ML20041A315

Technical Evaluation of Licensee Response to NUREG-0737 Item II.B.2-Design Review of Plant Shielding.

ML20041A315
Person / Time
Site:	Surry
Issue date:	12/31/1981
From:	Mccracken R EG&G, INC.
To:	Donohew J Office of Nuclear Reactor Regulation
References
CON-FIN-A-6427, RTR-NUREG-0737, RTR-NUREG-737, TASK-2.B.2, TASK-TM , EGG-PHYS-5689, NUDOCS 8202190416
Download: ML20041A315 (10)

Text

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hEGcG...n FORM EG&G-394

m. ii .n INTERIM REPORT Accession No.

Report No. EGG-PHYS-5689 Contract Program or Project

Title:

Operating Reactors -- TMI Lessons Learned, NUREG-0737 Response Evaluation Subject of this Documents Evaluation of Licensee Response to NUREG-0737 Item II.B.2 - Design Review of Plant Shielding Type of Document:

Technical Evaluation Report Author (s):

R. T. McCracken Date of Document:

December 1981 Responsible NRC/ DOE Individual and NRCIDOE Office or Division:

J. N. Donohew, Division of Licensing 1,

This document was prepared primarily for preliminary or internal use. it has n]t received full review and approval. Since there may be substantive changes, this document should not be considered final.

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EG&G Idaho, Inc.

Idaho Falls, Idaho 83415 l

Prepared for the U.S. Nuclear Regulatory Commission Washington, D.C.

Under DOE Contract No. DE-AC07 761D01570 NRC FIN No. A6427 i .

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[dd TECHNICAL EVALUATION OF RESPONSE TO ITEM II.B.2 0F NUREG-0737 ff DESIGN REVIEW 0F PLANT SHIELDING N t/ d.

SURRY POWER STATION UNITS 1 AND 2 - DOCKET NOS. 50-280/281 MN R. T. McCracken PRELIMINARY U.S. Department of Energy ,9 e Idaho Operations Office

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NI TECHNICAL EVALUATION REPORT NUREG-0737 ITEM II.B.2 - DESIGN REVIEW 0F PLANT SHIELDING SURRY POWER STATION UNITS 1 AND 2 Docket flos. 50-280/281 December 1981

! R. T. McCracken Reactor Physics Branch Physics Division EG&G Idaho, Inc.

ABSTRACT The Nuclear Regulatory Commission has required all licensees to perform a design review of plant shielding and to provide for adequate personnel access to those areas requiring occupancy in the mitigation of or recovery from an accident. This report contains an evaluation of the submittals for Surry Power Station Units 1 and 2.

FOREWORD This report.is supplied as part of the TMI Lessons Learned NUREG-0737 Response Evaluation program being conducted for the U. S. Nuclear Regulatory Commission, Office of Nuclear Reactor Regulation, Division of Licensing, by EG&G Idaho, Inc.

The U. S. Nuclear Regulatory Commission funded the work under the authorization B&R 20-19-01-06, FIN No. A6427.

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, CONTENTS 1

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1. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

2. EVALUATION CRITERIA . . . . . . . . . . . . . . . . . . . . . . . . 1

3. DISCUSSION AND EVALUATION . . . . . . . . . . . . . . . . . . . . . 2 3.1 Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.2 Systems Containing Scurces . . . . . . . . . . . . . . . . . . 2 3.3 Vital Access Areas . . . . . t . . . . . . . . . . . . . . . . 3 3.4 Projected Doses and Dose Rates,. . . . . , . . . . . . . . . . 3 3.5 Description of Modifications . . . . . . . . . . . . . . . . . 3

4. CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . .

4 l S. REFERENCES. . . . . . . . .~. . . . . . . . . . . . . . . . . . . . 6 t

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L TECHNICAL EVALUATION REPORT NUREG-0737 ITEM II.B.2 - DESIGN REVIEW 0F PLANT SHIELDING SURRY POWER STATION UNITS 1 AND 2

1. INTRODUCTION I 1

Evaluations of the accident at Three Mile Island Unit 2 have led to various recomendations resulting in a number of new requirements for operating reactors. These new requirements are described in NUREG-0E60,

" Nuclear Regulatory Commission (NRC) Action Plan Developed as a Res*11t of the TMI-2 Accident", dated May 1980, and NUREG-0737, " Clarification of TMI Action Plan Requirements", dated November 1980.

NUREG-0737 Item II.B.2 directed all licensees to perform a design review of plant shielding and to provide for adequate access to vital areas. This report provides an independent technical evaluation of the licensee response to this requirement.

2. EVALUATION CRITERIA The requirements for the plant shielding design review are contained in the position statement of NUREG-0737 Item II.B.2; the clarification statement provides additional guidance. The criterion for the evaluation of the design review is that the licensee shall have satisfied these requirements. The evaluation of licensee response consisted of verification that:

1. The specified radioactive source terms have been used.

2. The specified " systems assumed to contain high levels of radioactivity in a post-accident situation" have been con-sidered.

3. The specified " areas where access is considered vital under post-accident conditions" have been considered.

4. The doses to personnel shall not exceed the guidelines of General Design Criterion 19.

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3. DISCUSSION AND EVALUATION The Virginia Electric and Power Company response (References 1-2) to -

specific NUREG-0737 documentation requirements is summarized below.

