ML20009G393

Final Rept on Design Review of Plant Shielding & Environ Qualification of Equipment for Spaces/Sys Outside Containment Which May Be Used in Post-Accident Operations.

ML20009G393
Person / Time
Site:	Hatch
Issue date:	07/24/1981
From:	GEORGIA POWER CO.
To:
Shared Package
ML20009G389	List: ML20009G388 ML20009G393
References
RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.2, TASK-TM TAC-47943, TAC-47944, NUDOCS 8108040204
Download: ML20009G393 (16)

Text

.

• S FINAL REPORT ON THE

DESIGN REVIEW OF PLANT SHIELDING AND ENVIRONMENTAL QUALIFICATION OF J

EQUIPMENT FOR SPACES / SYSTEMS

, OUTSIDE CONTAINMENT WHICH MAY BE USED IN POST-ACCIDENT OPERATIONS i

EDWIN I. HATCH NUCLEAR PLANT UNITS 1 AND 2 l

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July 24, 1981 8108040204 810727

PDR ADOCK 05000 P

I. INTRODUCTION Item II.B.2 of NUREG 0737 (previously section 2.1.6.b of NUREG 0578) requires that a comprehensive review of plant shielding be completed '

to assure that vital access areas within the plant are not rendered inaccess.ible due to high. dose rates and that essential equipment is not degraded such that it fails to serve its safety function. The initial review was completed and transmitted to seet the required date of January 1, 1980.

The subject report identified certain areas within both Hatch Unit 1 and Unit 2 which were potential problem areas. Table 1 in the original report identified each of the potential areas. The result of the reevaluation of each of these areas is presented below in the order which they were identified on Table 1 of the original report.

II. SUFF.ARY AND CONCLUSION This evaluation completes the requirement as specified in Item II.B.2.

of NUREG 0737. As explained in the original report a detailed component level review of all essential electrical equipment is being conducted under the auspices of I.E. Bulletin 79-01B and has not therefore been covered in depth within this report. If any deficiencies are identified under 79-01B they will be addressed .'n that report.

With the plant modifications outlined in this report, it has been determined that the areas addressed as potential problem areas in the original report are no longer considered problems. The plant modifications are designed and under revision. It is anticipated that the installation of shielding will be complete by January 1, 1982.

III. CONTAINED SOURCE TERMS The source terms which were used in the original evaluation were again used for the purposes of this evaluation. The following release

fractions were used as a basis for determining the concentrations for i

contained sources:

l Containment atmosphere: 100% noble gases, 25% halogens Suppression chamber liquid: 50% halogens, 1% solids

[ Reactor steam: 100% noble gases, 25% halogens The above release fractions were applied to the total curies available for I the particular chemical spec ies (i.e. noble gas, halogen, or solid) for an equilibrium fission product inventory for a light water reactor core.

The important modeling parameters, decay time and dilution volume obviously have an important affect on the shielding analyses. The following sections l outline the rational for the selection of values for these key parameters.

1 l

.. . = _ .

A. Decay Time For the purposes of developing a set of accident radiation zone maps no decay time was assumed. The zone maps (Figure 7 through Figure 14) are developed to be used as a tool by the plant staff along with a set of decay curves (Figures 5 and 6) to quantitatively assess the plant status quickly at any time following an ab' normal occurrence.

For the purposes of evaluating personnel exposure in vital areas of the plant decay times consistant with the time for which access is required were used.

B. Dilution Volume i

The volume used for dilution is important, affecting the calculations of dose rate in a linear fashion. The following dilution volumes were used with the release fractions and decay times indicated above to

, arrive at the final source terms for the review.

Con *ainment atuosphere: Drywell and suppression chamber free volume Suppression chamber liquid: The volume of the reactor coolant system plus the suppression chamber volume at its minimum allowable level.

. Reactor steam: Reactor coolant system normal vapor volisa.

IV. REACTOR BUILDING AIRBORNE SOURCE In addition to the doses which are derived from the above sources an airborne activity source in the reactor building was evaluated. The airborne activity will give both gamma and beta dose rates but only gamma is of major i concern since the-beta radiation can be effectively shielded with protective clothing. Thyroid dose due to iodine inhalation is ignored since personnel will be expected to wear air packs.

A. Assumptions used in the analysis.

Listed below are the assumptions used to determine the gamma dose l

rate from the airborne activity:

Core Activity - Based on TID 14844 with core power level of 2550 MWt. ,

I Core Release to Containment - 25% iodine 100% noble gases Containment Volume - 256,000 cubic feet l RB Volume @ 130' - 0" El. - 299,300 cubic feet l

Total - 1,056,000 cubic feet Containment Leak Rate - 1.2 v/o per day i

SGTS Exhaust Rate - two cases considered:

1) 3,000 cfm

2) 3,000 cfm (first 10 days); then 6,000 cfm thereafter I

j 2 l

Inherent in this model is the assumption of complete mixing within the RB volume of the radioactivity leaked from the containment. The RB volume used does not include any of the common refueling floor volume which results in conservative concentrations. Also assumed was the continuous containment leak rate of 1.2 v/o per day which is conservative for the first day and very conservative for subsequent days post-LOCA.

