ML20244D551
| ML20244D551 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 06/09/1989 |
| From: | Burns K, Littlefield P, Reid D, Yasi D VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20244D518 | List: |
| References | |
| OLA, NUDOCS 8906190130 | |
| Download: ML20244D551 (21) | |
Text
. _ _ _ _ _ _ _ _ _ _ ___
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION before the ATOMIC SAFETY AND LICENSING BOARD
)
In the Matterof ) .,
)
VERMONT YANKEE NUCLEAR ) Docket No. 50-271-OLA POWER CORPORATION )
) (Spent Fuel Pool (Vermont Yankee Nuclear ) Expansion)
Power Smdon) )
)
Swom Written RebuttalTestimony of Donald A. Reid, Peter S. Littlefield, Daniel E. Yasi and Kevin J. Bums, Submitted by Vermont Yankee Nuclear Power Corpontion Pursuant to 10 C.F.R. I 2.1113(a)
Pursuant to 10 C.F.R. I 2.1113(a) and this Board's Memorandum of April 21,1989, the Licensee, Vermont Yankee Nuclear Power Corporation, submits the within sworn written rebuttal testimony of Donald A. Reid, Peter S. Littlefield, Daniel E. Yasi, and Kevin J. Burns in response to the submissions regarding Environmental Contention 3 filed by the New England Coalition On Nuclear Pollution dated May 23,1989.
8906190130 890609 PDR T ADOCK 05000273 PDR
l l
I. Identification of the Witnesses Messers. Reid and Littlefield are the same persons who are identified in " Sworn Written Testimony of Donald A. Reid, Michael J. Marian, Rudolph M. Grube, John -
M. Buchheit, Richard P.Pizzuti, and Peter S. Littlefield" filed by Vermont Yankee Nuclear Power Corporation in this matter on May 23,1989. In adition, the following persons have contributed to this rebuttal testimony:
Mr. Yasi is employed by Yankee Atomic Electric Company and is the lead Systems Engineer for the Vermont Yankee Pmject. His responsibilities include directing a staff of engineers who design plant process system modifications, perform engineering studies, and provide other engineering support services. He holds a Masters of Science degree in Applied Management and a Bachelors of Science degree in Mechanical Engineering. Mr. Yasi has 15 years experience in the nuclear power industry, including 8 years at Yankee Atomic Electric Company.
Mr. Burns is employed by Yankee Atomic Electric Company as a Senior Engineer in the Nuclear Engineering Department. He has 9 years of experience in the nuclear industry, all at Yankee Atomic Electric. His current responsibilities include the application of Probabilistic Risk Assessment techniques to the Vermont Yankee Nuclear Power Str2 tion. His previous responsibilities included thermal hydraulic transient analyses for the Vermont Yankee Nuclear Power Station. Mr. Burns holds Master of Science and Bachelor of Science Degrees in Nuclear Engineering.
2
II. Rebuttal
- 1. We have been advised that the merits of the Interveners' assertions concerning the risk of a severe reactor accident at Vermont Yankee are not properly the subject of this proceeding, and we have been instructed that we should not respond to these assertions. However, this Panel is concerned that, however procedurally correct (or required) that we ignore these assertions, any failure to respond to them is capable of being erroneously interpreted as our agreement to their accuracy. Even on cursory review, we have identified a number of false or erroneous statements, and we provide some examples.
A. The " Testimony of Gordon Thompson" (hereinafter Interveners' Testimony)(1) at Page 2 states: "My primary purpose in this testimony is to show that there is a significant danger associated with high-density pool storage of spent fuel at the Vermont Yankee plant, and that much safer alternatives exist."
This document, however, fails to recognize two basic determiteations of a study conducted by Brookhaven National Laboratory (2), namely, that no " prompt fatalities" would occur as a result of a worst case fuel pool accident and that the
" risk of injury is also negligible." Either the author of this document rejects the l~
work of the Brookhaven National Laboratory or his definition of a significant danger is significantly different than that contained in the Brookhaven study.
