ML19289F089
| ML19289F089 | |
| Person / Time | |
|---|---|
| Site: | 07002623 |
| Issue date: | 05/11/1979 |
| From: | Hoefling R, Ketchen E NRC OFFICE OF THE EXECUTIVE LEGAL DIRECTOR (OELD) |
| To: | |
| Shared Package | |
| ML19289F083 | List: |
| References | |
| NUDOCS 7906010349 | |
| Download: ML19289F089 (7) | |
Text
.
d UNITED STATES OF Af1 ERICA NUCLEAR REGULATORY COMMISSION BEFOR,i THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
DUKE POWER C0f1PANY
)
Docket No. 70-2623
( Anendment to fiaterials License
)
SNM-1773 for Oconee Nuclear Station
)
Spent Fuel Transportation and Storage
)
at McGuire Nuclear Station)
)
STATEMENT OF MATERIAL FACTS AS TO WHICH THERE IS NO GENUINE ISSUE (CESG AND CAROLI!"A ACTION CONTENTION 2) 1.
Approximately 42,000 persons live within 0.8 km of the route over which Oconee spent fuel will be transported. These persons could receive a cunulative dose of about 0.1 man-rem fron 300 shipments which is the equivalent of 0.003% of the dose received annually from naturally occurring sou rces.
(Affidavit of C. Vernon Hodge and R. Daniel Glenn(Glenn Affidavit),
Table I; Environnental Impact Appraisal Related to Spent Fuel Storage of Oconee Spent Fuel at McGuire Nuclear Station, Unit 1 Spent Fuel Pool (EIA), pp. 30-31).
2.
The average dose to a person along the route would be about 0.003 mrem per year.
The maximum individual who is 30m from the roadway as each of the 300 shipnents pass would receive a dose of 0.02 nren from the 300 shipments which is equivalent to 0.02% of the dose received annually from naturally occurring sources (Glenn Affidavit, Table I; EIA, p. 31).
79060103 K
. 3.
The doses to the population along the route of the proposed shipnents and to the maxinum individual are not significant and do not represelt an unacceptable incremental burden.
The nunber of health effects is too small to estimate.
(Glenn Affidavit, p. 7 and Table I; EIA p. 31.)
4 The Staff has analyzed two additional situations for exposure to the public involving (1) a traffic jam holding the truck and associated cask in a congested area for up to three hours, and (2) a vehicle closely following the cask over a major portion of the route.
(Glenn Affidavit, pp. 3-4; EIA, p. 31.)
5.
Assuming a traffic jam occurs in an area with a population of 10,000 persons per square mile uniformly distributed and the truck remained in the same location for 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, the population dose would be less than 0.2 man-rem and the maximum exposed individual 3 m. from the cask would receive 15 nren.
(Glenn Affidavit, p. 4; EIA, p. 31.)
6.
If a car is assumed to travel directly behind the truck carrying the cask for four hours at a distance of 30 n, the dose is calculated to be 0.16 mrem to each occupant.
If the same individual were to follow each of the 300 shipments, the total dose would be 48 mren.
(Glenn Affidavit, Table I; EIA, p. 32.)
2235 053 7.
For travel in the direction opposite to that of the shipments, the cumula'.1ve population dose for the assumed 300 shipments is about 0.04 pe rso n-rem.
The dose to a hypothetical individual who passes each of the 300 shipments would be about 0.03 mrem, or about 0.03 percent of the background dose received by an individual during one year.
(Glenn Affidavit, p. 3 and Table I.)
8.
The cumulative dose to persons traveling in the same direction and at the same speed as the shipment was calculated to be about 0.08 person-rem.
The dose to a hypothetical individual who travels concurrently with all 300 shipments would be about 3 mrem which represents about three percent of the background dose received by an individual in one year.
(Glenn Affivadit, p. 3 and Table I.)
9.
The doses to persons traveling over the transportation routes con-currently with spent fuel shipments are not significant and do not repre-sent an unacceptable incremental burden.
The number of health effects is too smal~ to estimate.
(Glenn Affidavit, p. 7 and Table I; EIA, pp. 31-32.)
10.
The spent fuel cask which will be used to transport spent fuel from the Oconee facility to the McGuire facility will require a certificate of compliance from the Nuclear Regulatory Conmission certifying that the cask meets all applicable Commission regulations.
(Affidavit of C. Vernon Hodge, Willian H. Lake, Jr. and R. Daniel Glenn (Hodge Affidavit), p. 3; EIA, pp. 17, 33.
2235 054
. 11.
Certification of a spent fuel cask to the Commission's regulations provides high assurance that the cask can survive a wide range of trans-portation accidents without the release of significant radioactivity.
(Hodge Affidavit, pp. 3-4; EI A, pp. 33-37).
12.
Spent fuel casks have been the subject of extensive testing, incluoing recent Department of Energy sponsored full-scale inpact tests; demonstrating that casks can contain ind shield their contents under accident conditions.
(Hodge Affidavit, pp. 4-5; EI A, p. 33).
13.
Actual experience with spent fuel cask shipments includes data on two accidents.
In neither case did a radioactivity release occur.
(Hodge Affidavi t, pp. 5-7).
14 Transportation accidents have been classified by degree of severity.
That classification shows that as accident severity increases, the con-sequences increase but that the risk (probability x consequences) remains small for all accident conditions.
This is so because the probability of extreme accidents occurring is remote.
(Hodge Af fidavi t, pp. 7-8; EIA, pp. 33-34).
15.
The 10 C.F.R. Part 71 accident test standards are equivalent to conditions for severe accidents. The acceptance standard for these tests is no release of radioactive contents except for radioactive gases and 2235 055
. contaminated coolant containing total radioactivity not exceeding the specified small quantities of 10 C.F.R. Part 71.36.
(Hodge Affidavit,
- p. 8; EIA, po. 33-34).
16.
The Staff has analyzed a minor accident of undetected leakage of coolant and moderate accidents associated witn loss of neutron shield water and cask overpressurization.
(EIA, pp. 34-35.)
- 17. The radiological consequences of these accidents are insignificant ranging from a total body dose to the maximum individual of 1x10~4 mrem for undetected leakage of coolant to e total body dose to the maximum individual of 1.1x10-2 mrem for the cask overpressurization accident.
In the case of loss of neutrcn shield water, a member of the public at 10m from the cask for 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br /> would receive a dose equivalent to 1% of natural background.
(EIA, pp. 34-35).
18.
For postulated accidents beyond the severe category, the likelihood of occ';rrence is extremely low and so the associated risk is low.
(Hodge Affidavit, p. 7; EIA, p. 34).
19.
The Staff has analyzed an extra severe collision on overturn acci-dent.
By virtue of cask design, massive rupture and sub. sequent releases are precluded.
(EIA, p. 36).
2235 056
. 20.
The Staff has furthe: analyzed the consequences of this accident presuming cask failure by breach of the closure head seal by failure or by the cask lid bolts being cheared off.
The Staff postulated some creep rupture of the fuel cladding.
(Hodge Affidavit, p.10; EI A, p. 37).
21.
Doses due to this postulated accident were calculated.
