ML20153D150

Affidavit of ML Wohl Rejecting Util 860123 Motion for Summary Disposition of Contention 4 Re Radiological Analysis of Spent Fuel Boiling Event

ML20153D150
Person / Time
Site:	Turkey Point
Issue date:	02/18/1986
From:	Wohl M Office of Nuclear Reactor Regulation
To:
Shared Package
ML17342A383	List: ML20153D082 ML20153D108 ML20153D150 ML20153D172 ML20153D204 ML20153D244 ML20153D252
References
OLA-2, NUDOCS 8602240033
Download: ML20153D150 (10)

Text

,

h.

a .,

w

.Y-UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD 4

In the Matter of )

) Docket Nos. 50-250 OLA-2 FLORIDA POWER a LIGHT COMPANY ) 50-251 0LA-2 i

)

(Turkey Point Plant, Units 3 and 4) ) (SFP Expansion)

AFFIDAVIT OF MILLARD L. WOIIL ON CONTENTION 4

1. My name is Millard L. Wohl. I am. a Reactor Engineer in the Technical Specifications Coordination Branch, Division of Human Factors Technology, U.S. Nuclear Regulatory Commission. Prior to November 24, 1985, I was a Nuclear Engineer in the Accident Evaluation Branch, Division of Systems Integrations where . I performed radiological consequence evaluations for the NRC Safety Evaluation (SE), dated November 21, 1984, on the expansion of the spent fuel storage capacity at Turkey Point Units 3 and 4. A statement of my professional qualifications is attached.

2. The purpose of this affidavit is to address Contention 4. With respect to Contention 4, I have read " Licensee's- Motion for Summary Disposition of Contentions" and " Licensee's Statement of Material Facts as to Which There Is No Genuine Issue To Be Heard With Respect to Intervenors' Contentions ," dated January 23, 1986. The material facts stated in relation to Contention - 4 are correct and I concur in the conclusions reached in the supporting affidavit.

3. Contention 4 states:

That FPL has not provided a site specific radiological analysis of a spent fuel boiling event that proves that offsite dose

- l 9602240033 860219 PDR ADOCK O DR _

o

'f limits and personal [ sic] exposure limits will not be exceeded

~

in allowing the pool ' to boil with makeup water from only seismic Category I sources.

Contention 4, as admitted by the Licensing Board and in view of the bases presented, esserts that personnel (onsite) and offsite dose limits in Parts 20 and 100 will be exceeded in a spent fuel pool boiling accident because Licensee extrapolated from calculations performed for the Limerick plant.- Intervenors maintain that there - may be critical differences - between the two pimits regarding the saturation values of noble gas and iodine inventories as a result of fuel failure and. increased enrichment. In addition, the amount of defective fuel rods may be more than one percent of the rods in the pool due to fuel failure and the gap activity of noble gases, such as krypton 85, and fission products such as radioactive iodine.

4. The Staff analyzes doses due to accidental releases by using guidelines of 10 CFR Part 100, which contain the criteria for suitability of proposed reactor sites. Part 100 provides that doses at the boundary of the site exclusion area not to exceed 25 rem to the whole body or 300 rem to the thyroid from iodine exposure. 10 CFR 100.11. The 25 rem whole body dose limit following a postulated accident specified in 10 CFR Part 100 also corresponds numerically to the National Council on Radiation Protection recommendation of a once in a lifetime accidental or emergency dose for radiation workers. 10 CFR 6 100.11, note 2. The Staff does not apply 10.CFR Part 20 limits to analyses of accidents, but uses Part 20 to evaluate the radiological consequences of routine plant operation and maintenance.

J

9. ..

g 5. As described above, the requirements for offsite doses during a spent fuel pool boiling accident are judged by the Staff to be those in 10 CFR Part 100. In response to a Staff request for additional'information, FPL responded to several questions concerning, in part, the results of the analysis of offsite doses in the _ event that pool boiling occurred.

The Licensce's response indicated that the methodology and assumptions used were consistent with those used in a similar pool boiling analysis performed for Limerick, which was previously accepted by the NRC. The Turkey Point analysis used plant-specific assumptions and assumptions generic to all pressurized water reactors (PWR).

6. The Staff agrees that the Licensee's analysis of spent fuel pool boiling is site-specific and is not based on an extrapolation from the

, Limerick study. The Staff's conclusion is based on the results of the Licensee's analysis and the independent offsite radiological consequences analysis performed by the Staff, both of which utilized Turkey Point plant-specific and PWR generic assumptions and demonstrated that the 10 CFR Part 100 guidelines will not be exceeded during a spent fuel pool boiling event at Turkey Point.