3.1 Source Terms The source terms used in the evaluation were based on 0% noble gases, 50% halogens, and 1% solid fission products for sump water; 100% noble

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gases, 50% halogens, and 1% solid fission products (undiluted) for primary coolant samples; and 100% noble gases, 25% halogens, and 0% solid fission products for the containment atmosphere. Estimates of resulting dose rates were made at 0,1, 8 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> following the accident.

3.2 Systems Containing Sources The systems considered as sources for the shielding review included the high head safety injection portions of the chemical and volume control -

system and safety injection systems, the low head safety injection system, the recirculation spray system, the sample systems, and the containment atmosphere cleanup (hydrogen recombiner) system. The containment atmos-phere source was also considered.

The letdown portion of the chemical and volume control system was not considered as a_ sysum containing sources because it is isolated at the in-itiation of an accident and remains isolated from containment. The resid-ual heat. removal system was not considered because all of the piping for 4

this system is located inside the containment. The gaseous and liquid

.{ radwaste systems are nonessential for accident recovery and were not con-sidered as sources for the shielding revica Any consideration of the stanct.y pc - 2 st.nent system, or its equiv-alent, was not specifically documented.

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l 3.3 Vital Access Areas l The licensee identified the control room, the technical support cen-ter, the counting lab / health physics area, the operational support center, and the security control center as vital areas requiring continuous oc-cupancy. The liensee did not clearly identify vital areas requiring in-frequent access. ~ 91deration of the chemical analysis lab, radwaste panel, motor control centers, and vital instrumentation locations was not explicitly documented. Modifications to the sample systems were described, implying consideration of the post-accident sampling facility as a vital area; however, this consideration was not specifically documented.

3.4 Projected Doses and Dose Rates Estimated dose rates at 0, 1, 8 and 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the accident were provided for the continuous occupancy areas considered by the licensee. Of the continuous occupancy areas considered, only the control room will meet the continuous occupancy dose rate criterion of NUREG-0737. The other areas will require some modification to meet the NUREG-0737 requirement.

I No estimates of dose rates, occupancy times, or predicted personnel ex-posures were provided for any vital areas of infrequent access. The licensee stated that radiation zone maps had been developed and would be revised after all modifications are installed. Radiation zone maps were not provided with References 1 and 2. The statements made by the licensee imply that radiation exposure of personnel during post-accident activities will be with-in limits specified by NUREG-0737; however, these statements are not conclu-sive.

3.5 Description of Modifications The licensee identified the following shielding and plant modifications to reduce personnel exposure as required by NUREG-0737:

, (a) Automatic temperature control valves will be added to the service water lines in the charging pump cooling water subsystem.

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(b) Manual valves, located in high radiation zones, which must be operated to line up and operate the post-accident hydrogen analyzer will be replaced with re-motely operated valves. .

(c) Modifications to allow for post-accident sampling of reactor coolant and containment atmosphere include pos-sible additional shielding for the post-accident sampling facility, shielding portions of the lines added as part of the post-accident sampling facility, and interim procedural changes and temporary shielding for the ex-isting sampling facility.

(d) Additional shielding, area relocation, or procedural modifications are being evaluated to limit radiation dose rates in the technical support center, the oper-ational support center, the counting lab, and the security control center.

(e) Modifications to permit interfacing with external pro-cess sytems for cleanup after an accident will be made.

(f) The drain system for the auxiliary building and safe-guards building sumps will be modified to allow pumping of these sumps to containment rather than the liquid waste tanks.

4. Conclusions Virginia Electric and Power Company's response to NUREG-0737, Re-vision 0(II. dated December 10, 1980, and Revision 1(2) , dated May 31, 1981 provide information relevant to Section II.B.2. for post-accident access to vital areas. The licensee completed the required review of post-accident accessibility. Post-accident radiation sources used in calculations were specified, including time after shutdown for source .

terms. Systems which were assumed to contain high levels of radio-activity in post-accident situations were specified, and systems not 4

r needed in post-accident operations were specified. Continuous occupancy vital areas were identified and estimates of dose rates in these areas were

, provided. The licensee did not document consideration of vital areas of infrequent access specified in NUREG-0737. Estimates of dose rates were not provided for vital areas of infrequent access. projected doses to personnel were not provided. Final radiation zone maps will be prepared after completion of modifications. Radiation zone maps were not provided with response documents. Corrective design modifications necessary to comply with NUREG-0737 requirements were described.

This submittal will be found acceptable subject to satisfactory re-solution of the following open items:

1. The licensee should explain why a standby gas treatment system (or an equivalent system) was not included among the systems to be used after an accident. If the plant is not equipped with a SGTS the licensee should so state.

2. The licensee should explain why the chemical analysis laboratory, radwaste panel, motor control centers, and vital instrument locations were not considered vital areas of infrequent access. If controls and readouts for vital instruments are located in the control room, the licensee should so state.

3. The licensee should specifically document consideration of the post-accident sampling facility as a vital access area and provide estimated dose rates and doses to personnel while obtaining the required samples.

4. The licensee should provide detailed estimates of radiation doses to personnel during necessary activities in post-accident situations. The licensee should state that

. radiation exposures to personnel during post-accident activities will be within the limits of NUREG-0737 and GDC-19.

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5. The licensee should provide post-accident radiation zone maps with vital access areas identified. [

5. References

1. Virginia Electric and Power Company Response to NUREG-0737 Post-TMI Requirements, Revision 0, (December 10, 1980) ,

2. Virginia Electric and Power Company Response to NUREG-0737 Post- .

TMI Requirements, Revision 1, (May 31, 1981) i e

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