The maximum individual whole body dise commitment is estimated as 0.28 rem. The population whole body dose commitmer.t for Population Center B is estimated at 370 person-rem. This population dose would mean 0.04 latent cancer fatalities or essentially no health effect.
(Hodge Affidavit, pp.10-11; EIA, pp. 37-38).
- 22. A spent fuel transportation accident may bring about scme reduction in shielding capability but, by regulation, such reduction shall not be sufficient to increase the external radiation dose rate to more than onc rem per hour at three feet from the external surface of the package.
(Hodge Affidavit, p. 12.)
23.
By virtue of this shielding requirement, the distance at which the dose rate would be 10 mrem /hr is estimated to be about 30 m.
It is unlikely that the general public woJld acquire significant doses under these circumstances, (Hodge Affidavi t, p.12. )
2235 057
. 24.
Any delay in the shipment of spent fuel would not produce significant effects. The health effects of any delay caused by an accident are not significant.
If a delay is caused by a stop of the cask vehicle because of a traffic jam in a high density population area, the population dose would be about 0.2 person-rem and the maximum individual dose would be about 0.015 rem. These doses would not result in any readily discernible health effects and thus would not be unacceptably large.
(Hodge Affidavit, pp. 12-13).
2235 058
)
UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
DUKE POWER COMPANY
)
)
Docket No. 70-2623 (Amendment to Materials License
)
SNM-1773 for Oconee Nuclear Station
)
Spent Fuel Transportation and Storage
)
at McGuire Nuclear Station
)
Affidavit of Brett S. Spitalny and R. Daniel Glenn 1.
Our names are Brett S. Spitalny and R. Daniel Glenn.
We have prepared statements of professional qualifications which are attached to this affidavit.
2.
This affidavit addresses Carolina Environmental Study Group (CESG) Contention 1 and Carolina Action Contention 1 which reads as follows:
Shipment of Oconee spent fuel to McGuire for storage is unacceptable as compared to other alternatives:
a.
Modification of the existing Oconee spent fuel pools to provide additional storage capacity.
b.
Construction of a new separate spent fuel storage facility at the Oconee site.
c.
Construction of a new and separate spent fuel storage facility away from the Oconee site, but other than McGuire.
The contention by Carolina Environmental Study Group (CESG) and Carolina Action (CA) suggests that the staff did not adequately examine the alternatives to the proposed action. The contention focuses on two 2235 059
. options available to Duke Power:
- 1) the modification of the existing pools at Oconee, and 2) the construction of a new facility either on or away from the Oconee site. We adopt as part of the basis for our affidavit, the analysis contained in the Environmental Impact Appraisal (EIA), (December,1978).
The first alternative, that of modification of the existing pool may be accomplished by three means, a) physical expansion of the pool, b) reracking with high density stainless steel racks, and c) reracking with neutron absorbing materials (poison racks).
The physical expansion of the Oconee pools is not possible (EIA p. 52).
The existing pools, both the Unit 1 and 2 shared pool and the Unit 3 pool, were not corstructed with the intent of expansion and therefore there is no capabilit/ to breach the integrity of the pools.
Since the Oconee 1 and 2, and Oconee 3 pools contain spent fuel, such an action is not feasible (EIA p. 52).
Another means of increasing capacity in the existing pools is to install hi~gh density stainless steel racks. Again, as evaluated in the staff's EIA, this action was one of two viaDie alternatives available to the applicant (EIA p. 52).
Experience had indicated that the time required to design, contract and procure the racks coupled with the time to license the action, made this alternative less attractive than its counterpart, transshipment (EIA, Chapter 9). Subsequently however, the applicant has requested, and received, an expedited delivery date from Combustion Engineering for the acquisition of these racks.
2235 060
. On February 2, 1979, Duke Power submitted an application to the NRC staff for the approval to install high density stainless steel racks in the Oconee Units 1 and 2 shared pool. The staff is trying to accommodate G t expedited schedule requested by Duke and provide Duke Power with a completed review by early June 1979, as presently scheduled.
This licensing action (reracking) will increase the storage capacity of the Oconee 1 and 2 basin from 336 to 750 assemblies. This increase of 414 assemblies will provide some relief to the immediate problen of a spent fuel storage shortage capacity at Oconee. However, it will not solve Duke Power Company's fuel storage problem. An additional measure is still needed. Duke has indicated that it still envisions transshipment as the preferable alternative (Duke Power Application to rerack Oconee I and 2, February 2,1979). The staff has indicated that both alternatives, transshipment or reracking, were feasible, and that neither alternative imposed any undue risk or significant impact on the quality of the human environment.
As evidenced by the application of February 2,1979 from Duke Power to rerack the Oconee Units 1 and 2 spent fuel pool, the consideration of reracking as an alternative to transshipment has become an additional measure actively being pursued by the applicant to ameliorate the shortage of spent fuel capacity.
The last option open to Duke for modifying the pools, is one of incorporating neutron absorbing materials (poison racks) to increase 2235 061
. the density of assemblies. This alternative was considered in detail in the Natural Resources Defense Council (NRDC) Contention 3c and 3d (B. Spitalny and R. Glenn affidavit).
In general, the staff would agree that the use of poison racks might be considered a reasonable means of ameliorating the shortfall of storage space, but in this situation is not considered cost-effective.
Due to the timing required in procuring and licensing this option, the shipment of spent fuel will still be required to accommodate the installa-tion of poison racks. The method of contending with the shortage of storage space at Oconee by transshipment and reracking as proposed by the applicant, has been shown to be cost-effective and results in negligibly small, and, therefore insignificant impacts on the quality of the human envi ronment.
The alternative of constructing a new and separate storage facility has also been addressed by the contention as being inadequately evaluated.
The economic consequences of constructing a new and separate storage facility remain constant regardless of whether the site is at Oconee, or at some other location.
Speaking independently of site selection, the construction of an Independent Spent Ftel Storage Installation (ISFSI) by the applicant was evaluated in this licer sing action's environmental impact appraisal (EIA pp. 50-52).
That evaluation concluded that the delays associated with licensing, construction and testing of such a facility would not allow completion in time to solve the immediate storage needs at Oconee. Additionally, 2235 062 the financial burden potentially passed on to the ratepayers is of proportion as not to be overlooked as an incidental cost.
In comparison, transshipment to McGuire and reracking Oconee Basin 1 and 2 would meet these immediate storage needs.
The installation of poison racks, would, however, allow sufficient time to construct a separate facility before Oconee again gets to a point that spent fuel storage space would be a problem. This alternative will not preclude the shipment of fuel hcwever, and subsequently does not result in an option advantageous to the one chosen by the applicant.
(Spitalny and Glenn affidavit, NRDC 3c & d)
Use of an onsite, but separate spent fuel storage installation would not significantly reduce the total dose received from similar shipments to offsite storage installations such as McGuire Nuclear Station.
The transshipment of one spent fuel assembly from Oconee to McGuire is estimated to result in a cumulative dose of 0.45 man-rem.
Of this total, 0.4 man-rem is due to loading and unloading of the shipping cask.
Since an onsite storage installation at Oconee could not be connected to the existing basins, it would still be necessary to make similar transfers using a shipping cask.