B. The Interveners' Testimony at Page 2 states: "Specifically, I show that a partial or total loss of water from the fuel pool can be expected to initiate an exothermic steam-zirconium or air-zirconium reaction which could lead to a release to the environment of a substantial amount of the long-lived radioactivity 3
contained in the spent fuel." It should be noted that nowhere does the Interveners' Testimony shar that either partial or total loss of water would lead to a release of a substantial amount of radioactivity. Instead, the Interveners' Testimony assumes that if either loss of water has taken place, all of the major radioactivity would be released. This was, however, the same assumption made in the Brookhaven study (2) that concluded that the " risk of injury was also negligible."
Data from Three Mile Island Unit 2, which melted a large fraction of the core, and Chemobyl, which bumed a fully operating com in place, do not support the assumption that 100% of the cesium is released. In fact, the estimates from Chernobyl are 6% to 20% for cesium.(3) It would appear that the similarity of the release mechanisms between this assessment and the Chemobyl event would dictate the use of a release fraction determined from a real event rather than an unsubstantiated assumption of 100%.
C. The Interveners' Testimony addresses the consequences of beyond design basis events associated with spent fuel stored in spent fuel racks but fails t3 l
addmss similar consequences of beyond design basis events associated with sr ent fuel stored in dry casks. The Interveners' Testimony is, in essene, a comparison of bevond design basis considerations for one storage method (fuel racks) to mihin design basis considerations for an altemative storage method (dry casks).
D. The proposed amendment mquests an increase in the number of spent fuel bundles permitted to be stored in the Vermont Yankee spent fuel pool from 2000 4
l to 2870. Rus, any discussion of the risks of spent fuel pool accidents must be limited to consideration of the effects of the addition of 870 spent fuel assemblies.
We note that while discussing consequences of a beyond design basis spent fuel accident at length, the Interveners' Testimony does not provide any discussion of the change in consequences due to the additional 870 bundles.
Rather, the Intervenor's Testimony merely assumes a certain 10131 inventory of radionuclides based on next-decade projections.
2 . The Interveners' Testimony at Page 9 states: " Loss of water could also be initiated by a severe reactor accident, in three ways. First, severe reactor accident scenarios can be associated with violent releases of energy through mechanisms such as hydrogen explosion, high-pressure melt ejection or steam explosion." He Interveners correctly use the term " severe reactor accident" for such hypotheses because no such violent energy releases are possible for any accident within Ver nont Yankee's design basis.
The applicable limits of 10 CFR 50.46 for emergency core cooling systems are a peak clad temperature ofless than 2200'F and a maximum hydrogen genemtion ofless i
than 1% (of the hypothetical amount that would be generated if all cladding were to react). Under these conditions, some fuel failure may occur due to clad perforation.
However, no fuel melt will occur.
This limit for hydrogen generation under design basis conditions limits the potential for hydrogen explosions. However,in keeping with the design in depth approach, 5
NRC regulations (10 CFR 50.44) require further standards for hydrogen control.
- Vermont Yankee is in compliance with these regulations, which require:
- 1. An inert containment atmosphere
- 2. Means for measuring hydrogen in containrnent
- 3. Means for controlling hydrogen gas concentrations in the containment following a postulated loss of coolant accident (LOCA)
Thus, hydrogen explosions are beyond the conditions specified in 10 CFR 50.46 i
and 10 CFR 50.44, and cannot occur as a result of any within design basis accioent at Vermont Yankee.
l High pressure melt ejection is a phenomenon hypothesized where molten core materials are postulated to be ejected from the reactor vessel at high pressure. This phenomenon involves significant amounts of molten fuel and failure of the reactor vessel to contain this molten material. Both of these conditions are beyond the l conditions specified in 10 CFR 50.46.
Steam explosion is a phenomenon hypothesized that results from the interaction of molten metals with water. This phenomenor, also requires significant amounts of molten core material, hence it is beyond the conditions specified in 10 CFR 50.46.
