of operation.

l c.

Co.ntainment atmosphere from two different elevations l 36 (sample points) in the containment.

l 9.

The containment penetrations and containment isolation valves l 36 for PASS sample lines are Class 2 and Seismic Category 1.

l Other portions of the PASS sample lines are Class 5.

l i

II.B-17 January 15, B88

)

e l

Enclosure to TXX-89226 CPSES/fSAR i

Page 12 of 52 l

RESPONSE TO NRC ACTION PLAN I

16 l 10.

Provisions are included for purging sample lines, for reducing l

plateout-in sample lines, for minimizing sample loss or l

distortion, for preventing loss of sampling capability due to l

loose material in RCS or Containment, and for flow restriction l

to limit reactor coolant loss from a rupture of a sample line.

l Sample lines are as short as practicable to minimize the volume l

of sample fluid outside containment.

36 l 11.

In order to assure that a representative sample reaches the PASS l

sample station, the Containment atmosphere sample lines are heat l

traced so that the temperature is maintained and condensation is l

minimized until the sample reaches the sample panel.

l Provisions are included to permit containmen* atmosphere l

sampling under both positive and negative containment pressures.

40 l 12.

In order to minimize the amount of radioactivity outside l

containment in a post-accident situation, the PASS will be l

capable of flushing all lines from PASS equipment near the l

external containment isolation valve to the sample return point.

l Liquid sample lines are flushed with demineralized water to l

the reactor coolant drain tank inside containment. Gas sample l

lines are purged back to the containment using nitrogen gas.

36 l13.

Electrical power to the non-safeguards components of the PASS l

are powered from a Class IE isolation transformer connected to a l

Safety-Train A motor control center.

This highly reliable non-l Class 1-E, electrical system (backed-up by the Train A diesel l

generator) provides assurance that samples can be obtained l

within three hours of the decision to sample, assuming loss of l

offsite power. Additionally, indication of sample pressure, temperature, pH, conductivity and k oss radioactivity e

obtained at the PASS Remote Operation Modules assuming a loss

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January 15, 1988 11.8-18

Enclosure to TXX-89226'.

t Page 13 of 52 CPSES/FSAR RESPONSE TO NRC ACTION PLAN l

of off-site. power (see Section 11 B.3.3).

Operation of'the l36 plant hot laboratory and counting room requires normal AC power. l The Containment Isolation valves which are part of the PASS l

arecontrolledbyaclass1Evalvecontrolpanel,locatedinthel Cable Spread Room.

(Electrical and Control Building, elevation l

807'-0").

This location provides a low post-accident radiation l

exposure to the panel operator.

l

~

bl4.

When plant modifications are completed and procedures l36 implemented, testing will be conducted to demonstrate the l

capability to obtain and analyze a sample according to the l

design criteria listed above.

l Onsite facilities (including PASS and hot laboratory facilities) and l36 procedures'are developed which provide the capability to quantify the l-'

following.

l1

-1.

Gross radioactivity and certain isotopes that are indicators of l36 the degree of core damage (i.e., noble gases. iodines, cesiums l

and significant non-volatile isotopes).

l 2.

Dissolved gases (i.e., H2 and 0 ), boron concentrations, l36 2

chloride and pH of liquids.

l 3.

Hydrogen and oxygen levels in the containment atmosphere.

The l36 CPSES design includes additional provisions for monitoring of l

hydrogert concentration via the Containment Hydrogen Monitoring l

System as described in FSAR Section 6.2.5.

l i

January 15, 1988

-Enclosure to TXX-89226 CPSES/FSAR Page 14 of 52 RESPONSE TO NRC ACTION PLAN 36' lII.B.3.3 PASS Equipment 36 l As stated above, the CPSES PASS is comprised of too independent l subsystems, the Reactor Coolant PASS (RC PASS) and the Containment Air l(CAPASS). A discussion of the major components of each of these 46 l subsystems is provided below. Also, a discussion of the control

.l equipment required to override (by-pass) Containment isolation is lprovided.

36 l!!.B.3.3.1 Reactor Coolant PASS 36 l The Reactor Coolant PASS (RC PASS) will be used to obtain micro-volume l samples of undiluted reactor coolant, reactor coolant off-gas, and l Containment sump water. The Containment sump sample will be taken lfromtheRHRsystemviaECCSrecirculation.

The RC PASS, as shown in l Figure II.B.3-1, includes the following major components:

~

16 l

RC PASS Sample Module 36 l

RC PASS Remote Operating Module 36 l

RC PASS Flush and Diversion Manifold 36 l

RC PASS Auxiliary Module 36 l

RC PASS Flush Module 36 l

RC PASS Sample Cooler 36 l The RC PASS Sample Module is located in the primary sample room; it l contains all the equipment required to perform the sample acquisition l process. This process is remotely controlled at the RC PASS Remote g

lOperatingModule.