7. Analyses of accident scenarios involving radioactive gas releases from stored spent fuels , such as the pool boiling accident, use assumptions outlined in Regulatory Guide (Reg Guide) 1.25, " Assumptions Used for Evaluating the Potential Radiological Consequences of a Fuel 3

Handling Accident in the Fuel Handling and Storage Facility for Boiling and Pressurized Water Reactors." Although Reg Guide 1.25 specifically covers - fuel handling accidents, the assumptions may be applied to other accidents involving the release of radionuclides such as a spent fuel pool

o i O

, boiling event. Reg Guide 1.25 assumes the release of the entire inventory of volatile radionuclide in the fuel assembly gap (the _ space between the fuel pellets and the " fuel rod cladding) and plenum from damaged assemblies under approximately 23 feet of water. Reg ,

Guide 1.25 inventories are applicable to assemblies with burnup up to 38,000 mwd /MTU batch average at discharge. The Reg Guide assumption is that an inventory of ten percent of the total fuel assembly lodines and noble gases (with the exception of 30 percent for Krypton-85) in the gap and plenum volumes is released upon clad perforation. In addition, an iodine decontamination factor (DF) of 100 is assumed for 23 feet of water cover and appropriate airborne radionuclide filtration / mixing, if any, is applied in the analysis before release to the outer atmosphere.

8. The Licensee used an upper limit of 50,000 mwd /MTU for the allowed burnup of assemblies in the spent fuel pool. Generally, evaluations of offsite radiological consequences of fuel handling accidents involving damaged fuel assemblies with burnup greater than 38,000 Ml?d/MTU batch average at discharge (extended burnup assemblies), may be performed using Reg Guide 1.25 assumptions, but with modified . gap and plenum fractional volatile radionuclide inventories. The fractional inventories range from a few percent (less than the ten percent recommended by Reg Guide 1.25) to as much as 40-50 percent for certain high burnups/radionuclide combinations. The gap and plenum fractional inventories for the highest power assembly are computed as a function of either burnup or time , temperature, and burnup using the GAPCON-THERMAL-2 computer code in conjunction with the ANS 5.4 fission gas release standard (model) presented in the American Nuclear

D 4

Society _in " Radioactive Gas Release from LWR Fuel,'" C. E. ' Beyer, draft ~

NUREG CR-2715, April 1987. In generating these estimated fractional inventories, a constant maximum-allowed peak linear heat generation rate (LHGR) for PWR's .is assumed. This is a . conservative assumption since each fuel assembly may not operate at this peak due to power variations and its location in the core. In fact , the assumption of a constant maximum-allowed peak LHGR appears to be conservative within a factor of 2-3 for gap and plenum volatile inventories.

9. In addition to the conservative - assumption regarding fuel assembly power operation noted above, three other - sources- of conserva-tism were used in the Staff's independent analysis. First, an iodine decontamination factor (DF) assigned to the pool was a factor of 80, which assumes less removal of iodine in the pool water than suggested by Reg Guide 1.25. (This factor is probably conservative by about a factor of three.) Second, the plateout of volatile iodine released from the fuel into the gap and fuel rod plenum was_ ignored. Third, the Staff assumed modified inventories (NUREG CR-2715) for volatile gap and plenum radionuclides. It is the Staff's position that about 10 percent or less t

of the iodine assumed to be released into the gap will remain volatile at the fairly low temperatures after the fuel has been allowed to cool for about a day or more. Thus, the Reg Guide 1.25 assumption of a volatile gap iodine fraction of 10 percent may be high by about .a factor

of ten. Thus, the Staff concludes that an analysis which uses gap activ-1 ity for past operation at the peak linear heat generation rate, a conser-vative pool decontamination factor, and assumes all gap todine being

s.

volatile, coupled - with a one percent fuel failure assumption, leads to a conservative estimate of thyroid doses. .

10. In performing its analysis,' the Licensee ' assumed fuel assembly i

gap. activities as specified in Regulatory Guide 1.25, namely 10 percent of the lodines and noble gases except for thirty percent of the Krypton-85, 1

suitably conservative assumptions. Additionally, the Licensee. assumed one percent fuel failure. The results of this analysis were 0-2 hour doses of 0.28 rem to the. thyroid and 0.00018 rem to the. whole body at the Excluskn Area ~ Boundary.' These : doses are a small fraction (less than 10 percent) of the 10 CFR Part-100 guideline values of 300 rem thyroid and 25 rem whole body.

11. Notwithstanding these Licensee analytic results, the Staff

- performed an independent offsite radiological consequence analysis for the spent fuel pool boiling scenario assuming one percent fuel failure (i.e.,

f j minor perforations in the cladding of one percent of the fuel discharged from the reactor during the last refueling)--a value commonly used in ,

accident analysis. The Staff made the censervative assumption. that all the stored spent fuel assemblies ha'd 22 percent - gap iodine activities, i corresponding to a burnup of 50,000 mwd /MTU , with the concomitant l

)

assumption that each assembly had operated at the . maximum allowable linear heat generation rate for its entire lifetime and calculated a 0-2 hour l thyroid dose of 0.85 rem at the Exclusion Area Boundary. This is a small fraction of the 10 CFR Part 100 guidelines value 300 rem for the thyroid. Even assuming that the Intervenors are correct that fuel failure greater than one percent were to occur, the resulting doses would be

. . .- ,  ; . ., _. __.._.__..;____.._.,.__..... . .__. , _ . _. . I

e within ' Part 100 guidelines. For example . if there is a 10 percent fuel failure, the resulting dose (8.5 rem thyroid) would meet Part 100.

