ML19257D816

Forwards Completion of Airborne Dose Assessment Required by NUREG-0578.Proposes Addl Corrective Action.Equipment Qualification Review Is Being Prepared in Response to IE Bulletin 79-01B.Drawings Available in Central Files Only

ML19257D816
Person / Time
Site:	Peach Bottom
Issue date:	01/31/1980
From:	Daltroff S PECO ENERGY CO., (FORMERLY PHILADELPHIA ELECTRIC
To:	Harold Denton Office of Nuclear Reactor Regulation
Shared Package
ML19257D817	List: ML19257D816
References
RTR-NUREG-0578, RTR-NUREG-578 IEB-79-01B, IEB-79-1B, NUDOCS 8002060533
Download: ML19257D816 (10)

Text

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PHILADELPHIA ELECTRIC COMPANY 2301 M ARKET STREET P.O. BOX 8699 PHILADELPHI A. PA.19101 SHIELDS L. OALTROFF vece Pas sscamt ELECf e'C PmODUCTION January 31, 1980 Re: Dochet Mos. 50-277 50-278

• ! r. Harold R.

Denton, Director Office of Huclear React or Re9ulation US Nuclear Regulatory Connission Nashington, DC 20555

SUBJECT:

Design Review Studies Required by Short Tern Lessons Learned Peach Sotton Atonic Power Station Reference-Letter dated January 2,

1930-3.

L.

Daltroff.

Ph ila de lp h ia Electric C onp any to iarold P..

Denton. Nuclear Regulatory Connissidn

Dear Mr. D en t on :

The above referenced letter transnitted the results of our olant shieldine desien review required by Short Tern Lessons Learned NUREG 0578. Section 2.1.6.b.

As stated in that transmittal, the results of the study uere connlete except for (1) an evaluation to determine the airborne dose in the area of the raduaste nanel, radiochenical laboratory and the control roon followin2 the accident described in SUREG 0579, and (2) the equionent qualification review.

The a t t a ch me n t to this let'ar conoletes the airborne dose assessnent required by UUREC 0573, and p rop os es additional corrective action.

In addition. the results of the reactor buildine dose assessment presented in the above referenced le tter has been re-evaluated based on additional information.

As we stated in the above referenced letter. our couianent qualification review is bein.2 done in resnonse to IZ Sulletin 79-01.

On January 14. 1930. IF Sulletin 79-013 uns issued ex,andin-the scene and n r ovi sin e additional cine fcr the licensees to

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h 1928 212 8002060 h

'f r. II.

R.

Denton, Director Page 2 review the environmental oualification of Class IE electrical equinnent.

Ile will submit a response to the NRC regarding the results of this study in accordance with the schedule identified in Bulletin 79-013.

Please contact us if further discussion of this matter is necessary.

Very truly yours,

/

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Attachment B

9 8

e 1928 213

Peach Bottcm Atcmic Power Station Supplemental Informtion for NUREG-0578 Section 2.1.6.b, Design Review of Plant Shielding m e results of further analysis of post-accident accessibility are discussed below in items A. 1, B.2 and B.3.

In addition, in the process of reviewing the calculations used to generate the results reported in our letter of January 2,1980 two errors were discovered: 1) an incorrect volume for one of the rooms in secondary containment was used and 2) an incorrect flow rate was used for the standby gas treatment system.

Thus, revised calculations were performed, and the results are also presented for a) airborne in secondary containment, item B.1, b) shine dose in the Technical Support Center, item A.2, c) shine dose at the Radwaste panel, item B.2, and d) shine dose in the radiochemistry laboratory, item D.3.

A.

Areas Recuiring Continuous Occupancy 1.

Control Rocm - The assumption of 1% containment leakage bypassing the standby gas treatment system has negligible effect on the whole bdy dose to personnel in the control room.

2.

Technical Support Center - he 180 day dose without shielding frcn direct shine frcm Unit 2 reactor building is 54.6 rem.

This value supercedes the value of 13.1 rem reported in Enclosure 5 of our letter dated Januar-f 2, 1980. W e addition of eight inches of concrete or equivalent shielding will reduce the dose from direct shine to less than 5 rem. The total whole bcdy dose for 180 days to personnel in the Technical Support Center will be less than 5 rem. We filter train, as described in our January 2,1980 letter remins unchanged.

B.

Areas Recuiring Infrecuent Access 1.

Secondary Containment - Figure 1 shows the dose rates versus time post-IOCA within secondary mntainment due to airborne radiation. These dose rates supercede the values reported in of our letter dated January 2, 1980. The airborne radiation is due to leakage frce primary containment at the rate of 0.50% per day. Entry intb secondary containment could be mde after several days for short periods of time, i.e. a few minutes; however, we do not believe entry into secondary containment would be required until recovery operations begin many days after the postulated accident.

