ML20212N662

Technical Evaluation Rept for Proposed Change to Operation of Ginna Fuel Pool Charcoal Filter Sys, Informal Rept

ML20212N662
Person / Time
Site:	Ginna
Issue date:	08/31/1985
From:	Duce S, Mandler J IDAHO NATIONAL ENGINEERING & ENVIRONMENTAL LABORATORY
To:	NRC
Shared Package
ML20212N660	List: ML20212N657 ML20212N662
References
CON-FIN-D-5023 EGG-PHY-7350, TAC-59801, TAC-60036, NUDOCS 8608290088
Download: ML20212N662 (12)

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1 INFORMAL REPORT p-e y c:- i TECHNICAL EVALUATION REPORT FOR PROPOSED

/daho CHANGE TO OPERATION OF THE GINNA FUEL POOL U ~ N8flOD88-

. i CHARC0AL FILTER SYSTEM Engineering ;y ~

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OlSCLAIMER l

i This book was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Governmer.t nor any agency thereof,

! nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product or process disclosed, or represents that its use would I

not infnnge privately owned r,yhts. References herein to any specific comrnercial l

l product, process, or service by trade name, trademark, manufacturer, or otherwise,

• does not necessar'ly constitute or imply its endorsement, recommendation, or favoring by the United States Government or any agency thereof. The views and opinions of authors expressed herein do not necessanly state or reflect those of the United States

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Government or ar.y agency thereof. ,

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EGG-PHY-7350 i

TECHNICAL EVALUATI,0N REPORT for i

! PROPOSED CHANGE TO OPERATION OF THE GINNA FUEL POOL CHARC0AL FILTER SYSTEM DOCKET NO. 50-244 TAC NO. 59801 3 INEL Reviewers - S. W. Duce /J. W. Mandler

, NRC LEAD Reviewer - M. Fairtile l INEL PROGRAM Manager - C. L. Nalezny NRC FSV Project MANAGER - M. Fairtile

. NRC Program Manager - M. Carrington n

i Idaho National Engineering Laboratory EG&G Idaho, Inc.

August 1986 EG&G Idahr , Inc.

Idaho Falls, Idal,o 83415 i

i Prepared for the

- U.S. Nuclear Regulatory Commission Washington, D.C. 20555 l Under DOE ~ Contract No. DE-AC07-76ID01570 l

.- FIN No. D5023 i

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l ABSTRACT This report addresses EG&G Idaho's technical evaluation of proposed changes to Technical Specification 3.11 and 4.11 from Rochester Gas &

O. Electric Corporations R. E. Ginna Nuclear Power Plant. The specification change proposed not using the spent fuel pit charcoal adsorber system whenever irradiated fuel, moved in the fuel pit, was older than 60 days post-irradiation The conclusion is that the Technical Specification change would not involve a significant change in the type or magnitude of effluent releases from the fuel storage area at the R. E. Ginna facility.

'In addition the change involves no significant hazards considerations for the purposes of 10 CFR 100.

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SUMMARY

On October 9,1985 Rochester Gas and Electric, licensee for the R. E.

Ginna Nuclear Plant, submitted a request for license amendment. This request addressed the frequency of use of the spent fuel pit charcoal adsorber system. This report discusses EG&G Idaho's evaluation of the propesal. The proposal was evaluated against applicable regulations (i.e., 10 CFR 100, 10 CFR 20, 10 CFR 50 Appendix I) as to the effect that the acceptance of the proposed amendment would have on the, licensees ability to meet the requirements of the regulations. Facility supplied information, FSAR drawings, and telephone conversations were used in the evaluation. Dose calculations were also performed,. The conclusion of EG&G Idaho's technical evaluation is that the Technical Specification change would not involve a significant change in the type or magnitude of effluent releases from the fuel storage area and that the change involves no significant hazards considerations for the purposes of 10 CFR 100.

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CONTENTS ..