12. To summarize, the Licensee's analysis of the spent fuel pool boiling event used a methodology similar to that at Limerick and used appropriate generic and site-specific assumptions consistent with Reg Guide 1.25. The results of Licensee's analysis and the Staff's independent analysis confirm the offsite dose guidelines in 10 CFR Part 100 will not be exceeded.

The foregoing and the attached statement of professional qualifications are true and correct to the best of my knowledge and belief.

h 5L.1 f. W _

Millard L. Wohl Subscribed and sworn to before me this g day of February,1986.

fYj $=k_)

Notary Public h

My commission expires: 7/i /F'(,

h MILLARD L. WOHL PROFESSIONAL QUALIFICATIONS .' .

TECHNICAL SPECIFICATIONS COORDINATION BRANCH DIVISION OF HUMAN FACTORS TECHNOLOGY I am presently employed as a Reactor. Engineer in the Technical Specifications Coordination Branch, Division of Human Factors Technology.

Until November 24, 1985, I was employed as a Nuclear Engineer in the Accident Evaluation Branch, Division of Systems Integration, U.S. Nuclear Regulatory

. Comission, Washington, D.C. My duties in this position were to conduct site

'and accident analyses Emergency Response Facility Appraisals, and various other safety-related studies for nuclear power, test, and research reactor facilities. This includes probabilistic risk assessment (PRA) analyses for 4

power reactor environmental impact statements as well as evaluation of .

applicant / licensee PRA's.

I attended Case Western Reserve University (formerly Case Institute of Technology) and received a B.S. degree in Physics in 1956. I received a M.S.

degree in Physics from Indiana University in 1958. I did additional graduate work as a special student in Nuclear Engineering Science at Columbia University and in Nuclear Engineering at Case Western Reserve University from 1962 through 1964. I have 'nad short courses in Reactor Safety,' Emergency Preparedness, Probabilistic Risk Assessment, and Human Reliabiltty.

l

- o

_2

.1 1

I was a teaching assistant in Physics at Indiana University from ,1956 - 1958.

L I have taught physics, physical science, mathematics, and statistics-in the evening divisions of Baldwin-Wallace College, the Ohio' State University and Cuyahoga Community College from 1958 - 1973.

In 1957, I participated in the Special Power Excursion Reactor Tests at the j SPERT-II Facility, National Reactor Testing Station. Arco, Idaho.

In 1958. I joined the NASA Lewis Research Center staff in Cleveland, Ohio.

~

. My initial duties involved the writing of Monte Carlo computer codes for the

, determination of radiation shielding requirements and propellant radiation heating for proposed nuclear-powered rocket designs. Other assignments involved methods development and shielding and nuclear safety artlyses for

! numerous proposed mobile nuclear vehicle applications including the ~

~

4 Multi-purpose Nuclear Airplane. I was co-author of a study on disposal of radwaste in space, perfonned for the USAEC. Numerous other technical '

publications evolved in the course of the NASA work, some presented at ANS meetings. Additionally, during the period 1958 - 1973 I had substantial research contract management responsibilities.

i 1

In 1973 I joined the General Atomic Company in La Jolla, California, as a nuclear engineer. At General Atomic I perfomed a variety of nu' clear safety-related analyses for the High-Temperature Gas-Cooled Reactor (HTGR).

. These included the analysis of Design Basis Depressurization Accidents (DBDA) l and containment integrity stuties, as well as computer code upgrading and l modification.

.m. _ . - _ . , _ . , _ . ..~, , _ . , , , . . - . _ ,.- _ _._ ,_.m. _ ___ ..,,,__,..-..-_,-_m,, , , , ....,-,,-,m.m ., ..y...

I E

i

-1 1

In-1975, I joined the Accident Analysis Branch in the Division of Technical Review, U.S. Nuclear Regulatory Commission.- My responsibilities involved f

site characteristic studies and accident analyses. More recently, I have had expanded responsibilities, including Design Basis and Severe Accident (PRA)

Analyses for staff Safety Evaluations and Environmental Impact Statements.

These analyses include reactor core and piping system radiological accident analyses, steam generator repair accident analyses, core reload accident evaluations, spent fuel pool rerack accident evaluations, containment

. enclosure shielding analyses, and severe accident consequence and risk analyses. Additionally, I.have participated in operating plant Emergency ResponseFacility(ERF) appraisal. Also, I have had substantial contract management and expert hearing witness responsibilities.

Presently, I am involved in the upgrading of nuclear power plant Technical ~

Specifications in the newly formed Technical Specifications Coordination Branch, Division of Human Factors Technology.

- . - - . + , _ _ = _ , - - - , , - . , - , , . . , + , ---..v-