ML20041D227

Evaluation of Licensee Response to NUREG-0737 ,Item II.B.2 - Design Review of Plant Shielding Edwin I. Hatch Nuclear Plant Units 1 & 2 - Dockets 50-321 & 50-366, Technical Evaluation Rept

ML20041D227
Person / Time
Site:	Hatch
Issue date:	01/31/1982
From:	Olson W 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-5714, NUDOCS 8203040594
Download: ML20041D227 (14)

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TECHNICAL EVALL't' ION OF RESPONSE TO ITEM II.B.2 0F NUREG-0737 48

/C 5i C-DESIGN REVIEW OF PLANT SHIELDING-EDWIN I HATCH NUCLEAR PLANT A T/6 7

UNITS 1 AND 2 - DOCKET NOS. 50-321, 50-366 >

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INTERIM REPORT Accession No.

Report No. EGG-PHYS-5714 Contract Program or Project

Title:

Operating Reactors -- TMI Lessons Learned. NilREG-0737 Response Evaluation Sub! vet of this Document:

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

Technical Evaluation Report Author (s):

W. O. Olson Date of Document:

January 1982 Responsible NRCIDOE Individual and NRC/ DOE Office or Division:

J. N. Donohew, Division of Licensing l

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

l EG&G Idaho. Inc.

Neo Falls. Idaho 83415

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

Under DOE Contract No. DE-AC07 761001570 NRC FIN No. A6427 INTERIM REPORT

TECHNICAL EVALUATION REPORT NUREG-0737 ! TEM II.B.2 - DESIGN REVIEW 0F PLANT SHIELDING EDWIN I HATCH NUCLEAR PLANT UNITS 1 AND 2 Docket Nos. 50-321, 50-366 January 1982 W. O. Olson Reactor Physics Branch Physics Division EG&G Idhho, Inc.

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r ABSTRACT The Nuclear Regulatory Commission has required all licensees to perfcrm a design review of plant shielding and to provide for adequate personnel estess to those areas requiring occupancy in the mit1gation of or recovery from an accident. This report contains an evaluation of the submittals for the Edwin I. Hatch Nuclear Plant 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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l CONTENTS

1. INTRODUCTION. . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 i .

2.

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

3. DISCUSSION AND EVALUATION . . . . . . . . . . . . . . . . . . . . . . 2 3.1 Source Terms . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3.2 Systems Containing Sources . . . . . . . . . . . . . . . . . . . 3 3.3 Vital Access Areas . . . . . . . . . . . . . . . . . . . . . . . 5 3.4 Projected Doses and Dose Rates . . . . . . . . . . . . . . , . . 6 3.5 Description of Modifications . . . . . . . . . . . . . . . . . . 7

4. CONCLUSIONS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

5. REFERENCES. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 10 i

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i TECHNICAL EVALUATION REPORT NUREG-0737 ITEM II.B.2 - DESIGN REVIEW 0F PLA:iT SHIELDING EDWIN I HATCH NUCLEAR PLANT UNITS 1 AND 2

1. INTRODUCTION Evaluations of the accident at Three Mile Island Unit 2 have led to various recommendations resulting in a number of new requirements for operating reactors. These new requirements are described in NUREG-0660,

" Nuclear Regulatory Commission (NRC) Action Plan Developed as a Result 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 t.3 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 cf 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.

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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 t

The Georgia Power Company response to NUREG-0737 is in two distinct segments; Reference 1. which is a shielding design review of the reactor building and Reference 3, which adds leakage from the containment to the .

reactor building and assumes no access to the reactor building for the first 30 days following an accident. Reference 2 is a statement of li-censee commitments regarding the implementation of NUREG-0737.

The Georgia Power Company response to specific NUREG-0737 documentation requirements is summarized below.

3.1 Source Terms Source terms that were used in the evaluation were:

1 Source A: Containment atmosphere - 100% of noble gases and 25% of halogens diluted in a volume equivalent to the drywell and suppression pool free volumes.

i Source B: Reactor liquid - 100% of noble gases, 50% of halogens, and 1% of solids diluted in the reactor coolant system normal liquid volume at operating temperature and pressure.

Source C: Suppression pool liquid - 50% of halogens and i

1% of solids diluted in the volume of the l reactor coolant system plus the suppression pool volume (Ref.1). The suppression pool volume is further defined in Reference 3 as the volume at its minimum allowable level. .

Source D: Reactor steam - 100% of noble gases and 25% of -

halogens diluted in the reactor coolant systen l

normal vapor volume.

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1 The time assumed for the source terms was immediately after the accident. The licensee provided gamma energy emission decay curves for adjusting these source terms and dose rates to later times.

The licensee states that "The above release fractions were applied to the total curies available for the particular chemical species (i.e.,

noble gas, halogen, or solid) for an equilibrium fission product inventory for a light water reactor core". The licensee did not further identify the core as being appropriate to the problem.