Thus, the dose savings is only 10 percent of the total received or about 15 man-rem, largely attributable to the reduced dose to the truck drivers. (EIA pp. 29-32, Spitalny and Glenn affidavit, NRDC 3c and d.)
2235 063
. Recent studies by utilities and confirmed by the Department of Energy have indicated that costs for constructing and operating a facility of this type range upward to approximately $30,00C per assembly. This is sharply contrasted by the approximate cost of shipment at $2,000 per assembly. (EIA, Table 10-1, P.58; DOE /EIS-0041-D, p.II-27.) To assure adequate spent fuel storage capacity for its operating reactors, the applicant has increased the size of spent fuel storage basins at those reactors it'has presently under construction.
The potentially small reduction of exposure achievable by building separate storage facilities at Oconee coupled with the large additional costs involved do not support construction of such facilities as in the best inte. rest of the applicant or its ratepayers at this time.
(EIA pp. 50-52; Spitalny and Glenn affidavit, NRDC 3c and d.)
In regards to a new and separate facility at a site other than Oconee or McGuire Nuclear Stations, little or nothing is to be gained over transshipment to McGuire.
In fact, the proposal suggested in CESG's and CA's Contention 1(c) would result in requiring the shipment of spent fuel in a manner identical to that of shipping to McGuire. The intervenors have suggested an option which they oppose as presented in their Contention 2.
This action could most likely also result in greater environmental impacts due to construction of such a facility offsite and the development of land which presumably would have to be acquired by Duke Power.
This would result most likely in additional increased costs to the utility and ratepayers. An indirect cost would also be incurred 2235 064
. in that the construction of a new facility fails to take advantage of economic commitments already made by the applicant.
Duke's proposal, on the other hand, takes advantage of storage capacity at McGuire for which monetary commitments have been made.
In summary, for the foregoing reasons, we have determined that the proposal to transship is an environmentally sound option, with negligibly small, and, therefore insignificant impacts.
In general, the use of neutron absorbing (poison) racks is an accepted practice.
However, in this case it may not be the optimum alternative.
Additionally, the construction of a new facility either on or away from the Oconee site has been shown, like poison racks, to not be cost-effective. Although we find these alternatives technologically feasible, they are not preferred alternatives when compared to the proposed action to transship and store Oconee spent fuel at McGuire.
We hereby certify that the above statements are true and correct to the best of my knowledge and belief.
$. S o M ~.
Brett S. Spitaihy R. Daniel Glenn Subscribed aid sworn to before me this // 7.3 day of May, 1979 2235 065
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(R "Cbth flotary Public My Commission exoires 'lc L'c4 /,/98 L
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
DUKE POWER COMPANY
)
Docket No. 70-2623 (Amendment to Materials License
)
SNM-1773 for Oconee Nuclear Station
)
Spent Fuel Transportation and Storage )
at McGuire Nuclear Station)
)
Affidavit of Brett S. Spitalny I, Brett S. Spitalny, being duly sworn to depose and state:
1.
I am the Project Manager for the McGuire/Oconee spent fuel transportation and storage proposal, Office of Nuclear Material Safety and Safeguards.
2.
I have prepared a statement of professional qualifications which is attached to this affidavit.
3.
This affidavit addresses Carolina Environmental Study Group (CESG)
Contention 3 and Carolina Action (CA) Contention 3 (CESG and CA Contention 3) which reads as follows:
Factors set forth in items 1 [CESG-Contention 1 &
Carolina Action-Contention 1] and 2 [CESG-Contention 2 & Carolina Action-Contention 2] above require the preparation of an Environmental Impact Statement because the proposed action is a major federal action of the Commission significantly affecting the quality of the human environment.
CESG and CA Contention 3 states that the Staff should have prepared an Environmental Impact Statement in lieu of an Environmental Impact Appraisal because the proposed action will have a significant adverse effect 2235 066
-, on the quality of the human environment. CESG's and CA's stated basis is that the impacts from transportation will impose and unac.'9 table burden of radiation dose to the public as a result of routine and non-routine operation, and that the Staff has not properly evaluated specified alterna-tives to the proposed action.
This affidavit further addresses CESG and CA Contention 3 with respect to the factors set forth in CESG Contention 1 and CA Contention 1 and whether those factors demonstrate that the proposed action is a major federal action of the Commission significantly affecting the quality of the human environment such that preparation of an environmental impact statement is requi red.
Resolution of this contention is necessarily dependent directly on the resolution of CESG Contention 1 and CA Contention 1.
My affidavit and the affidavits of R. Daniel Glenn. Dr. M. Parsont, and Dr. J. Nehemias, as wall as the Environmental Impact Appraisal (EIA (December,1978) show (1) that the environmental impacts from the proposed action are negligibly small and therefore insignificent and, 2) that there are no preferred alternatives to the Applicant's request to ship Oconee spent fuel to McGuire for storage when compared to other alternatives.
I adopt the material set forth in the EIA pertinent to the CESG Contentions 1 and 2 and CA Contentions 1 and 2 as part of my testimony and affidavit in this case.
2235 067
. Although Duke Power Company has applied for an amendment to modify the Oconee spent fuel capacity by re-racking, modification of the existing Oconee spent fuel pools to provide additional storage capacity is less preferred on an economic basis.
Modification of the Oconee pool is rougly comparable to the request to transship Oconee fuel to McGuire with respect to radiation exposure from routine operations, although neither activity has other tnan a negligible environmental impact including the impacts of radiation dose.
Transshipment and storage of Oconee fuel at McGuire has negligible or no measureable environmental impact, and, therefore, certainly no signi-ficant environmental impact and far outweighs the construction of a new and separate spent fuel storage facility at or away from the Oconee site from a time, cost, and environmental impact standpoint. This conclusion is based on several factors.
The radiation doses from transshipment and storage at McGuire, although extremely low, would be comparable to transshipment to a new and separate spent fuel pool if constructed at the Oconee site. The economic costs of such a new, separate pool at the Oconee site would exceed many times the transshipment proposal.
(EIA, p 49-59, EIA, Ch. 5)
The time required to design, license and construct such a new, separate spent fuel facility exceeds the time available to Duke by a number of years. (EIA, pp 49-59) Construction of such a new, separate spent fuel facility at either the Oconee site or at another site other than the McGuire site most likely would result in greater environmental impacts from construction, where the transshipment 2235 068
. option has been shown to have negligibly small, and therefore, insignificant environmental impacts. Environmental impact.; of such construction of a new, separate spent fuel pool facility at the Oconee site or at another site other than at McGuire, since the impacts of the proposed action are insignificant and construction of such a separate spent fuel pool is not a reasonable alternative, have not been, and are not required to be evaluated in this case other than in a general sense for purposes of this affidavit.
As the NRC Project Manager of this licensing action, I have directed and taken part in the preparation of the Environmental Impact Appraisal (EIA) in support of the Staff's negative declaration (43 FED. Reg. 61057).
The Staff's EIA has considered all facts that are material to this issue and concludes there are no significant impacts from the proposed action.
I have prepared testimony with respect to CESG Contention 1, alternatives, and CA Contention 1, alternatives. The EIA clearly demonstrates that the proposed action will result in negligible or insignificant impacts with respect to air, aquatic, and terrestrial environs.