Thesamplemoduletrapssmallvolumesof(undiluted]

l reactor coolant, reactor coolant off-gas, and Containmtat sump water lintwoshielded,pressurevessels. One vessel traps a liquid sample

]andtheothertrapsanoff-gassample.

L 36 l The actual sample removal is a manual task, accomplished by the use of l

l pressure lock, micro-volume syringes.

These syringes are inserted l into the pressure vessels through needle ports in the shields.

j January 15, 1988 II.B-20 l

l l

1

^

Enclosure to TXX-89226

.CPSES/FSAR C

"E" RESPONSE TO NRC ACTION PLAN i

The samples contained in the syringes are either immediately injected l36 into a septum bottle (contained in a cart mounted, shielded cask) or l

are placed into a shielded syringe cask, also mounted on a transport l

cart.

The micro-volume samples acquired by the above process supply l

sufficient sample volume to perform the required radiological l44 analyses, chloride analysis, and the boron analysis.

l Should it become necessary to.obtain a larger volume of sample, for l66 shipment to an offsite laboratory, a 5ml grab sample can be acquired l

in the RC Sample Module by operations at the RC PASS Remote Operating l

Module.

This sample is trapped in a lead " pig" with nominal 2" lead l

shielding.

The pig is manually disconnected from the sample module l 36 (after flushing) and is loaded into a shielded transport cart for l

transport to the hot laboratory.

The sample will be stored either in l66 the Hot Lab or an area, within the Radiation Controlled Area, l;

designated by Radiation Protection personnel.

1 For the pH and conductivity analyses, the Sample Module contains in-l36 line instrumentation with remote read-out at the Remote Operating l

Module. Remote calibration of the pH and conductivity cells is l 40 controlled from the RC PASS ROM.

A calibrated metering pump takes l

suction from a bottle of known standardizing solution.

This l

equipment is furnished with the RC PASS sample module.

In addition, l

the Sample Module contains temperature and pressure sensing l

instrumentationneces3arytopermitananalyticaldeterminationofthel quantity of dissolved gas contained in a reactor coolant sample, and l

it has an emergency spray system to aid in the removal of radioactive l

~

contamination inside the sample module, should a leak occur.

l Indication of such a failure can be ter '

from the radiation

[ 36 weeb monitor located inside the sample module.

his monitor has a remote l

read-out in the RC PASS Remote Operating Module.

l Amendment 66 II.B-21 January 15, 1988-

Enclosure to TXX-89226 P SWR Page 16 of.52 j.'.~

RESPONSE TO NRC ACTION PLAN l

46 lII.B.3.3.3 PASS Isolation Valve Control Panel 46 l The PASS Isolation Valve Control Panel (P!VCP) is a Class IE, Seismic l Category I electrical control panel.

It is located in the cable j spreading room, a Q radiation area.

low 46 l The PIVCP is provided to permit control and manual override (by-pass) l of the Containment Isolation Phase "A" signal for all isolation valves lassociatedwithboththeRCPASSandtheCAPASS.

Without such an-l override, post-accidentsamplingwouldnotbepossibleinthepresence l of a Phase "A" signal.

-46 l The PlVCP has control switches for the operation of twelve (12) 54

-lisolationvalves.

Three (3) of these valves can also be controlled l during normal operation of the Sample Valve Control Panel (SVCP).

54 l Train oriented selector switches are provided at the PlVCP to select lcontrolsattheSVCPorPIVCP.

The remainino nine (9) valves are lcontrolledexclusivelyatthePlVCP.

In addition, two train oriented lkeyoperatedswitchesconfigured" NORMAL-BYPASS"areprovidedonthe l PIVCP to override containment isolation Phase "A" signal.

Bypass of l Phase "A" Signal is annunciated in the control room.

46 lThe'PIVCPisunderdirectadministrativecontrolofthecontrolroom l operator.

Train oriented control switches configured "CLOSE-LOCAL" 54 l are provided in the control room.

Placing these switches in the l " LOCAL" position permits operation of the 12 isolation valves from the

~

l PlVCP. However, by placing the control room switches in the "CLOSE" l position, the override capability at the PIVCP is negated.

46 l Administrative control of the containment isolation override is lthereforemaintainedbythecontrolroomoperators.

36 lII.B.3.4 Sample Analyses l

January 15, 1988

Enclosure to TXX-89226

-Page 17 of 52 CPSES/FSAR RESPONSE TO NRC ACTION PLAN After the post accident-samples are obtained by the sample modules, l36 anti are placed in a shielded cask, the cask is transported via a l

transport cart to the hot laboratory for analyses.