Abetract. . . . . . . . . . . . . .,. . . . . . . . . . . . . ... . . . i S u ma ry . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i i Introduction. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

. Background. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . -1 0

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Technical Evaluation. . . . . . . . . . . . . . . . . . . . . . . . . . 2 Conclusions . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 1

I References. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 e

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1 TECHNICAL EVALUATION REPORT FOR PROPOSED CHANGE TO OPERATION OF THE GINNA

. FUEL POOL CHARCOAL FILTER SYSTEM

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.. INTRODUCTION

.The purpose of this Technical Evaluation Report is to review and evaluate the proposed change to the Rochester Gas & Electric Corporation (RG&E) Ginna Station Technical Specifications 3.11 and 4.11[1] which raquire the use of the spent fuel pit charcoal adsorber system whenever irradiated fuel is being moved in the spent fuel storage pool.

l The evaluation considered the proposed change in view of the l requirements of 10 CFR 100.11, 10 CFR 20.106, 10 CFR 20 Appendix B l Table II,10 CFR 50 Appendix I, and the model Radiological Effluent

Technical Specifications NUREG-0472.[2] Other documents used in the evaluation were a letter from D. Crutchfield (NRC) to J. Maier (RG&E) dated Sept. 24,1981;[33 a letter from D. Crutchfield (NRC) to J. Maier 1- (RG&E) dated March 3,1982;[4] a letter from G. Lear (NRC) to Sholly Coordinator dated March 31,1986;[5] RG&E's R. E. Ginna Nuclear Power

- Plant Updated FSAR Figure 9.4-2 Heating, Ventilating, and Air Conditioning

,- System - Auxiliary and Intermediate Buildings (supplied by M. Fairtile NRC); and RG&E's submittal for the proposed Technical Specification change. In addition telecons were held with license personnel, the NRC Project Manager, and EG&G Idaho personnel on July 9. 1986 and July 23,

! 1986.

4 I BACKGROUND The 2xisting Technical Specifications 3.11 and 4.11 require the use of j the spent fuel pit charcoal adsorber system whenever an irradiated fuel i assembly is being moved in the spent fuel storage pool. The objective of

! the specification is to limit radioiodine releases and subsequent doses in

! ,, the event an irradiated fuel assembly is significantly damaged while being moved in the spent fuel storage pool.

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The proposed change to Technical Specifications 3.11 and 4.11 would 4

change the required use of the spent fuel pit charcoal adsorber ta tho::e periods when the irradiated fuel being moved has decayed less than 60 days

~. since irradiation. A 10 CFR 50.91 review was performed by the NRC. Using 4 ..

the thyroid dose criteria of 10 CFR 100 for an accidental release, the NRC I performed a calculation to determine what the thyroid dose of'an individual would be in the event a 60 day post-irradiation fuel element failed and released the iodine inventory to the fuel pit. In a telecon facility personnel claimed that the calculation took no credit for any charcoal adsorber system, with the resultant dose being 0.2 rem. The NRC finding, stated in the G. Lear March 31, 1986 letter, was that the i proposed application for amendment involved no significant hazards f consideration.

Neither the NRC nor Franklin Research personnel in their review of the Radiological Effluent Technical Specifications (NUREG-0472) considered the spent fuel pit charcoal adsorber system as an effluent treatment system.

Instead, this charcoal adsorber system was considered as an internal

.' cleanup system.

TECHNICAL EVALUATION ,

j. The evaluation of the proposed amendment reviewed the need for the spent fuel pit charcoal adsorber system to be in use to meet the.

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requirements of 10 CFR 100, 10 CFR 20, 10 CFR 50 Appendix I, and

. NUREG-0472 for gaseous effluent treatment systems. In addition, an evaluation was made as to the need for this system to be in operation during.a' Three Mile Island type accident.

