ML20054L139

Forwards Responses,Revised Responses & Partial Responses Re FSAR Questions in Order to Facilitate Completion of Review. Responses Will Be Included in FSAR Revision 6

ML20054L139
Person / Time
Site:	Catawba
Issue date:	06/28/1982
From:	Parker W DUKE POWER CO.
To:	Adensam E, Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML20054L140	List: ML20054L139 ML20054L146
References
NUDOCS 8207070182
Download: ML20054L139 (7)

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DtJKE POWER COMPANY Powen flustonwo 422 Soutes Cauwcnn STurzT, CuAumTTE, N. C. naa4a i

wik LIAM O. PARMER,JR.

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June 28, 1982 5 "" ": ^* 7 o*

Seeau Pacoucnow 373-40s3 l

1 Mr. Harold R. Denton, Director Office of Nuclear Reactor Regulation U. S. Nuclear Regulatory Commission Washington, D. C. 20555 Attention:

Ms. E. G. Adensam, Chief j

Licensing Branch No. 4 Re: Catawba Nuclear Station Docket Nos. 50-413 and 50-414

Dear Mr. Denton:

In order to facilitate the completion of the review of the Catawba FSAR, Duke Power Company is transmitting herewith responses, revised responses, or partial responses to the following FSAR questions:

281.5 440.96 410.10 440.110 410.33 450.4 4

420.7 480.6 430.9 480.8 430.18 480.9 i

440.43 480.11 440.94 t

These responses will be included in FSAR Revision 6.

5 Ver truly yours, A

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William O. Parker, Jr.

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Attachment cc:

Mr. James P. O'Reilly, Regional Administrator U. S. Nuclear Regulatory Commission Region II 101 Marietta Street, Suite 3100 i

Atlanta, Georgia 30303 pv l

j B207070182 820628 i

PDR ADOCK 05000413 i

A PDR

Mr. Harold R. Denton, Director June 28, 1982 Page 2 cc:

Mr. P. K. Van Doorn NRC Resident Inspector Catawba Nuclear Station'

- Mr. Robert Guild, Esq.

Attorney-at-Law 314 Pall Mall Columbia, South Carolina 29201 Palmetto Alliance

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2135 Devine Street Columbia, South Carolina 29205 Mr. Jesse L. Riley Carolina Environmental Study Group 854 Henley Place Charlotte, North Carolina 29207 Mr. Henry A. Presler, Chairman Charlotte-Mecklenburg Environmental Coalition 943 Henley Place

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Charlotte, North Carolina 28207 A

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d Table 8.3.1-1 Sheet 3 Catawba Nuclear Station Maximum Loads To Be Supplied From 0ne Of The Redundant Engineered Safety

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Power Distribution Systems SEQUENCE NO.

(a)

CONNECTED REQ'D.

REQ'D.

AND INITI-LOCA PER FOR FOR ATION TIME EQUIPMENT NAME TION SYSTEM VOLTAGE DIESEL LOCA BLACKOUT REMARKS

' No. 1 Diesel Starting Air Com-DB VG 575 20 HP 20 HP 20 HP 2-10HP/ diesel (cont'd) preFsor Motor l

AC Emergency Lighting Pnlbd.

AB ELA 600 30 KVA (e),(f) 30 KVA 1-30KVA/ diesel m

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Diesel 600/120V Panelboard DB EPY 600 5 KVA 5 KVA 5 KVA 1-5KVA/ diesel Diesel Generator Engine DB LD 575 7.5 HP 7.5 HP 7.5 HP 1-7.5HP/ diesel Prelube Oil Pump Motor l

Diesel Generator Engine Lube OR LD 575 3 HP (e) 3 HP 1-3HP/ diesel Oil Transfer Pump Motor Diesel Generator Engine Lube DB LD 600 48 KW 48 KW 48 KW 2-24KW/ diesel 4