Impacts from occupational exposure and the dose to the general public are insignificant. (EIA, pp 29-32)
The affidavits of Messrs. Hodge, Glenn and Lake confirm that the burden of radiation dose as a result of routine and non-routine operations (CESG and CA Contention i) is also insignificant.
In addition, the Carolina Environmental Study Group (CESG) and Carolina Action (CA) have failed to point out where the Environmental Impact Appraisal (EIA) is in error, or how the matters described in the document constitute 2235 069
._ (1) significant impacts or (2) a major Federal action of proportions significantly affecting the quality of the human environment.
- Indeed, Carolina Action has failed to provide any supporting factual basis for its Contentions 1 and 2 in response to discovery requests of the parties.
Based on my analysis, which has considered the EIA and the evidence offered in evaluation of the factors described in CESG Contentions 1 and 2 and Carolina Action Contentions 1 and 2, I have determined with respect to CESG Contention 3 and CA Contention 3 that the impacts from the proposed action will be negligibly small, and therefore, insignificant; and, consequently, the proposed action to transport and store 300-270 day old Oconee spent fuel assemblies at McGuire constitutes an insignificant effect on the quality of the human environment.
I hereby certify that the above statements are true and correct to the best of my knowledge and belief.
ho *M1, W
Brett S. Spitalny 4
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Subscribed and sworn to before me this llId day of May,1979 g,if(*(gc f
/ fQ SSbb 0 0 Notary Public My Commission expires: ')cOz//ff'd.
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
DUKE POWER COMPANY
)
)
(Amendment to Materials License
)
Docket No. 60-2623 SNM-1773 for Oconee Nuclear Station
)
Spent Fuel Transportation and Storage )
at McGuire Nuclear Station)
)
AFFIDAVIT OF BRETT S. SPITALNY I, Brett S. Spitalny, am employed by the Nuclear Regulatory Commission in the Office of Nuclear Material Safety and Safeguards, Fuel Reprocessing and Recycle Branch, as the project manager for the amendment to materials license SNM-1773.
I have prepared, or assisted in the preparation of, the NRC Staff's " Environmental Impact Appraisal Related to Spent Fuel Storage of Oconee Spent Fuel at McGuire Nuclear Station - Unit 1 Spent Fuel Pool", Docket No.' 70-2623, dated December, 1978.
That document is true and correct to the best of my knowledge.
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__Brett S. Spitalny I Subscribjdandsworntobeforeme this // Lday of M4y,1979.
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Notary' Publi c My Commission Expires:
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UNITED STATES OF AMERICA NUCLEAR REGULATORY CCyMISSICN BEFCRE THE ATCMIC SAFETY AND LICENSING 3 CARD In the Matter of
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)
CUKE PCWER CCMPANY
)
)
(A,menament to Materials
)
Docket No. 70-2623 License SNM-ll73 for Cconee
)
Nuclear Station Scent Fuel
)
Trans;cetaticn and Storage
)
at McGuire Nuclear Staticn)
)
AFFICAVIT CF OR. JCHN V. NEHEMIAS I, Dr. Jchn V. Menenias, being duly sworr, do depose and state:
1.
I am a Senior Health Physicist in the Division of Site Safety and Environmental analysis, U.S. Nuclear Regulatory Com. mission (NRC).
2.
I have prepared a statement of professicnal qualifica:icos which is attached to this affidavit.
3.
This affidavit addresses in part, Natural Rescurces Defense Council Ccntention 4(a).
I nerecy certify that the above statements are : rue and accurate to the best of my kn wledge and belief.
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Dr. John V. Nehemias Subscri:ed and sworn to
- efere me this/, daj c' 2235 072 "aj, 1979.
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Contention J(a):
ALARA can be acnieved by cn-site expansion of scent fuel cool stcrage capacity at Cconee, including building another spent fuel ccol.
Ibis contention addresses the fact that the proposed transshipment of Cconee scent nuclear fuel to v Guire Nuc! ear Station for storage c
will involve scme radiaticn exposure to the ;utlic and to acrkers involved in the transshi; ment.
- ntervencr's point is that these radiation ex:Osures cculd te entirely eliminated by simply expanding the spent fuel s:crage cacacity at Oconee, either by re-racking the present spent fuel pcol to cermit storage of a larger number of fuel elements, or by cuilding another spent fuel pcci at Oconee.
We understand that re-racking the present spent fuel pool at Cconee would provide cnly enough additicnal fuel storage capacity to accommodate about two years' supply of spent fuel. At er before that time, additional spent fuel storage capacity would be recuired, either by building another spent fuel pool at Ccenee, cr by trans-snicping the spent fuel, utilizing available scace at v Guire.
c Thus, re-racking the present scent fuel pcol at Cconee would delay only slightly tre necessity to transfer spent fuel to another location, and wculd not eliminate that necessity.
(a) Re-racking the cresent Oconee scent fuel cool Ex;erience with pricr one-time fuel col cdifications of this kinc indicates that sucn re-racking cper
- t-n a
caused an averaga of about 3 man-rems to the 1:orkers involv2d.
The highest cunulative occucational dns2 feca such o?arations has been 20 man-reas, which.:as incurred during the ro.-;ification of the scent foal neol at Eaddam
- ?ec k.
See attached Table I.
"o public exposure should resul t.
~
Cuke power has estinated, we believe conservatively, that occucational doses during modification of the s, cent fuel pool at Oconee '.zould be 125 nan-rees.
Rased on experience with similar nodifications at other plants, :;e '.ould ex?ect that actual doses nay be scnenhat lower. ' e have recuested the applicant to prepare a r. ore realistic estinate, ar.d to crovide additional information about Scw dose rates and occupaticnal doses will be kept as icw as is reasonably achiavable. (AL111)
It seems reascnable to assune that the likely cccupational radiation exposure frcm the re-racking oceration at Cccr.ee would be in the range of 20 to 30 nan-rems.
(b) Transshi: ment of Oconee scant fuel to cGuire The radiaticn deses to the ;.biic resuIf irg fr:n the r,.,,. c. s. e. - - =. e. s.
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.ws 2235 074
3 portion of the total dose could be eliminated by construc-tion of a new spent fuel pool at Oconee.
The principal radiation dose resulting from this trans-shipment, however, '.,ould be delivered to.:orRers, and is estimated at 16 can-r2ms to drivers for 300 shipments.
In addition, occupational dose which would not be eliminated by construction of a new spent fuel pool at Oconee, results frcn activities related to transfer of the spent fuel into a shipping cask, movement of the cask fecm the spent fuel pool to the new location, and transfer frca the shipping cask to the new storage facility.
It seems reasonable to assume that the likely occupational radiation exposure from transshiprent to McGuire would be in the range of 20 to 30 Tan-rems.
(c)
Construction of a new scent fuel cool at Oconee The actual activities involved in construction of a new spent fuel pool at Oconee would not involve any radiation exposure to the public, or to the cersonnel involved.
':owever, when the new spent fuel cool has been cons tructed, as in the case o' t-irsshi ment to 'cGuire, fuel trars fer
. uld still be re;uired.