Thus, the mechanisms for violent energy release postulated by the Interveners cannot occur as a result of a within design basis accident at Vermont Yankee.
- 3. The Interveners' Testimony at Page 9 states: "less of water could also be initiated by a severe reactor accident,in three ways..... Second, a severe reactor accident could 6
create on-site contamination of sufficient severity that access of personnel was precluded. In that event, water in the pool could evaporate over the days following the accident." Based upon the following discussion, this scenario cannot occur as a result of any within design basis accident at Vermont Yankee.
The worst case accident within Vermont Yankee's design basis is the Loss of Coolant Accident (LOCA). For this accident, the 10 CFR 50.46 peak clad temperature limit of 2200 F may result in some fuel clad perforations, however, no fuel melt will l
occur. Even assuming a much larger NUREG-05889) source term, reactor building access would still be available within 90 days. Furthermore, the Spent Fuel Pool Cooling System (SFPCS) and the other needed support systems are environmentally qualified to operate for at least 90 days following a Design Basis LOCA. 'Ihus, pool boiling will not occur and the spent fuel pool's inventory loss will be by evaporation only, for which make-up water can be provided from a number of different sources without requiring access to the reactor building.
4 . The Interveners' Tesdmony at Page 9 states: "Ioss of water could also be initiated by a severe reactor accident, in three ways .... Third, an accident occurring during refueling could lead to a draining down of the pool to the top of the fuel assemblics (see Exhibit 2) followed by evaporation of the remaining water. Access of personnel for mitigative actions could be precluded by contamination arising fmm involvement of the reactor core in the event." Based upon the following information, this scenario cannot occur as the result of a within design basis accident at Vermont Yankee.
The Vermont Yankee Station Safety Analysis identifies the worst case refueling accident as one resulting from the accidental dropping of a fuel bundle onto the top of 7
_ _ _ _ _ _ _ _ _ - _ - - - - - - - - - - - - - ^ - - - - - - . _ . __ ___ _ -
the core. The radiological consequences for this accident are well below the guidelines set forth in 10 CFR 100. Also, for this design basis refueling accident, access to the reactor building is not precluded after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, and make-up water to the pool can be provided from a number of different sources without requiring access to the reactor building.
Within Vermont Yankee's design basis, any other accident occuring during refueling is bounded by the design basis refueling accident described above. The postulated refueling accident described in the Interveners' Testimony, in which all water is lost from the pool, is beyond Vermont Yankee's design basis.
The Vermont Yankee Final Safety Analysis Repon describes the spent fuel pool design as follows: "The fuel storage pool is designed so that no single failure of structures or equipment will cause inability (1) to maintain irradiated fuel submerged in water, (2) to reestablish normal fuel pool water, or (3) to safely remove fuel from the plant."
- 5. The Interveners' Testimony at Fage 8 states: "One mechanism for water loss is 1
earthquake-induced failure of the pool structure." This statement is not true ifit was intended to refer to a design basis earthquake occurring at Vermont Yankee. The spent fuel pool at Vermont Yankee was evaluated for the increased loadings imposed by die subject reracking. The evaluation concluded that the spent fuel pool stmetural capacity exceeds the seismic design requirements of the Vennont Yankee Final Safety Analysis Report for all anticipated seismic loadings. Therefore, no structural failure of the spent fuel pool will result from a design basis canhquake.
8
- 6. The Interveners' Testimony at Page 12 states: "less of watercould arise as a result of.... drop of a shipping cask ...." As described below, this event cannot occur as the result of a within design basis accident at Vermont Yankee.
The Vermont Yankee Reactor Building crane is used to move shipping casks in the area of the spent fuel pool. The reactor building crane is single failure proof and the NRC staff has found the crane acceptable in accordance with NUREG-061.2, " Control of Heavy Loads at Nuclear Power Plants".