3.2 Systems Containing Sources Systems that were included as sources for the shielding design review (Reference 1) and the sources used in each:

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! Containment spray system - Source C Core spray system - Source C High pressure coolant injection (HPCI) system Liquid - Source C Steam - Source D  :

Reactor core isolation cooling (RCIC) system Liquid - Source C Steam - Source D Residual heat removal (RHR) system - Source C was used for the low pressure coolant injection mode.

! Sampling systems Containment air sample - Source A Reactor coolant sample - Source B Reactor water cleanup system - Source B

. Recombiner - Source A l

In defining the limits of the connected piping subject to contamination j in the systems listed above, normally shut valves were assumed to remain shut.

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1 The reactor water cleanup (RWCU) system was included only to the second containment isolation valve. Those portions of the standby gas treat-ment system up to the first closed isolation valve which could contain atmosphere from the containment were included for Hatch Unit 2 (as an extension of the containment atmosphere). ,

The containment atmosphere, as a system, was not documented as a sytem assumed to contain high levels of radioactivity following an accident. The licensee should state whether the atmosphere within the containment, as a system, was considered in the shielding evaluation.

If it was not, radiation zone maps and estimated doses to personnel must be revised to include this term.

Similarly, it was not documented whether liquid systems within the containment (drywell and torus) have been considered in estimating dose rates and estimated doses to personnel outside the containment. The i licensee should state whether these systems have been included, and if they have not, their contributions to dose rates and estimated doses to personnel should be determined.

It is noted that the radiation from the torus is discussed as a contributing factor in the high dose rate levels at the 130-foot level of the reactor building. The above two paragraphs request documentation that the liquid and gaseous systems within the drywell and torus have been included as sources for the shielding design review.

Reference 3 evaluates the effects of leakage of the containment at-mosphere into the reactor building. The containment atmosphere (25% of the core iodine and 100% of the core noble gases diluted in the 256,000

! cubic foot containment volume) was assumed to leak into the portion of the reactor building below the refueling floor (1,056,000 ft3 ) at a leak .

rate of 1.2 volume percent per day. The air in the reactor building was assumed to be exhausted through the SGTS with two rates considered, either -

at 3000 cfm, or at 3000 cfm for the first 10 days and 6000 cfm thereafter.

The resulting reactor building atmosphere was used as a system containing a radioactive source.

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I With this change in treatment (consideration of the reactor building atmosphere as a system containing radioactivity) the licensee should doc-ument that the complete SGTS has been considered as a system containing high levels of radioactivity following an accident.

Also in Reference 3, the licensee describes a mcdification which moves the high level sampling functions from the 130' elevation of the reactor building to the hot machine shop. While Reference 1 documents consideration of the sampling system at the original location, the licensee should document that the new sampling system has been considered as a system centaining high levels of radioactivity and thus include it in the shielding review.

3.3 Vital Access Areas In Reference 1 the licensee considers as vital areas requiring access:

Control Room On-site Technical Support Center (TSC)

On-site Operation Support Center (OSC)

! Sampling and monitor areas l

Areas which are listed in NUREG-0737 " Clarifications" which were to be considered for inclusion as vital areas, but were rct included in the above list are:

Chemical Analysis Laboratory ,

Post-LOCA Hydrogen Control System Motor Control Centers

! Instrumentation Locations i Radwaste Control Panels

. Emergency Power Supplies If these areas were not considered vital areas for post-accident access, the licensee should so state and explain why. If access to these areas is 5

considered vital following an accident, dose rate determinations and es-timates of doses to personnel for necessary access should be made.

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3.4 Projected Doses and Dose Rates The licensee has stated that the dose rates are less than 15 mR/hr in the control room and technical support center at all times following an accident.

The licensee has provided dose rate maps for the reactor building before and after shielding of portions of the core spray lines with one inch of steel. These maps show dose rates immediately following an ac-cident. The contributions due to leakage of the containment atmosphere into the reactor building (at various times) are calculated to be added separately. Decay curves for the gamma energy emission rate are included for sources A, B, and C of Section 3.1 (with source D to be decayed using the curve for source A) to correct the dose rate maps to later times. Dose rate map regions are designated as A-I (<15 mR/hr), A-II (15-100 mR/hr),

A-III (100 mR/hr-5 R/hr), and succeeding steps by order-of-magnitude increments.

The licensee has not provided dose rate maps or estimated doses to individuals for areas other than the reactor building, control room, and technical support center (as detailed above). The licensee should provide dose rate maps for other potentially occupied areas and should provide estimated doses to individuals for necessary occupancy times in vital areas.

Additionally, the licensee should document the dose rate changes due to inclusion of the full SGTS, implied in Reference 3, which was not in-cluded in Refercnce 1. The licensee should also document dose rate changes and estimated doses to personnel resulting from using the high level sampling .

system introduced as a modification in Reference 3.

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I Reference 3 provides gamma energy emission decay curves for Sources ?

and C, but states that these apply to the diluted primary coolant and drywell atmosphere, respectively. This does not agree with Reference 1, where these curves are assigned to the reactor liquid and suppression 1

pool liquid, respectively.

l 3.5 Description of Modifications The licensee states that an in-line sampling system, to be used for both Units 1 and 2, will deliver samples to a sample cask in the hot ma-chine shop. No design details were provided for this system.

