ML17193A522
| ML17193A522 | |
| Person / Time | |
|---|---|
| Site: | Dresden  |
| Issue date: | 11/04/1980 |
| From: | Wohl M Office of Nuclear Reactor Regulation |
| To: | |
| Shared Package | |
| ML17193A517 | List: |
| References | |
| NUDOCS 8011070214 | |
| Download: ML17193A522 (5) | |
Text
In the Matter of UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD COMMONWEALTH EDISON COMP~Y Docket Nos. 50-237 50-249 (Spent Fuel Pool Modification)
(Dresden Station, Units 2 and 3)
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SUPPLEMENTAL TESTIMONY OF MILLARD L~ ~!OHL ON CONTENT! ON 6 I, Millard L. Wohl, do state as follows:
I am employed by the United States Nuclear Regulatory Commission, as a nuclear engineer in the Division of Systems Int~gration, Accident.Evaluation Branch.
A statement of my professional qualifications is attached to this t~stimony.
This testimony addresses Contention 6 in the Memorandum and Order of the Atomic Safety and Licensing Board of September 9, 1980, concerning the matter of accident consequences.
- 6.
The application inadequately addresses the increased consequences of accidents considered in the FSAR, SER, and FES associated with the operating license review of Dresden Units 2 and 3 due to the increased number of spent fuel assemblies and additional amounts of defective fuel to be stored in the spent fuel pool as a result of the modification.*
The accidents considered in the documents referred to in Contention 6 are (1)
Control Rod Drop, (2) Tne Refueling Accident, (J) Main Steam Line Break Outside Orywell, an_d (4) Loss of Coolant Inside the Drywell.
The increased number of spent fuel assemblies stored in the spent fuel pools has no impact on the conse-quences of any of the above accidents..
tAcc~dents {lJ, (3J, and (4) are not relevant ~o the proposed action, and are not-further discussed.)
A Fuel Handling Accident, postulated to occur in the spent fuel pool area would have consequ_ences similar to those of the Refueling Accident, which occurs* in the core area and is defined as the impact of one assembly on another assembly I
- with damage to all the fuel rods in one assembly.
In the Staff Safety ~valuation Report of uctober 17, 1969, it was assumed for the Refueling Accident that tl) perforation of 4~ fuel rods (7 x 7 assembly) with consequent release of 20% of the noble gases and 10% of the halogens from the damaged rods into the reactor building occurs, t2*) 90% of the halogens released from the perforated fuel rods remain in the refueling water; t3) the remaining
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airborne fission products (ZO%*of the noble gases and 1% of the halogens contained in the fuel) within the build fog are assumed to be discharged to the 9tmosphere through the Standby Gas Treatment System, having a 90% iodine filter removal.
efficiency, and then through the stack over a two-hour period. It is assumed that the accident occurs 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after shutdown, resulting in a ZS rem 0-2 hr.
thyroid dose and a whole body dose of <l rem at the site boundary.
For the fuel assembly movements necessary in the present proposed spent fuel storage capacity increase action, the main crane, considered to be single-failure-proof by the staff tmeets the intent of NUREG-0554, May 1919, entitled 11Single-Fai lure-Proof Cranes for Nuclear Power Plants") wi 11 be utilized for all fuel movements, thereby reducing the likelihood of a Fuel Handling Accident.
The offsite radio-logical consequences ot a Fuel Handling Accident (which is of very low likelihood due to the redundant crane design) would be similar to those of the previously analyzed Refueling Accident and well. within the consequence limits of 10 CFK Part lUO.
- Thus, even though the total number of fuel assembly movements due to the fuel storage J
capacity modification over the plant lifetime increase by about five percent, the actual risk due to Fuel Handling Accidents during plant lifetime is lower than our previous estimates.
With respect to the consequences of additional amounts of defective fuel being stored in the spent fuel pools as a result of the modification, the final remaining 7 x 7 fuel assemblies were removed from the Dresden 3 core in February 1980. There are still 16 fuel assemblies of the 7 x 7R type in the Dresden 3 core, which are designed to avoid j:>ellet-clad interaction type failures. These will be removed in January 1982.
The sixty-one 7 x 7 fuel assemblies remaining in the Dresden 2 core will be removed in January 1981, along with most of the 7 x 7R assemblies.
All replacement assemblies in recent reloads have been of the improved 8 x 8 type.
Sampling for fuel defects leading to leakage in lUOO 8 x 8 assemblies from Dresden 2 and Dresden 3 have shown none.
Fuel performance should, thus, be greatly improved in future cycles.
Use of a redundant crane will substantially lower the likelihood of postulated Fuel Handli~g Accidents over the plant lifetime,thereby reducing overall risk.
It can, therefore, ~e concluded tnat offsite radiological consequences due to postulated Fuel Handling Accidents should not be increased in the new spent fuel storage configuration from those of the Refueling Accident analyzed in the staff SER of October 1969. Additionally, use of the improved 8 x 8 fuel, which has thus far shown no failures leading to leakage, will reduce noble gas release from stored spent fuel.
MILLARD L. WOHL PRUFESSIUNAL QUALIFICATIONS ACCIDENT EVALUATION BRANCH DIVISION OF SYSTEMS INTEGRATION I am employed as a nuclear engineer in the Accident Evaluation Branch, Division of Systems Integration, U. S. Nuclear Regulatory Commission, Washing-ton, D. C.
My duties are to conduct site and accident analyses and various other safety-related studies for nuclear power and non-power reactor facilities.
I attended Case Western Reserve University {formerly Case Institute of Technology) and received a B. S. degree in Physics in 1956.
I received an M. S. degree in Physics from Indiana University in 1958.
I did graduate work in Nuclear Engineering at Columbia University and Case Western Reserve University from 1962 through 1964.
I was a teaching assistant in Physics at Indiana University from 1956 - 1958.. I have taught physics and mathematics in the evening divisions of Baldwin-Wallace College, the Ohio State University and Cuyahoga Community College from 1958 - 1973.
In 1958, I joined the NASA Lewis Research Center in Cleveland, Ohio.
My initial duties involved the writing of Monte Carlo computer codes for the determination of radiation shielding requirements and propellant heating for proposed nuclear-powered rocket designs.
Other assignments involved methods development and shielding and nuclear safety analyses for numerous proposed mobile nudear vehicle applications.
Numerous technical publications evolved
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Additionally, during the period 1958 - 1973, I had substantial research contract management responsibilities.
In 1973, I joined the General Atomic Company in La Jolla, California, as a nuclear engineer. At General Atomic I performed a variety of nuclear safety-related analyses for the High-Temperature Gas-Cooled Reactor (HTGR).
These included the analysis of depressurization accidents and containment integrity studies, as well as computer code upgrading and modification.
In 1975, I joined the Accident Analysis Branch in the Division of Technical Review, U. S. Nuclear Regulatory Commission.
My responsibilities involved site characteristic studies and accident analyses.
Presently, I have similar but expanded responsibilities.