Since secondary containment is inaccessible for several days, two modifications must be mde. The capability to obtain post-accident primry coolant and primary containment gas samples requires modifications to our sampling systems.

Mcdifications will be mde to provide post-accident sampling capability outside the reactor building. Conceptual design of 19mn20 nd 4

. these new sampling provisions has been completed. The generic approach to postaccident sampling develcped by General Electric Canpany for the LUR Owners' Group was used as the basis for our conceptual design. Liquid samples will be taken from reactor pressure vessel instrunent lines, FJR sample lines and torus water level instrument lines. Gaseous samples will be taken fran the drywell radioactive gas sample lines. All samples will be processed in shielded sampling stations located outside secondary containment or they will be transferred to the recote counting laboratory for analysis.

Modification is also required to the controls and instrunentation associated with the makeup water supply to the spent fuel pools. Madifications will be made to permit raintenance of pool water level from outside secondary containment.

Both modifications will be car 1pleted by January 1,1981, unless precluded by equipment unavailability.

Drawings M-13 through M-18 show equiprent 1ccations for various elevations for the Unit 2 reactor building and the carren radwaste building. 'Ihe numbers 1, 2, and 3 respectively refer to 1-the dose rate due to airborne in secondary containment, 2-contact dcse rates on EOCS piping, and 3-the contact dose rate on the standby gas treatment system filters. Figure 1 gives the dose rate due to airborne in secondary containment.

Figure 2 gives the contact dose rate for 24" diameter ECCS piping. 'Ihis is a bounding assumption: all ECCS piping was assumed to be 24".

Figure 3 gives the contact dose rate for the standby gas treatnent system citarcoal filter. 'Ihe affect of the shine from the filters on fan qualification will be addressed in our response to Bulletin 79-01B.

2.

Padaaste Panel - The total 180 day whole body dose to personnel at the control panel area in the Radwaste Building is 6.82 rem (6.77 from airborne and 0.05 fran torus shine). The value of the torus shine dose component supercedes the value presented in Enclosure 5 of our letter dated January 2, 1980. This assumes an occupancy factor of 1.

Occupancy factors less than 0.4 would be o p ted; thus the 180 day whole body dose would be less than 2.7 rem. hhole body and thyroid rates from airborne are shown in table 1.

Fresh air breathing apparatus would be required for extended stay periods in the radwaste building.

-3.

Padicchemistry Laboratory - The total 180 day whole body dose to personnel in the Radiochemistry Laboratory is 7.86 rem (0.48 rem from airborno and 7.38 rem from torus shine). The value of the torus shine dose component supercedes the value presented in Enclosure 5 of our letter dated January 2,1980.

This assumes an occupancy factor of 1.

Occupancy factors less than 0.4 would be expected; thus, the 180 day whole body dose would be less than 3.2 rem.

I928 215

. Although access to the radiochemistry laboratory would be possible, the dose rates, as shown in table 2, in the first few hours after an accident will adversely affect the accuracy of the counting equiprent.

Provisions will be rude to pennit the counting of samples in areas with lower levels of background.

We presently envision using the counting roan at Unit 1, where the Technical Support Center is located. Upgrading our capability to perfonn post-accident counting and analysis will be canpleted by January 1,1981, unless precluded by equipmnt unavailsbility..

9 8

1928 216

PEAQI BOT 104 AT004IC 10WER STATICN TABLE 1 DOSE RATES FID4 AIRBORNE Radwaste Building Radiochenistry Post-IOCA Control Panel Area Tahnratory Time Dose Rates (RDf/HR)

Dose Rates (RD4/HR)

(Hours) ihole Body

'Ihyroid thole Body Thyroid 1

4.56-l*

2.42+3 4.11-2 1.92+2 2

3.45-1 2.39+3 3.10-2 1.89+2 4

2.26-1 2.27+3 2.02-2 1.80+2 8

1.26-1 2.08+3 9.03-3 1.59+2 24 3.91-2 1.63+3 3.12-3 1.17+2 48 1.89-2 1.31+3 1.46-3 9.04+1 72 1.32-2 1.12+3 1.02-3 7.70+1 96 1.04-2 9.79+2 7.88-4 6.65+1 240 4.83-3 5.47+2 3.57-4 3.61+1 480 1.75-3 2.20+2 1.29-4 1.45+1 720 7.95-4 8.83+1 5.17-5 5.78 4,380 3.49-5 0

2.52-6 0

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• 4.56-1 = 4.56 x 10 Note: No m y ncy factors included 1928 217

PEAQI B0fl04 AKHIC POWER STATION TABLE 2 Total Dose Rates In Radiodwmtstry raMratory Post-IOCA Dose Rate Time (Hours)

(RADS /HR) 1 0.65 2

0.42 4

0.23 8

0.09 24 0.01

-4 96 8.8 x 10

-4 240 3.7 x 10

-5 720 5.5 x 10 e

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