The spent f;el.:uld have to be tearsf arred, are fuel assa-biy at a time, frca the existir.g 2235 075
scent fuel pool into a shipping cask, moved in the cask frcm the scent fuel pool to the new location, and transferred from the shipoing cask to the new storage facility.
These activities will involve radiation exposures to the oerscnnel taking part in the transfers.
Although no estimate of the potential total dose fron these operations has been prcvided, it seems reascnable to assume that such doses will also be in the range of 20 to 30 man-rems.
The total man-rem doses projected to result from the three actions being considered are estimated to be in the same ceneral dose range.
Therefore, there would be no basis for concluding tha t any o f the three is clearly to be preferred frca the point of view of radiation risk, nor that any significant dose saving would be expected to result frcm the selection of any one of the three.
We conclude that the exposures likely to result fecm the transship-ment of Cconee spent fuel to McGuire, as described by the a?piicant, would be ALARA.
Each aspect of the proposed actions have been considered from the point of view of keeping radiation ex;osures ALAR;., eliminating unnecessary exposures, and taking all reascnable precauti:ns to reduce exp;s;res.
'le have transmi tted to the
?:olicint additicnal reques:s for inf:r ? ti:n :n t're :r:;csed 2235 076
5_
re-racking of the spent fuel poci at Oconee, in ceder fcr us to be able to reach the conclusion that it too wculd ce consistent with ALARA censiderations. Similarly, if the acolicant pecccses to ccnstruct a new spent fuel stcrage facility at Occnee, we nill review any such application with regard to ALARA consideraticns.
'While the NRC has not issued specific guidance related to ?LARA censideraticns involved with fuel storage or transfer, we have issued Regulatory Guides 3.3, "Information Relevant to Ensuring That Occupational Radiation Ex;csures at Nuclear ?cwer Stations Will 3e As Low As Is Reasonably Achievable," and 3.5, " Operating Philosochy for staintaining Cccucational Radiation As Law As !s Reascnably Achievable." These guides spell cut our ALARA philos 0chy and descrite the ALARA approach to reduction of exposures. These consideraticns nave been applied in cur review of the apclicant's prccosals regarding scent fuel transfer and storage at Ocenee and v Guire.
c I hereby certify that the above statements are true and ccrrect to the best of my knowledge and belief.
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Dr.
crn /. 1enemias s
Subscribec and s.vorn to before le this,> day cf
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John V. Nehemias PROFESSIONAL QUALIFICATIONS Radiological Assessment Branch Qivision of Site Safety and Environmental Analysis I am a Senior Health Physicist in the Radiciogical Assessment Branch, Division of Site Safety and Envircnmental Analysis, Office of Nuclear Reactor Regulation.
My formal education consists of study in Physics at Rensselaer Polytechnic Institute where I received a B.S. in 1948 and at Columbia University where I received an A.M. in 1949.
I received a Ph.D. in Environmental Health (Radiological) frca the University of Michigan in 1950.
Before joining AEC/NRC, I served three years at Brookhaven National Laboratory as a health physicist, six years at the University of Michigan as health physicist and assistant director of a radiation effects laboratory, and three years as Director of Radiological Health Surveys for the National Sanitation Foundation.
In the latter position, I
/,
designed, organized, and directed the environmental survey for the
~,
Enrico Fermi nuclear plant.
I joined the AEC in September 1960, as a health physicist in the Office of Health and Safety. My principal duties there related to development of radiation protection standards. With the two exceptions noted belew, I have continued with AEC (and NRC) since that time. My principal responsibility was in the development of Standards until September 1974; during most of those years I served as a branch chief-through several name changes and reorganizations-most recently as Chief, Occupaticnal Health Standards Branch, March 1972 to September 1974.
Since September 1974, I have served as Senior health physicist in the Radiological Assessment Branch. My principal function is the review of power reactor applications, both at the construction permit and operating i
license stage, to determine the adequacy of proposed occupational radiation protection programs and the related efforts proposed to assure that occupational radiation exposures will be maintained as lcw as is reascnably achievable.
2235 079
u~
~
O From June 1953 to September 1965, I took a leave of absence frca AEC and served as principal member of the Occupational Safety and Health Division of the International Labor Office in Geneva, Switzerland.
My work was principally in the development of international standards.
In December 1971, I was transferred to the Criteria and Standards Divisien, EPA, serving as Chief, Criteria and Standards Branch, until my return to AEC in March 1972.
I have published about 40 technical articles in professicnal journals and other publications in the general areas of low-level counting, environmental monitoring, radiation effects on biological systems, and control of occupational radiation exposure.
I have been a Certified Health Physicist since 1960, and am a Charter member of the Health Physics Society and of the Saltimore-Washington Chapter.
C 2235 080 W
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g UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
DUKE POWER COMPANY
)
)
(Amendment to Materials License
)
Docket No. 70-2623 SNM-ll73 for 0conde Nuclear
)
Station Spent Fuel Transportation
)
and Storage at McGuire Nuclear
)
Station
)
AFFIDAVIT OF DR. MICHAEL A. PARSONT 2235 081
e f
AFFIDAVIT OF DR. MICHAEL A. PARSONT My name is Michael A. Parsont.
I am Chief of the Radiological Health Standards Branch of the NRC Office of Standards Development.
As part of my duties in this position I am responsible for directing an NRC program to evaluate and assess the radiological health impacts to the public from NRC proposed and licensed facilities.
A copy of my Professional Qualifications is attached.
My affidavit responds to Petitioner's contention 4 Part b. which refers to residual health risks from the dose resulting from transshipment of spent fuel from the Oconee facility as major costs tipping the balance against the proposal to transship and store Oconee spent nuclear fuel in the McGuire, Unit 1 spent fuei pool.
Contention 4 is as follows:
The proposed action increases the exposure to radiation cf workers and the general public beyond what is ALARA.
a.
A'_ ARA can be achieved by on-site expansion of spent fuel storage capacity at Oconee, including building another spent fuel pool.
b.
The residual health risks which remain even if the present NRC regulations on exposures to workers are met are major costs of the proposed action which tip the balance against the proposed action ('r.77-85).
In the context of my testimony, Residual Health Risks from exposure to ionizing radiation are genetic risks and may be exp essed in subsequent generations as congenital abnormalities, cons titutional and degenerative diseases and overall ill-health (other ill1 esses having 2235 082
. some degree of genetic determiration).
In addition, the cancer risk from exposure to ionizing radiation is of concern to Petitioner.
My response to this part of Contention 4 is based on the following considerations:
1.
Somatic risks (i.e. the risk of cancer) and a significant portion of the genetic risks of health effects from ionizing radiation are directly and linearly proportional to radiation dose and dose rate.
2.
There are 2 viable options, both of which will be taken, for Duke Power Company to resolve its immediate shortfall in spent fuel storage capacity--these being the expansion of storage capacity of Oconee Units 1 and 2 Spent fuel pool by re-racking and at other nuclear stations owned by Duke Power.
I have estimated the genetic effects for the range of doses involved in the 2 options for resolving the Oconee spent fuel storage capacity shortage based on the 1972 National Academy of Science Report of the Co'imittee on the Biological Effects of Ionizing Radiation, BEIR.I)
(The recently published update of the BEIR Comittee, BEIR-III,2) p esents information on genetic effects which does not significantly differ from the 1972 BEIR Report.)