2. Pasults of the analysis The results of the analysis are presented in Table I.

TABLE I

{ Post-LOCA Airborne Gamma Dose Rates Time Post-LOCA Case A Case B (days) R/hr R/hr 5 34 34 10 18 18 20 5.9 2.9 30 2.0 1.0 V. RESULTS OF THE "; EVALUATION A. Sample cask retrieval - Unit 2

1. Description of the Problem The location of the sample cask for the liquid sample on Unit 2 was over the RHR corner room. The operator would have had to enter elevation 130' and disconnect the sample cask and remove it to the hot machine shop.

2. Results of the Evaluation It was determined that an operator would receive a dose of 100 mr, without consideration of the airborne source, while disconnecting the sample cask and removing it from the 130' elevation to the hot machine shop. Although this dose is concidered acceptable during the accident situation, the location of the sample cask will be changed due to the installation of the inline sample system. One inline sample system will be used for both Hatch Unit 1 and Unit 2. The sample cask will be located in the hot machine shop where, the radiation level does not l

consistute a personnel or equipment bazard under the worst postulated accident conditions.

B. Access to elevation 130 in the Reactor Building

1. Dcscription of the Problem l .

The original report indicated that general access to the 130' elevation of the Reactor Building in the area above the torus chamber was limited by the dose rate from the torus compartment which is shielded by a two (2) foot concrete floor slab. Access to specific areas was further limited by the core spray lines on both units which run through the 130' elevation with virtually no shielding. Subsequently the potential airborne l

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dose rate in the Reactor Building was found to be very high also.

2. Results of the evaluation In order to improve access and limit the radiation dore received by equipment in the area of the core spray lines on elevation 130', appropriate portions of the lines will be shielded with one inch of steel. The degree to which they are to be shielded has been determined such that thirty (30) days af ter the postulated accident the dose rate from all of the contributing sources will not preclude an operator from making an inspection tour or performing some minor maintenance such as the replacement of a combination starter.

Providing for access thirty (30) days af ter an accident has been based on the following: ,

a) Access to the Reactor Building after an accident is not required for proper ECCS actuation and operation.

b) The airborne source in the Reactor Building would be largely dissipated.

c) The installation of core spray line shields required for access in thirty (30) days will not require major structural modifications to the building.

d) The installation of massive shields which are required for access prior to thirty (30) days is not practical from a construction point of view.

C. High radiation doses to equipment in the corner rooms where the RHR and RCIC pumps are located and in the HPCI rooms in the Reactor Building.

1. Results of the evaluation.

As indicated above all essential equipment is being evaluated at the i component level under the auspices of I.E. Bulletin 79-01B. The

( '

equipment in the RER, RCIC and HPCI roogs will be exposgo to the maximum gotal integrated dose 1.86 x 10 rads, 2.4 x 10 rads and l 2.4 x 10 rads respectively. If during the evaluation conducted for I.E. Bulletin 79-01B a component is found to be deficient corrective action will be detailed within that report. There is no personnel access required to these rooms to mitigate any accident condition.

l l D. Access to +.he areas inside the southern portion of the railroad air lock ci.

Unit 1.

I

1. Results of the evaluation .

j The area outside the portion of the railroad airlock doors that are in a direct line of sight of the core spray line could be subjected to a high radiation field in the unlikely event of an accident. The station radiation operating procedures recognize this possiblity and require the area to be evaluated and access to be restricted as required.

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E. Access to the area outside the truck door on Unit 2.

1. Description of the problem.

The yard area outside the truck door on Hatch Unit 2 has been identified as a potential area of concern because of the possibility of' excessive exposure to plant personnel'in the yard area outside the door. This is normally an unlimited access area. The dose contribution would be primarily from the core spray line near the truck door.

2. Results of the evaluation One inch of steel will be installed to shield the vertical piping run from elevation 130'- 0" to 144' - 6 ". The shielding on the line will reduce the dose at the door to soma degree; however, as has been previously indicated it is not practical to shield the spray line to the extent required to eliminate the yard dose concern. The station radiation operating procedures recognize that during the early stages of an event the dose rate outside the door could be high. The procedures require that the area be surveyed and access restricted as required.

The Accident Zone Maps which were attached to the original report written for submission by January 1, 1980 have been modified to reflect the results of this study and assigned figure and revision numbers to assure traceability In addition, the decay curves for sources B (Figure 5 - Diluted Primary Coolant) and C (Figure 6 - Drywell Atmosphere) have been included to provide the information required to predict the dose rate in each area at any time after the event.

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