Appropriate portions of core spray lines will be shielded with 1 inch of steel to reduce dose rates at the 130-foot level of the reactor building, permitting brief entry to this level 30 days after the postulated accident.

High radiation levels outside the railroad air lock of Unit 1 and outside the truck door of Unit 2 will be reduced somewhat by 1 inch of steel shielding on portions of the core spray lines. Access to yard areas adjacent to these doors will also be restricted following an ac-

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cident. i l

Maximum total doses of 1.86 x 106 R, 2.4 x 106 R and 2.4 x 106 R have been calculated for the RHR, RCIC and HPCI rooms, respectively. Equipment is being evaluated for radiation resistance to these doses under I.E Bulletin 79-01B and corrective action taken will be reported there. The licensee states that no personnel access is required for accident mitigation.

4. CONCLUSIONS e A plant shielding design review was performed for Georgia Power Company's Edwin I. Hatch Nuclear Plant Units 1 and 2. Post-accident source 7

terms consistent with NUREG-0737 were used. The core used in developing the isotopic inventories was not specified. Systems which were assumed to contain high levels of radioactivity in post-accident situations were spec- '

ified. The licensee did not address all systems specified in NUREG-0737.

The licensee identified certain vital access areas. The licensee did not ,

document consideration of all vital access areas specified in NUREG-0737.

Estimated dose rates were provided for the control room and technical support center. Radiation zone maps immediately post-accident, with order-of-magnitude radiation dose rate level indicators were provided for the reactor building. Gamma energy emission decay curves were provided for adjusting the maps to other times. Estimated dose rates, occupancy times to perform tasks and doses to personnel were not provided for other areas requiring vital access. A summary description of modifications to be made was provided. The description of the high level sampling system was insuf-ficient for the purposes of this review.

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

1. The licensee should further identify the core (and its operating history) and methods used in the calculation of the isotopic inventories used for the source terms.

2. The licensee does not state that the containment atmosphere has been included as a system containing high levels of radioactivity following an accident. If it has been included in making dose rate estimates it should be so stated. If it has not been included, the dose rate estimates should be revised to include this term.

3. The licensee does not state that liquid systems within the -

containment have been included in making dose rate estimates for areas outside the containment. If they have been in-cluded it should be so stated. If they have not been included, the dose rate estimates should be revised to in-clude this source term.

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It is noted that radiation from the torus has been discussed as a contributing factor to the high dose rate levels at the 130-foot level in the reactor building. The above two paragraphs request specific documentation of the inclusion of the atmos-phere and liquid systems within the drywell and torus in the shielding evaluation.

4. In the changes that were assumed from Reference 1 to Reference 3, the full standby gas treatment system comes into use and ac-cumulates high levels of radioactivity. The licensee should document the inclusion of the full SGTS in the' shielding de-sign review.

5. The licensee has comitted to install a new in-line sampling system to deliver samples to the hot machine shop. The licensee should document inclusion of this system in the shielding de-sign review and provide dose rate maps and estimated doses to personnel to obtain the required reactor coolant and contain-ment atmosphere samples under accident conditions.

6. The licensee should explain why certain l'reas, listed in NUREG-0737 as potentially vital areas, were not included in the list of areas considered, namely; chemical analysis laboratory, post-LOCA hydrogen control center, motor control centers, instrumenta-tion locations, radwaste control panels, and emergency power supplies. If these are not considered vital areas the licensee should so state and explain why. If these are considered vital, dose rate estimates and estimated doses to personnel resulting from vital access should be supplied.

. 7. The licensee has provided estimated dose rates for the control room and technical support center and dose rate maps for the reactor building. The licensee should provide dose rate maps for other potentially occupied areas and should provide es-timated doses to personnel to perform necessary tasks. The licensee should confirm that these doses and dose rates meet the requirements of NUREG-0737 and GDC-19.

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8. The licensee provides gamma energy emission decay curves for sources B and C in Reference 3, stating that these apply to the diluted primary coolant and drywell atmosphere, re- ,

spectively. This does not agree with Reference 1, where these curves are assigned to the reactor liquid and suppression pool ,

liquid, respectively. The licensee should clarify, 5 REFERENCES

1. " Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces / Systems Outside Containment Which May Be Used in Post-Accident Operations - Edwin I. Hatch Nuclear Plant Units 1 and 2",

January 25, 1980.

2. Letter, Georgia Power Company (W. A. Widner) to NRC (Director of Nuclear Reactor Regulation), December 15, 1980 - with enclosure-

" Intentions of Compliance for NUREG-0737 Requirements".

3 Letter, Georgia Power Company (J. T. Beckham, Jr.) to NRC (Director of Nuclear Reactor Regulation ), July 27, 1981 - with enclosure-

" Final Report on the Design Review of Plant Shielding and Environmental Qualification of Equipment for Spaces /Systeins Outside Containment Which May Be Used in Post-Accident Operations - Edwin I. Hatch Nuclear Plant Units 1 and 2", dated July 24, 1981.

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