In addition, the NRC staff has specifically accepted the Vermont Yankee pmgram for ensuring that any special lifting device, used in conjunction with the reactor building crane for any cask movement, meets all applicable operating license and NUREG-0612 criteria. This program ensures that any heavy load, such as a shipping cask, cannot be dropped in the event of a single faihtre in the lifting system.
Based on Vermont Yankee meeting all NRC requirements for Generic Issue A-36 (NUREG-0612), no credible single failure in any lifting system used for cask movement will result in a load drop.
7 . The Interveners' Testimony at Page 9 states: " Finally, the pool might lose water as the result of an act of sabotage." Further, the Interveners' Testimony at Page 12 states:
" Loss of water could arise as a result of.... sabotage." This event cannot occur as a result of a design basis sabotage threat at Vermont Yankee and Vermont Yankee offers the following information: 10 CFR 73.55 provides the NRC regulations necessary for physical protection of licensed activities in nuclear power reactors against radiological sabotage. As such, Vermont Yankee maintains an NRC approved security plan that includes provisions for coping with various security contingencies. One aspect of the 9
plan is the ability to cope with a sabotage threat from either an internal or an external source. This plan would cover the reactor building where the fuel pool is located. Due to the safeguards nature of the plan, Vermont Yankee cannot disclose any specific details concerning either the nature of a projected threat or the planned response to mitigate that threat.
- 8. Based on the foregoing clarifications and corrections to the information presented in the Inte: Tenors' Testimony, it is clear that the risks claimed in that testimony of a zircaloy fire are not possible as a result of any within design basis event occurring at Vermont Yankee. To the contrary, wholly apart from the errors contained in the 1
Interveners' Testimony's assessment of the consequences of such an event, the probability of such an event (which is not assessed explicitly), is controlled by the possibility of occurrence of one of the beyond design basis events that the Interveners' Testimony assumes happens.
III. TechnicalInformation Reganiing ASLB Memorandum dated May 25,1989W.
- 1. Item 2, third sentence of the May 25,1989 Board memorandum questions "whether
[the Interveners'] model of low-density racks corresponds to the racks at Vermont Yankee immediately prior to the current application (at which time VY had authority to store 2000 elements)..." The short answer to this is: No, the previous racks were high density racks to the same extent as the recently installed racks (with a 2870 spent fuel assembly capacity) are high density racks. The basis for this is as follows:
The real issue is whether fuel stored in the "new" (i most recently installed)
Vermont Yankee fuel racks is significantly more susceptible to zircaloy reactions (under 10
beyond design basis conditions) than fuel stored in the "old" (iL the racks in place <
prior to July 1988) Vermont Yankee fuel racks. To adequately address this issue, the - )
following two questions rnust be considered:
A. What design characteristics for spent fuel racks play a significant role in the potential for zircaloyreactions?
B. What is the change in these design characteristics in going from the old to the new racks?
A review of the Sandia (1979)(9) and Brookhaven (1987)(2) reports indicates that j I
l the important design characteristics are:
- 1. Decay heat of spent fuel L
- 2. Open channel vs. closed channel rack configuration f 1
- 3. Size ofinlet orifice )
i
- 4. Flow area of channels (center - to - center spacing) j i
- 5. Rack to rack and rack to wall spacing
- 6. Rack baseplate to fuel pool floor spacing i The change in these design characteristics between the old and new racks, and the impact on zircaloy reaction potential, is discussed below. This discussion is based on -
i the previously referenced Sandia and Brookhaven studies which treat spe,. luel heatup -
i in air. ;
l 11 _
i
E , w>m.
, , !yp
(
f 1.' Decay Heat ofSpent Fuel '
I Spent fuel with higher decay heat levels can reach higher temk.imes, hence,' . has a greater likelihood of a zircaloy reaction. No change in the spent fuel itselfis being proposed by this' amendment. Thus, the change in rack design has no impact on the ;
spent fuel decay heat generation rate.