Physically the spent fuel pit charcoal adsorber system is located just upstream of the Auxiliary Building IC exhaust fan in the spent fuel storage pit area ventilation system. Its function is to remove radioiodine that may be released during irradiated fuel movement. A HEpA

] ,, filter system exists downstream of the exhaust fan and acts as the gaseous

{ effluent treatment system. ,

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s A dose calculation was made to determine what the 10 CFR 100 accident dose would be to a member of the public in the event a 60 day post-irradiation fuel bundle was damaged while being moved in the fuel storage pit. The assumptions used for this calculation and the intermediate calculations are shown in Table 1. The resultant infant thyroid dose due to a 2 hr release from the fuel storage pool was 2.09 rem. This value is in good agreement with a decay corrected dose of 1.94 rem for an unfiltered release calculated using the values reported in the Sept. 24, 1981 letter to J. Maier (RG&E). These values are larger than the 1

< 0.2 rem value calculated by the NRC and reported in the G. Lear March 31, 1986 letter. However, tbare is a question as to whether credit was taken for the charcoal filter in this calculation. If a charcoal 4

filter is asst.med in our calculations the resultant values of 0.209 and 0.194 rem,'are in good agreement with the < 0.2 rem value. Regardless of which value is correct, 2.09 rem, 1.94 rem, or < 0.2 rem, the result is 4

still the same; the resultant dose is not a significant fraction of the thyroid dose allowed by 10 CFR 100, t

Review of source term releases from other commercial nuclear generating facility fuel pool areas (i.e., NUREG/CR-2346, NUREG/CR-0715, NUREG/CR-1629, and NUREG/CR-1992) indicate that there are differences, but i

no more than approximately a factor of two between nonrefueling and refueling periods for unfiltered iodine release rates. Also the l pre-filtered release rates from fuel handling areas are typically less than three percent of the total plant release rate. This shows that l operations in the fuel storage pit can increase the iodine release rate, i

but by no more than a factor of two under normal conditions, and that l release rates from fuel handling areas are typically not a significant-fraction of total facility I-131 release rates.

Therefore, in facilities that have charcoal adsorber systems downstream of the fuel pit ventilation

system the factor of two increase poses no significant problem as to

i. meeting the 10 CFR 20 and 10 CPR 50 Appendix I requirenents. However, in

!> the Ginna system, where the charcoal adsorber system has always been used

.. when moving fuel in the fuel storage pit, not using this charcoal adsorber when moving fuel that is 60 days post-irradiation will probably cause i

facility iodine releases to increase. The magnitude of the increase is

unknown but, based on the source term information cited above, it will likely be within the factor of two observed at other facilities, and will probably not cause the total facility release rate to approach limits.

As the spent fuel pit charcoal adsorber system is used only as an intermediate cleanup system and is not required as the final gaseous effluent cleanup system, the requirements of the G!nna Technical Specifications do not directly apply. However, if it were deemed ,

necessary to have the system operating in the event normal gaseous

• adioiodine release rates were approaching release limits during periods when fuel older than 60 days po.. -irridation is being moved in the fuel pit, the system could be put in operation before limits are exceeded. In relation to the necessity to have the charcoal system operable during a TMI type accident, it would be helpful, but not required to have the system operating continuously. If it were deemed necessary to have the fuel pit charcoal adsorber system operating it can be put on line from the control room.

- CONCLUSION Taking into account the above discussion, a determination has been made that the spent fuel pit charcoal adsorber system is not required to be used wher irradiated fuel, which has an irradiation date that is 60 days or older, is being moved in the spent fuel storage pool. This change would not involve a significant change in the type or magnitude of e'/ fluent releases from the fuel storage area at the R. E. Ginna facility.

In addition the change involves no significant hazards consideration for the purposes of 10 CFR 10r.

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TA,BLE 1. 10 CFR 100 Accidental Dose Assumptions and Calculations Assumptions:

A) Iodine partition factor (water to air) 0.007 B) Iodine release from bundle to pool 0.1 C) Bundle inventory 4.23 E5 Ci at end of irradiation. Value derived from TMI Unit.2 core inventory ann core loading.