I have ostimated risk to cancer from BEIR-III data because it represents more recent considerations of radiation effects.
- 1) Advisory Committee on the Biological Effects of Ionizing Radiation.
"The Effects on Populations of Exposure to Low Levels of Ionizing Radiation," National Academy of Sciences-National Research Council, Washington, D. C. November 1972.
- 2) Committee on the Biological Effects of Ionizing Radiations.
"The Effects on Populations of Exposure to Low Levels of Ionizing Radiations, National Academy of Sciences--National Research Council, Washington, D. C.
1979.
2235 083
. The range of doses used in my calculation of the genetic effects is based on several considerations as follows:
- 1. The upper end of the range of population dose is based on estimates of the applicant. This was presented as 150 person-rem in Table 10-1 of the Environmental Irrpact Analysis for expanison of the Oconee spent fuel storage pool capacity by re-racking.
This estimate was subsequently reestimated at 125 person-rem.
- 2. Mr. Glen of Battelle Northwest Laboratories estimates that re-racking could start at about 60 person-rem but would unlikely range upward to 150 person-rem.
- 3. Dr. Nehemias states that, bcsed on actual experience, re-racking dose would be closer to 20 person-rem.
- 4. The applicant's recc timated re-racking dose and the population dose from transshipment (120 nerson-rem) are effectively the same from the standpoint of effects.
Therefore, the range of doses from the 2 options extend from 20-150 person-rem based on whichever information is accepted.
In addition, for perspective, these doses are quite small (.004%.03%) compared to the expected normal operation occupational exposure at Oconee 1, 2 and 3 over the assumed 30 year facility lifetime.
The estimated genetic effects from BEIR ) and from the re-racking and I
transshipment options are presented in Tables I and II, respectively.
The range of doses brackets the dose estimates given above.
Although there is general agreement that a significant proportion of somatic and genetic health risks are directly proportional to the magnitude of the radiation dose, there is controversy over the magnitude of the dose-effect response at low-radiation dose and dose rate.
This controversy is based on the results of studies of various exposed populations.
These studies report that exposure to low-level radiation 2235 084
. may be about an order of magnitude (about 10 times) more effective in producing health effects than the estimates given in the BEIR Report.I)
Applying the factor of 10 to the estimates of ganetic effects given in Table II results in a maximum equil'brium estimate of 0.3 effects.
In my opinion, because of the small numoer of genetic effects, even if the BEIR estimates were low, this action does not represent a major genetic health cest.
Although contention 4 does not specifically refer to socratic effects, I have calculated the range of total and fatal cancers which might result from the options considered.
I have used the risk estimate presented in BEIR-III which are summarized in Table III.
The estimates for the option are given in Table IV.
For a single exposure the maximum estimate of total cancers, assuming BEIR-III was low by a factor of 10, would be 0.8, and the estimate for fatal cancers would be 0.2.
2235 085
Table I.
ESTIfiATED GENETIC EFFECTS 6 live Estimated Risk C
Disease Classification Natural Effect per 10 Incidence births al of 5 rem per per 106 person-rem (c) cc (per 106 live generation (b)
O births) m vn First Generation Equilibrium First Generation Equilibrium m
N Dominant diseases 10,000 50 to 500 250 to 2500 6 to 60 30 to 300 Chromosomal and relatively very slow relatively very slow recessive diseases 10,000 slight increase slight increase Congenital anomalies 15,000 Anomalies expressed later 10,000 5 to 500 50 to 5000 0.6 to 60 6 to 600 Constitutional and 15,000 degenerative diseases Human Illness having genetic component 0.25 to 250 0.03 to 30 TOTAL 60,000 60 to 1000 300 to 7750 7 to 120 36 to 930 Risk per 106 people 1,200(d)/ year Geometric fiean 29 183 (d} From the 1972 BEIR Report1/ able 4 p. 57.
The Human Illness entries (.005x50 and.05x5000) and new totals are T
my estimations.
( } A generation is assumed to be 30 years.
(C Risk per 106 person-rem = (cases /106 live births) x (30 years /5 rem) x (4 x 106 live births / year per 2 x 108 people) = 0.12 x cases /106 live births.
(d) Cases /106 live births x (4 x 106 live births per year / 2 x 108 people),
Table II.
GEflETIC EFFECTS COMPARISON FOR TWO OPTIONS Dose Genetic Effects Total Genetic Option 1/
(Person-rem)
First Generation Effects at Equilibrium 1
20-150 0.0006
.004 0.004 - 0.03 2
120 0.003 0.02 1/ ption 1 is reracking at Oconee.
0 Option 2 is transshipment to McGuire.
2235 087
TABLE III Co=parative Lifetime Cancer Risk Estimates for the General Population from Exposures to Low-Dose, Low-LET Radiation, Single Exposure
- and Continuous Exposure **, Both Sexes Co=bined af Source of Continuous Esti=ates Single Exposure exposure (per million pooulation exposed per rad)
BEIR 1979 Incidence Relative Risk 636-1031 592-946 Absolute Risk 268-399 (525)b 254-373 (490)
Mortality Relative Risk 177-353 150-293 Absolute Risk 70-124 (157)b 68-119 (141)b BEIR 1972 Factors **
Mortality Relative Risk 621 568 Absolute Risk 117 (270)b 115 (256)b UNSCEAR 1977 Mortality 100 100
- The BEIR 1979 single-exposure esti= ate was based on a 10-rad dose and was divided by 10 for conparison with the other values; the estinate for con-tinuous exposure is based on a lifeti=e exposure of 1 rad / year.
- BEIR 1972 post-natal, age-specific risk factors used with 1969-1971 life-tables, with plateau extending throughout the years of life re=aining af ter irradiation, esti=ste (b) in the 1972 BEIR Report.
The average age of the 1969-1971 life-tables is older than that of the 1967 U.S. population used in the 1972 BEIR report.
For this reason, the numbers obtained here for continuous exposure are larger, on a per rad basis, than those obtainable from Tables 3-3 and 3-4 of the 1972 BEIR report.
/ aken from BEIR-III, Table 5, p.342 T
Geometric Mean (my addition) 2235 088
Table IV. CANCER CASE COMPARIS0N FOR TWO OPTIONS (Single Exposure)
Dose Total Option 1/
(Person-rem)
Incidence Fatal 1
20 - 150
.01
.08
.003
.02 2
120
.06
.0002 00ption 1 is reracking at Oconee.
Option 2 is transshipment to McGuire.
2235 089
I hereby certify that the above statements are true and accurate to the best of my knowledge and belief.
/.
/
Or. Michael A.
arsont Subscribed and sworn to before me this lith day of May,1979, alaw
-Not ry Public My Commission Expires:
Mf,7 2235 090
PROFESSIONAL QUALIFICATIONS of Dr. Michael A. Parsont My name is Michael A. Parsont, I am Chief, of the Radiological Pealth Standards Branch in the Office of Standards Development of the U.S.
Nuclear Regulatory Comission.
I have served in this position since November 1978.