- 2. _Open Channel vs. Closed Channel Rack Configuration
. Both the old and new rack designs are cl6sid channel' designs, with flow inlet orifices at the rack baseplate only. Thus no ch'ange in zircaloy reaction potential is'.
expected due to this design characteristic.
- 3. Size ofInlet Orifice The old rack design had an inlet orifice diameter for each channel of 3.625 inches.
The value for the new rack is 3.75 inches. These values are very similar, and no .
significant impact on spent fuel cooling in air is expected. If anything, the slightly larger inlet orifice for the new racks will tend to promote more natural circulation air flow.
- 4. Flow Area of Channels
. The flow area of the rack channel also affects the natural circulation air flow rate.
~
The center - to - center spacing is often used as an indication of flow area. Thus, low
' density racks have a large center - to - center spacing (with a large flow area) and high-
. density racks have a smaller center - to - center spacing (with a smaller flow area).
12 L
' The old Vermont Yankee racks have a center- to - center spacing of 7 inches, and the new racks 6.2 inches. Since the "old style" BWR racks investigated by Sandia and ,
Brookhaven (and referred to by Interveners' Testimony _ on page 4) typically have ~a -
center - to - center spacing as large as 8 to 11 inches, both the old and new Vermont .
Yankee racks could be considered "high density".
The smaller flow area of the new racks implies less natural circulation ' air flow. -
However, the Vermont Yankee spent fuel is stored with the fuel bundle channel in place. The flow area within the fuel channel is the most
- uyviunt parameter since, as shown by the Sandia study, convective heat transfer from the spent fuel rods to air is .
the dominant heat transfer mechanism. He flow area within the spent fuel channel is a ;
function of the spent fuel, and is independent of rack design.' Dus, the decreased rack channel spacing is expected to have little effect on spent fuel heatup in air. j
- 5. Rack to Rack and Rack to Wall Spacing j The spaces between racks and between the racks and the pool walls affect the air flow which enters the pool. De old racks required a 1/4 inch space between racks and -
a 9 inch space between the racks and the wall. De new racks have'a 2 inch space-between racks but only require a 2 inch space between the racks and wall. Thus, the change from old to new racks means that less flow will enter the pool through the wall -
rack space, but more will enter between racks. It is not clear whether this tradeoff will result in a net increase or a net decrease in air cooling. The Sandia and Brookhaven j
~ 1' studies assume no space between racks, hence the rack to rack space was not a variable l m these srM. !
\ '
L -l l-a 13 .
, 1 e
1_____ _ _ _ _ . _
- 7,
- 6. Rack Baseplate to Fuel Pool Floor Spacing 1 u
The Sandia study assumed a constant rack baseplate to floor spacing of 16 inches.
This value was about 6 inches with the old Vermont Yankee racks and is about 10 inches with the new Vermont Yankee racks. 'Ihus, the new racks have a larger space for air flow below the racks which promotes natural circulation air flow.
For the case of a partial pool drainage postulated by the Interveners' Testimony, steam cooling is the important heat transfer mechanism since the water level blocks air from entering the bottom of the racks. For this case, the Interveners' Testimony states (at page 7) that "if no other heat transfer modes are signifk: ant, cladding temperature is independent of the decay heat output". . While not necessarily agreeing with this statement, we note that the spent fuel decay heat load is independent of rack design. Of all the rack design parameters discussed above, the following have little impact on the partial pool drain down scenario:
i Size ofinlet orifice Rack to rack and rack to wall spacing Rack baseplate to fuel pool floor spacing.
)
l This is because these parameters primarily affect the air inlet flow, and no air inlet is allowed for this case because of the blocked inlets. The following parameter:
1 Flow area of channel j
I 14
_=--h------------_ - . - - . - _ - _ - - . - _ - d
l has little effect on steam cooling effectiveness. As noted above for the air cooling case, i
it is the fuel channel flow area that is most important, and this is independent of rack i design. Fmally, for the parameter:
Open channel vs. closed channel rack configuration there is no difference between the new and oki Vermont Yankee racks.