Sample l66 preparation will be done in the Hot Lab fume hood.

Temporary l

shielding will be used to reduce the exposure to personnel preparing l

the post-accident sample for analysis.

_ l%

Radiological analyses o thesampleswillbeperformedinthe[ Counting 44 (Room (adjacent to the ot Laboratory As shown in Figure 11.82-12, l

e post-accident radiation analysis shows that the maximum radiation l 36 1

level to be expected in the plant counting room is 15 mrem /hr.

This l

hs on the order of 1500 times the normal background.[High activity l

samples such as post-accident coolant and containment air samples are l

expected to yield instrument count rates 10 times greater than that l

resulting from the elevated background.

Therefore, the results of l

the counting analyses should have an accuracy within a factor of two.

l4 However, samples having lower activity and requiring the analyses l~

ofspecificisotopesthatmaghaveexcessivebackgroundinterference l

Gn the olant enuntinggcan be transferred to the Emergency l

Operations Facility counting room for analyses, where the background l

/-

radiation level should be several orders lower than thehant countIn')l The PASS will be used to acquire routine samples, such that it l44 will be used at least once every six months.

The routine use of the l

PASS will assure its operability and will also assure proper operator l 36 training and familiarity with the system.

l i

l l

Amendment 66 anuary 15, 2 8 l

l 11.0-25 l

'. Enclosure to TXX-89226 CPSES/FSAR pop 18 f 52 RESPONSE TO NRC ACTION PLAN 36 l Table II.B.3-1 lists the chemical and radiological analyses l capabilities that wi' be utilized to perform the required post 40 l accident sample ana;

i. A discussion of these sample analyses is lprovidedbelow.

40 l II.B.3.4.1 Gross Activity (Liquid) 66 l A microliter sample taken with a micro-syringe at the sampling port of l the Post Accident Sampling System (PASS) will be placed in a shielded l container and transferred to the Hot Lab.

The sample will then be 36 l diluted in an analytically acceptable manner to a factor that will l allow conveniently placing approxiniately 500,000 - 1,000,000 dpm of l activity in a counting planchet for evaporation.

After evaporation lofthesampleintheplanchetitwillbetransferredtoacalibrated l proportional counter in the counting room for analysis.

This l procedure can be performed with a maximum error of 10%.

For post-

~

l accident samples, sensitivity would not be of concern.

36 lII.B.3.4.2 Gamma Spectrum (Liquid) 36 l A volume of diluted sample (prepared for the gross activity sample) l with an activity of approximately 500,000 to 10,000,000 dpm will be l placed in a standard gamma counting geometry and transferred to a l calibrated Germanium detector gamma spectroscopy system in the lcountingroom.

This procedure can be performed with a maximum error l of 10%.

~

36 l 11.B.3.4.3 RCS OFF-Gas Gamma Analysis 44 lAminimumvolume(approximately

1) of reactor coolant off-gas will

{

l be removed from the RC PASS Sample Module using a micro-syringe.

,l 36 l Larger volumes may be taken, depending on the level of sample 44 l activity,tooptimizetheanalysistime.

The sample will be injected l into a standard gamma counting geometry for analysis.

Amendment 66 ll.B-26 January 15, 1983 L_

F,nclousre to TXX-89226 CPSES/FSAR Page 19 of 52 RESPONSE TO NRC ACTION PLAN The same technique, gas chromatography, that is used for hydrogen gas l 44 analysis will be used for oxygen gas analysis. The maximum error for l this analysis is 10%, in the range of 0 to 30%.

l 11.B.3.4.11 Containment Air Particulate and Iodine Gamma Spectrum l36 The PASS will collect particulate and radiciodine isotopes on filters l 44 for analysis. The radiological measurements can be accomplished with l

a maximum error of 5-15%.

By acquiring appropriate sized volumes of l36 sample, the sensitivity of the analysis can exceed the MPC valves of l

10CFR20.

l II.b.3.4.12.

Containment Air Noble Gas Activity 00 A minimum volume (approximately 10 pl) of Containment air will be l.44 j

removed from the CA PASS Sample Module using a micro-syringe.

Larger l-volume samples may be taken, depending on the level of sample l 36 activity, to optimize the analysis time.

The sample will be injected l44 into a standard gamma counting geometry for analysis.

l II.B.4 TRAINING FOR MITIGATING CORE DAMAGE Action Plan Requirements:

"A program is to be developed to ensure that all operating personnel are training in the use of installed plant systems to control or mitigate an accident in which the core is severely damaged.

The training program should include the following topics."

"1.

Incore Instrumentation a.

Use of fixed or movable incore detectorr to determine extent of core damage and geometry changes.

January 15, 1988 II.B-29 x _ x__ _ ___ _ _

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