In this capacity, I supervise and direct the activities of six staff professionals in areas concerning the determination of health risks and effect from exposure to ionizing radiation, radiation epidemiology and regulation of the use of medical devices and pharmaceuticals containing radioactivity.
In addition I am responsible for developing radiological health standards and guides and for the evaluation and assessment of the radiobiological health impacts on the public # rom proposed and licensed facilities.
Such efforts include the determination of relationships between low-level radiation exposure and health effects from direct radiation and radioactive materials emitted from planned or existing nuclear facilities and from the medical use of radioactive materials.
I am also responsible for directing, coordinating and evaluating technical support research performed by national laboratories and industrial contractors to establish the bases for regulations, standards and guides.
I serve as an advisor and coordinator in radiobiology for technical assistance contracts.
I represent the NRC at international symposia, and other meetings in areas of radiological impact assessment.
From September 1972 until November 1978 I served as a radiobiologist and an environmental scientist on the staffs of the Office of Standards Develop-ment and Nuclear Reactor Regulation, respectively.
In these positions I performed evaluations of the health effects of ionizing radiation; prepared the Radiological Assessment and Radiological Monitoring Sections of Environmental Impact Statements; and performed numerous studies related to the impact of NRC proposed and licensed facilities on the environment.
I received a B.S. in Public Health from the University of California at Los Angeles (1955), a M.S. in Radiology from Colorado State University (1962) and a Ph.D. in Radiation Biology from Colorado State University (1967).
I completed additional undergraduate studies in genetics and endocrinology at the University of California, Berkeley and graduate studies in Sanitation Engineering and Public Health at the University of California at Berkeley and Los Angeles, respectively.
I have more than 13 years of professional experience in Public Health, Radiation Biology, Environmental Sciences, research evaluation and coordination and standards development.
This experience was gained at the Alameda County Health Department, Alameda, California; Sandia Labora-tories, Albuquerque, New Mexico (Aerospace Nuclear Safety); NUS Corporation, Rockville,14ryland (Environmental Studies); and the AEC-NRC.
2235 091
N ')
o UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
DUKE POWER COMPANY Docket No. 70-2623 (Amendment to Materials License
)
SNM-1773 for Oconee Nuclear Station
)
Spent Fuel Transportation and Storage
)
at McGuire Nuclear Station)
)
AFFIDAVIT OF C. VERNON H0DGE, WILLI AM H. LAKE, JR.
AND R. DANIE GLENN Introduction Our names are C. Vernon Hodge, William H. Lake, Jr. and R. Daniel Glenn.
Copies of our professional qualifications are attached.
This affidavit addresses a contention which reads as follows:*./
Transportatior, of spent nuclear fuel from the Oconee Nuclear Station for storage at the McGuire Nuclear Station will create an unacceptable hazard by significantly increasing the radiation doses to persons in the region near the proposed transportation routes between the two facilities, specifically:
(c) There is likely to be an unacceptable incremental burden of radiation dose to persons in the vicinity due to an accident **/ or delay in transit.
- /- This contention is raised by both the Carolina Environmental Study Group and Carolina Action as Contention 2 of " Stipulations" dated October 18, 1978. Only Part (c) is addressed in this affidavit.
Parts (a) and (b) are addressed elsewhere by the NRC Staff.
- / ccident as related to this contention includes the likelihood of
--A a melting or breach of cask accident.
2235 092
T a Discussion Spent fuel is highly radioactive and requires heavy shielding for safe handling.
Massive, durable, heavy casks are required to transport these ma terial s.
The form of spent fuel is the same as that of new fuel except for differences in chemical composition and physical properties due to irradiation - spent fuel pellets containing fission products in both gaseous and solid state. The pellets are clad with light metal and assembled in rods and elements which are tightly inserted into spent fuel casks.
Both the form of the material and the heavy casks in which it is shipped protect against consequences to public health and safety that would otherwise result from transportation accidents.
A spent fuel cask is generally cylindrical in shape and about 20 feet long.
he basic components include a steel inner vessel which contains the fuel elements and spacers or neutrons absorbers to assure nuclear s ubcri tical i ty.
The inner vessel is surrounded by several inches of shielding (dense metal for attenuation of gamma radiation) encased in a steel jacket.
Several inches of hydrogenous material (such as water) for attenuation of neutron radiation surround the gamma shield. A steel outer jacket completes the package. The cask may also be equipped with sacrificial impact limiters to absorb forces involved in impact accidents.
The closed inner vessel is filled with the primary coolant (air, helium, water) to aid in the dissipation of heat generated by radioactive decay.
2235 093
t
, The designs of spent fuel casks are regulated by the Department of Transportation (DOT; 49 CFR Parts 170-189) and by the Nuclear Regulatory Commission (NRC; 10 CFR Part 71). The NRC reviews the designs for certification of compliance with the requirements of 10 CFR Part 71.
The review addresses the capability of the package design under both normal and accident conditions to retain its radioactive contents, to shield the external environment from the radiation of its contents, to dissipate its internal heat to the external environment at a safe rate, and to assure nuclear subcriticality.
In addition, the package design is reviewed with respect to quality assurance in acceptance, operations, and maintenance.
Standards for these aspects are also prescribed in 10 CFR Part 71.
In seeking to protect public health and safety from the effects of trans-portation accidents, the NRC regulations prescribe a performance standard and an acceptance standard for each package of radioactive material.
In the case of a spent fuel cask, the performance standard is a series of tests applied sequentially and the acceptance standard is essentially no release of radioactive material.
It must be recognized that under the test conditions some coolant or gaseous material entrained in the coolant or in the gap between fuel cladding and fuel pellet may be released from the cask.
Release of this material would not be significant to public health and safety; the acceptance standard limits the quantities of such releases to assure that they would not be significant.
2235 094
3
, These casks are designed to withstand, without release of radioactive material in excess of the regulatory limits specified in 10 CFR Part 71.36(a)(2), a severe accident damage test sequence to simulate the effects of severe impact, puncture, fire, and immersion in water as specified in Appendix B of 10 CFR Part 71.
The test sequence includes:
- 1) a free fall from a height of 30 feet onto an essentially unyielding horizontal surface, striking the surface in a position for which maximum damage is expected; 2) a free drop of 40 inches striking (in a position which is expected to cause maximum damage) the top end of a vertical cylindrical steel bar, 6 inches in diameter and at least 8 inches long, mounted on an essentially unyielding horizontal surface;
- 3) a thermal test in which the cask is exposed to a heat input equivalent to that of an oil fire (1475 F for 30 minutes); and 4) immersion in water to the extent that all portions of the cask are under at least 3 feet of water for a period of not less than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />.
These test conditions make up the design basis accident for a spent fuel cask, meaning that package designs which meet the criteria under the above conditions provide reasonable assurance that the cask will withstand most severe transpor-tation accidents without the release of significant radioactivity.
Spent fuel casks have been subjected to many tests and analyses to find the most vulnerable aspects of the package designs.
Recently, the Department of Energy (DOE) sponsored full-scale impact testing of casks mounted on trucks and a rail car by colliding the vehicles with concrete abutments or speeding locomotives.