In conclusion, the differences between the old and new Vermont Yankee spent fuel racks do not result in a significant increase in the likelihood of a zircaloy fire under beyond design basis conditions. For the case of a partial water loss with steam cooling, no significant design changes are noted between the old and the new racks.
For the case of total loss of pool water followed by air cooling, both the old and the !
new racks havudvantages, with neither havir:g significant advantage over the other based on the available anatyrs Gandia and Brookhaven reports).
Therefore, in the context in which the Interveners' Testimony defines and employs 2 the terms, both the old racks and the new racks are "high density" racks and the old racks were not " low density" racks. In particular, NECNP's "model of low density racks" does not correspond "to the racks at Vermont Yankee immediately prior to the current application." I
]
l l
- 2. Item 4a of the May 25,1989 Board memorandum questions: " .in what manner, if 'l at all, do they [ descriptions provided by the parties in their testimony) factor in the -
benefits attributable to the passive nature of dry cask storage and the attendent 15 i l
operational benefits, set fonh in the proposed rulemakmg on dry cask storage (54 Fed.
Reg.19379,19380 (May 5,1989))...? In response, Vennont Yankee provides the j following information.
The referenced Federal Register notice lists two specific benefits afforded by utilizing dry cask storage. The first benefit is that there is "no need to provide operating cooling systems to purify and circulate cooling water or other fluid. . active mechanisms, such as pumps and fans, are not required." Vermont Yankee points out, that as previously testifiedW, any spent fuel stored in dry casks must first spend a period of time, nominally 5 to 10 years, stored in the existing spent fucI pool to allow time for decays in waste heat generation and fission products. As such, the existing spent fuel pool and its associated equipment must continue to be opanted. Thus, with .
the use of dry cask storage, there can be no operational benefits obtained by eliminating the use of active spent fuel pool cooling equipment.
)
Additionally, the language contained in the referenced Federal Register notice regarding the lack of a need for operating cooling systems occurs in the same paragraph that states "...it is anticipated that most spent fuel stored in casks will be five years oki or more." Obviously, the Federal Register notice does not imply tha: existing spent f fuel pools should be abandoned and their cooling systems dismantled. Rather, it compares possible options (dry cask storage versus construction and use of a second l
wet storage pool) assuming an operating facility's existing spent fuel storage capacity i
has been exhausted. Therefore, the referenced Federal Register notice has no. I applicability to this amendment. '
i l
16 !
- - _ _ _ _ _ - _ . 1
The second benefit listed by the referenced Federal Register notice is "that the
{
potential for corrosion of the fuel cladding and reaction with the fuel is reduced, because an inen atmosphere is expected to be maintained inside dry spent fuel storage casks." Vermont Yankee cautions that this statement should not be taken to infer that the water environment in a spent fuel pool is corrosive to zirconium clad fuel. For comparison, Vennont Yankee points out the following.
A study performed by Battelle Northwest Laboratories) provides spent fuel clad
- oxidation rates for spent nuclear fuel stored in water storage pools of 0.009 m/ year.
This translates to approximately 1/1000 in. during 3000 years. A second report, issued by the Eleenic Power Research Institute (8), concludes: " Evidence that pool storage of k
spent fuel is a safe, efficient technology continues to accumulate....Further, fuel l examinations have yielded no evidence that Zirraloy< lad fuel degrades as a result of water storage."
i i
i i
I l
l 17 i
l
C W Donald A.Reid Peter S.Littlefield/
Daniel E. Yasi Kevin J. Bums Commonwealth of Massachusetts:
VJoksrM 0:xNTYss.
Then personally appeared Donald A. Reid, who being first duly swom, made oath that the foregoing statements are true, this M of June,1989, before me:
~
wrb b ,
Notary Public My Commission expiresaw J9 /Pt/
l Commonwealth Of Massachusetts:
N c4va rY ss.