D) Facility vent flow 71,282 cfm (plant value)

E) Breathing rate 1400 m3 /y infant and 8000 m3 /y adult F) Thyroid mass 2 g infant 16 g adult G) Duration of release 2 h H) X/Q = 4.8 E-4 s/m3 I) All iodine entering the body via inhalation path goes to thyroid J) Infant thyroid gamma dose is <15 % of beta dose (prior information)

I-131 Released:

4.23 E5 Ci

• 1 E6 uCi/Ci

• 0.1

• 0.007 = 2.96 E8 uCi Dilution Air:

71282 cfm

• 60 min /h

• 2 h *28320 cc/ft3 /1E6 cc/m3 = 2.422 E5 m3 Stack Average Concentration:

2.96 E8 uCi / 2.422 E5 m3 = 1.22 E3 uCi/m3 Fenceline Concentration:

1.22 E3 uti/m3

• 33.6 m3/s

• 4.8 E-4 s/m3 = 19.7 uCi/m3 Thyroid Concentration:

Infant 19.7 uCi/m 3

• 1400 m 3

/y

• 2 h / 365 d/y

• 24 h/d

• 2g = 3.17 uCi/g Adult 19.7 uCi/m 3

• 8000 m 3

/y

• 2 h / 365 d/y

• 24 h/d

• 16g = 2.25 uCi/g Infant Beta Dose (rem):[6]

73.8*0.180 Mev*3.17 uCi/g*7.F1 d*((1-exp(-0.693*80 d/7.61 d)) = 319 Infant Beta Dose From 60 Day Old Bundle:

319

• exp( .693

• 60 d / 8.05 d) = 1.82 rem

. Total Dose:

1.82 + (1.82 * .15) = 2.09 rem 5

, REFERENCES ,

, 1. Letter from R. W. Kober, Rochester Gas and Electric, to H. R. Denton, NRC,

Subject:

Rochester Gas and Electric Corporation, R. E. Ginna

.. Nuclear Power Plant Docket No. 50-244, dated October 9, 1985.

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2. United States Nuclear Regulatory Commission, Standard Radiological Effluent Technical Specifications for pressurized Water Reactors, NUREG ll 0472, Rev. 3, Draft 7", September 1982.

3. Lett:er from D. M. Crutchfield, NRC, to J. E. Maier, Rochester Gas &

l Electric Corporation,

Subject:

SEP Topics, XV-2, Spectrum of Steam

System Piping Failures Inside and Outside Containment XV-12, Spectrum of i Rod Ejection Accidents, XV-16, Radiological Consequences of Failure of Small Lines Carrying Primary Coolant Outside Containment, XV-17, Steam ,

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! Generator Tube Failure, and XV-20, Radiological Consequences of Fuel Damaging Accidents - R. E. Ginna, dated September 24, 1981.

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. 4. Letter from D. P. Crutchfield, NRC, to J. E. Maier, Rochester Gas &

Electric Corporation,

Subject:

Ginna Nuclear Power Plart Final Safety Evaluation Report for SEP Spics XV-2, Spectrum of Steam Iystem Piping j Failures; XV-12, Rod Ejection 8.ccid6nts; XV-16, Failure of Small Lines

! Carrying Primary Coolant Outside Containment; XV-17, Steam Generator Tube l

Failure; and XV-20, Fuel Oaaaging Accident Inside and Outside Contair, ment, dated March 3,1982.

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5. Letter from G. E. Lear, NRC, to Sholly Coordinator,

Subject:

Request for Publication in Bi-Weekly FR Notice - Notice of Consideration of Issuance t

of Amende:r.t to Facility Operating License and Proposed No Significant Hazards Consideration Determination and Opportunity for a Hearing (TAC No. 60036), dated March 31, 1986.

E 6. R. E. Lapp & H. L.'Andrews, Nuclear Radiation Physics Fourth Edition, ,

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p 274, Prentice Hall Inc., Englewood Cliffs, New Jersey, 1972.

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