Oil Sump Tank Heater Diesel Battery Charger

-DB EPQ 600 20 KVA

_20 KVA~

20 KVA 1-20KVA/ diesel Diesel Generator Room Sump DB WN 575 10 HP 10 HP 10 HP 2-5HP/ diesel Pump Motor Diesel Bldg. Generator DB VD 575 60 HP 60 HP -

60 HP 2-30HP/ diesel Vent Fan Motor i

Control Room Air Handling AB VC 575 50 HP 50 HP 50 HP 1-50HP/ unit Unit Fan Motor Train A 1EMXG Containme'nt Air Return CV VX 575 1 HP 1 HP 1-1HP/ diesel Isolation Damper Control Room Area Filter AB VC 575 25 HP 25 HP 25 HP-1-25HP/ unit Train Pressure Fan Motor Train A 1EMXG Rev. 6

o CNS

Response

The isolation valves used for containment isolation of the process sampling lines are electric motor operated and therefore fail "as is."

These valves are used in groups for each penetration with the isolation valves inside containment supplied by one train of safety related power while the valve outside containment receives power from the other train of safety related power.

Both interior and exterior valves receive an appropriate automatic signal to close.

Isolation of these lines is thus assured even with assump-tion of a single failure.

This meets the intent of GDC60 in Appen-dix A to 10CFR50.

281.9 Provide information that satisfies the attached proposed license (1.9, II.B.3) conditions for post-accident sampling.

(Attachment 281-1).

Response

The Catawba Post Accident Liquid Sampling System is identical to the system reviewed and approved for the McGuire Nuclear Station (NUREG-0422) and meets the requirements of NUREG-0737, II.B.2 as discussed in the following response to Attachment 281-1:

1. 0 Compliance With NUREG-0737 1.1 Each unit has a reactor coolant and a containment air sampling system.

The basics of both systems are the same and both systems are remote controlled.

A small sample is taken and diluted.

A small portion of the di-luted sample is saved, while the excess is flushed to a radwaste system.

Total sampling time is approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.

1. 2 a)

Isotopic analysis is run at the station counting room using a portion of gas stripped from the liquid sample and diluted with inert gas and a portion of diluted l

liquid sample.

The size of the samples and the dilu-l tion allows them to be handled and counted with the j

available equipment.

I b)

Hydrogen levels in the containment atmosphere can be determined by the hydrogen monitor located in the Auxiliary Building.

c)

Dissolved gases are stripped from the liquid sample and diluted 1:1000 in the Post Accident Sampling Panel.

The gas sample is then analyzed with a gas chromato-graph.

Other gases as well as the H can be determined.

2 Results of tests performed on this function are given i

in the report attached.

280-5 Rev. 6

(

o CNS Chloride analysis can be performed on a diluted liquid sample by ion-chromatography.

Currently a radiochemistry laboratory is being developed at the Physical Sciences Building to handle these analyses.

A spot test using silver nitrate (Fiegl procedure) should be run first to determine if chloride concentration exceeds.08 ppm.

A negative result obviates the need for ion-chromatograph analysis.

Boron is run on the diluted liquid sample by the chlori-metric method ASTM (D3082C)31.

d)

In line monitoring capability as part of the sampling panel is used to determine pH and conductivity on the undiluted sample.

1. 3 (Later)

1. 4 Reactor coolant samples are depressurized in the sampling panel.

Any gases released plus gases stripped from the sample are col-lected and diluted to a known volume.

The diluted gas sample is then analyzed on a gas chromatograph where H, 0, N, and 2

2 2

CO are determined.

2

1. 5 Capability of performing chloride analysis by ion-chromatography is available at the Power Chemistry Laboratory in the Physical Sciences Building of the Training and Technology facility lo-cated in Huntersville, N.C.

A sample can be transported to this location and analyzed within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />.

However, as stated in 1.2c above, a preliminary spot test per-fromed where the sample is collected will determine whether the chloride concentration exceeds.08 ppm and if further analysis is necessary.