Then personally appeared Peter S. Littlefield, who being first duly sworn, made oath that the foregoing statements are true, this 9* of June,1989, before me:
MM Notary Public My Commission expires & M nf/
/
18
Commonwealth of Massachusetts:
ss.
Then personally appeared Daniel E. Yasi, who being first duly sworn, made oath that the foregoing statements are true, this of June,1989, before me:
Notary Public My Commission expires Commonwealth of Massachusetts:
mru cwiY ss.
Then personally appeared Kevin J. Burns, who being first duly swom, made oath that the foregoing statements are true, this 4* of June,1989, before me:
ON&
Notary Public My Commission expires the if 4#
W 19
REFERENCES 1
Testimony of Gordon Thompson dated May 23,1989.
2
" Severe Accidents in Spent Fuel Pools in Suppon of Generic Safety Issue 82",
NUREG/CR-4982, dated July 1987.
3
" Report on the Accident at the Chemobyl Nuclear Power Station", NUREG-1250, dated Jannery 1987.
4
" Interim Staff Position on Environmental Qualification of Safety-Related Electrical Equipment" NUREG-0588, Revision 1, dated July,1981.
5 ASLB Memorandum (Issues for Consideration at 6/21/89 Oral Argument), dated May 25,1989.
6 Sworn Written Testimony of Donald A. Reid, Michael J. Marian, Rudolph M. Grube, John M. Buchheit, Richard P.Pizzuti, and Peter S. Littlefield, Submitted by Vermont Yankee Nuclear Power Corporation, daied May 22,1989.
7
" Behavior of Spent Nuclear Fuel in Water Pool Storage", A. B. Johnson Jr., Battelle Pacific Northwest Laboratories, BNWL-2256/UC-70, dated September 1977.
8
" Surveillance of LWR Spent Fuel inWet Storage", Electric Power Research Institute, EPRI NP-3765, Final Report, dated October 1984 9
" Spent Fuel Heatup Following Loss of Water During Storage", Sandia Laboratories, NUREG/CR-0649, dated March 1979.
20
[0(.h E Til UmC VYN-1ss ASLB - Fed Ex.
RKGABSTE.VY
'89 J3113 P4 :55 Certificate of Service 00Cri~ .
I I I, R. K. Gad III, hereby certify that on June 9,1989, I made service of the within document by depositing a copy thereof with Federal Express 1 (Monday delivery) (except where indicated by an asterisk, in which case by.
l mailing a copy thereof postage prepaid) as follows:
Charles Bechhoefer, Chairman Gustave A. Linenberger, Jr.
l Administrative Judge Administrative Judge l Atomic Safety and Licensing Panel Atomic Safety and Licensing Panel U.S.N.R.C. U.S.N.R.C. '
l East West Towers Building East West Towers Building .
I 4350 East West Highway 4350 East West Highway Bethesda, Maryland 20814 Bethesda, Maryland 20814 l James H. Carpenter Anae Spielberg, Esquire.
Administrative Judge Harmon, Curran & Tousley Atomic Safety and Licensing Panet Suite 430 U.S.N.R.C. 2001 S Street, N.W.
East West Towers Building Washington, D.C. 20009 4350 East West Highway Bethesda, Maryland 20814 Adjudicatory File
- John Traficonte Esquire Atomic Safety and Licensing Board Assistant Attorney General Panel One Ashburton Place U.S.N.R.C. Boston, Massachusetts 02108 l Washington, D.C. 20555 Geoffrey M. Huntington, Esquire Samuel H. Press, Esquire l Environmental Protection Bureau George E. Young, Esquire State House Annex Vermont Department of Public 25 Capitol Street Service Concord, New Hampshire 03301 120 State Street Montpelier, Vermont 05602 Ann P. Hodadon, Esquire Patricia A. Jehle Esquire U.S.N.R.C.
One White Flint North ' '
11555 Rockville Pike -
Rockville, Maryland 20852 l . .- g s N
R. K. Gad III
. . . . .