(Detection limit, 2 x 10 6/25000=0.8 x 10 3 ppm.

Before 1:1000 dilutions = 0.08 ppm).

1. 6 Radiation exposures are kept low through the use of distance and dilution of the sample.

The sample panels are remotely controlled taking advantage of distance and the shielding of the walls.

Liquid samples can be diluted 3000:1 and air sam-pies can be diluted 10,000:1.

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1. 7 Boron analysis is performed on the diluted sample collected by the sampling panel.

The amount of dilution is chosen to minimize radiation exposure.

For the postulated 10 Ci/g ex-t treme in the sample drawn, a 1:1000 dilution would provide protection below the 75 rem exposure to the analyst.

With the limit of the ASTM D3082C Boron Method of 0.1 i.023 ppm, the minimum detectable concentration in the liquid sample

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would be 100123 ppm @ 1:1000 dilution.

I 280-5a Rev. 6 New Page l

CNS 1.8 Not applicable.

1. 9 a)

The post-accident sampling panels provide the capability to promptly obtain a liquid sample and a gas sample under reactor accident conditions as described in Regulatory Guides 1.3 or 1.4 and has the capability to dilute samples within the shield for measurement in order to reduce per-sonnel exposure.

The size of the samples and the dilution allows station personnel to analyze any liquid or gas sample with the available counting room equipment.

b)

The health physics and chemistry laboratory facilities located in the Auxiliary Building provide the capability for prompt radioactivity spectrum analyses of noble gases, radiciodines, radiocesium, and non-volatile radionuclides.

Highly radioactive samples are prepared at a sample pre-paration laboratory provided with sample shielding and a ventilation system to control airborne radioactivity.

No difficulties are expected performing these analyses.

1.10 The attached report on Functional Testing of the Post Accident Liquid Sampling Panel performed in the laboratory provides data on the accuracy, range, and sensitivity attainable by an operator.

These are adequate to provide pertinent data for the radiological and chemical status of the systems sampled.

1.11 All internal components of the air and liquid system and the sample lines are purged before the diluted samples are re-trieved.

The small line size in conjunction with an orifice will restrict reactor coolant flow in the event of a rupture.

The size, length of line and number of bends have been kept to a minimum.

Each air panel is vented to the stack vent.

2.0 Using dist ace and dilution, the exposure levels are kept low.

l The air sau.ple is diluted 10,000:1 and the liquid samples are j

diluted 3,000:1.

During sampling, the panels are controlled remotely.

3.0 Regulatory Guide 1.97 Revision 2 is under evaluation and a discussion of conformance will be provided in FSAR Section 1.8.

4.0 To comply with the requirements of NUREG-0737 Item II.B.3, Part 4 the Post Accident Sampling Panels (samplers) are powered from 240/120 VAC auxiliary control power system.

This assures that all components associated with post ac-cident sampling are capable of being operated within 30 l

minutes of an accident in which there is core degradation, l

and loss of offsite power assumed.

Detailed description of l

the 240/120 VAC auxiliary power system is presented in FSAR Section 8.3.2.1.1.2.

280-Sb Rev. 6 New Page l

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j CNS 5.0 (Later) 6.0 Procedure AP/0/A/5500/31, " Estimate of Failed Fuel Based on I-131 Concentreation" is being prepared and will be available six months prior to fuel load.

7.0 Not Used 8.0 The Nuclear Sampling System described in FSAR Section 9.3.2.2.1 provides the capability to measure reactor coolant dissolved oxygen at levels below 0.1 ppm if reactor coolant chlorides are determined to be > 0.15 ppm.

9.0 Upon finalization of the design of the Post' Accident Sampling System, procedures will be written to operate the system, de-termine sampling frequency and establish operator training re-quirements.

These procedures will be available six months prior to fuel loading.

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280-5c Rev. 6 i

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