In these tests, the casks were not damaged significantly and conclusions were drawn that the abilities of the casks to contain and shield their contents were not impaired 2235 095
1 r in the tests.1 In a full-scale fire test described in the same document, the rail cask was set in a large pool of jet fuel which was ignited and burned for about two hours. After an hour and a half, the lead shielding had all been melted but was still contained. After that time, a small crack appeared in the outer steel skin of the cask and molten lead was slowly expelled.1/ The results of these full-scale tests are interpreted by the NRC staff as data which increases its confidence in the NRC regulations and in the NRC reviews of each cask design submitted for approval.
Spent fuel casks have been allowed in the public transportation system for the past thirty years or 30.
In a recent survey conducted by the NRC, the annual shipping rate for spent fuel in the United States was estimated for 1975 as about 270 shipments per year.2/
As of 1972, about 3600 shipments of spent fuel had been made.3_/ Two accidents to spent fuel casks have occurred during that time. On December 8,1971, a truck carrying a spent fuel cask was overturned on a highway in Tennessee.E The accident was apparently caused by an oncoming tractor-trailer veering into the lane of the cask vehicle on a curved portion (150 foot radius) of the road. The driver of the cask vehicle negotiated about 300 feet of the curve, but lost control of the vehicle.
The vehicle came to rest upside down in the ditch beside the road with the leading end of the cask embedded three feet deep in soft soil.
The cask had skidded about thirty yards along a ditch with the tractor-2235 096
i
, trailer attached. Only minor cask damage was discovered in the initial investigation and no additional damage was discovered in subsequent more detailed inspection.
The driver of the cask vehicle was killed in the accident; no other injuries occurred. An Atomic Energy Commission (AEC) emergency response advisory team from its nearby Oak Ridge Operations Office arrived on the scene within an hour of the accident and determined that no radioactive material had escaped from the cask.
Later, careful health physics surveys also revealed no additional radiation from the cask due to the accident.
The wreckage and cask were removed from the highway and traffic was restored by law officers.
The cask was transported to its destination on the same day, having been delayed by the accident by seven hours.
In the other accident, which occurred February 9,1978, on a highway in Illinois, a truss-type trailer in which a spent fuel cask was being carried experienced a structural failure.E The vehicle was traveling about 50 mph when it struck a sharp road surface heave, causing the top trailer longerons (structural supports) to buckle and the trailer bottom to drop to the road surface, after which the driver maneuvered the vehicle to a stop at the edge of the road.
Early observations, later confirmed by closer visual examination, indicated no visible damage to the cask. No injuries or other property damage occurred.
An Illinois agency responded to the 2235 097
i
, accident.
Radiological surveys showed no abnormal conditions.
The trailer was cut away from the cask and the cask was loaded into another trailer and transported the next day to its destination.
Traffic was restored by law officers; the highway had been closed to traffic for fifteen hours.
The analysis of transportation accidents involving radioactive materials shipments begins with a study of how severe they are, how frequently they occur, and what might be the possible consequences of them.
In 1972, the Atomic Energy Commission (AEC), the predecessor agency to the Nuclear Reculatory Commission (NRC), issued " Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants," WASH-1238.
In this document, accident severity is broken down into five categories.
In order of increasing severity, they are minor, moderate, severe, extra severe, and extreme. The categories are described in Table 6-1 of " Environmental Impact Appraisal Related to Spent Fuel Storage of Oconee Spent Fuel at McGuire Nuclear Station - Unit 1 Spent Fuel Pool" (EIA).6/ This table includes estimates of the frequencies of these accident severities.
These frequences become more meaningful when they are applied to the proposed shipping campaign. Assuming this campaign consists of 300 shipments in one year of 170 miles each, the number of years between accidents of the same severity is given by:
Minor 14 years Modera te 50 years Severe 2500 years Extra Severe 25,000,000 years Extreme 1,000,000,000 years 2235 098
. In WASH-1238, the staff considered that spent fuel casks would meet the regulatory standards for containment, shielding, and criticality in accidercs classed as minor, moderate, and severe. The DOE sponsored accident tests described above indicated that the particular spent fuel cask tested would meet these standards in accidents of greater severity.
Accident scenarios of greater severity than the severe category have been considered in both WASH-1238 and in the EIA. The evaluated consequences in these documents (Appendix B of WASH-1238; Section 6.1 of EIA) lead to the conclusion that the risk to public health and safety from transportation accidents involving radioactive materials shipments is small.
The contention refers to a melting or breach of cask accident.
Such an accident would belong to either the extra severe or extreme category described above. As discussed above, the probability of an accident severe enough to cause either of these types of package damage is extremely small.
The joint probability of both melting and breach of cask occurring in the same accident woula be even smaller.
In many accident scenarios considered, the wreckage resulting from a collision serves to shield a package from fires.
Extensive quantities of fuel are required to sustain fires capable of elevating to high values the temperatures of packages that happen to be located nearby.
Even in the event of the accidents postulate 4 above, it can be shown that mel ting of the nuclear fuel is not credible.
2235 099
. Considering for the moment the source of heat within the cask, it should be noted that the maximum internal heat load for a spent fuel cask is limited by an approval condition specific to each cask design.
In setting a maximum acceptable internal heat load, the applicant must demonstrate that the heat can be passively dissipated (that is, without the assistance of active auxiliary heat removal systems that may be mounted on the cask vehicle) from the cask following the accident damage tests discussed above, while the cask meets all the shielding, containment, and sub-criticality requirements of 10 CFR 71.36.
The maximum internal heat load is not sufficient to melt the fuel (uranium dioxide melting temperature exceeds 4500 ) or the fuel cladding (typical cladding material melting temperature ranges from 2600 F to 3300 F).
It is concluded that melting under normal or accident conditions from an internal heat source is not credible.
With respect to external heat sources, the regulatory accident tests include a half hour fire, equivalent to a heat source at 1475 F with an emissivity coefficient of 0.9 radiating to the cask which is assumed to absorb 80 percent of the incident radiant heat which completely surrounds it.
Each spent fuel cask is evaluated against the design basis accident conditions described above to assess the effects of the accident conditions on the ability of the cask to dissipate heat after the sequence of tests.
Authorization to use a cask means that the cask can dissipate such heat and therefore the contents will be maintained at temperatures below the melting temperature.
2235 100
, It has been determined that such an accident would not result in fuel melt but might result in some creep rupture of the fuel cladding.
Creep failure is a temperature time phenomenon evident at elevated temperatures, each material being characterized by its own onset temperature well below its melting temperature.
It is observed as progressive failure under fixed stress and temperature.
Increasing the load on the material or increasing the temperature results in accelerated failure.
Such failure may yield subsequent release of noble gases and possibly small quantities of volatile solid radionuclides such as cesium and tellurium through an assumed breach in the containment vessel. The consequences to public health and safety from such releases are not significant.
For example, the maximum individual whole body dose commitment is estimated in Table 6.3 of the EIA as 0.28 rem.
The population whole body dose comitment estimated in Table 6.3 of the EIA for Population Center B is 370 person-rem.
The average individual dose commitment is estimated as 0.032 rem.
Considering that one million person-rem of whole body population dose results in about 120 latent cancer fatalities,2_/ this population dose would mean 0.04 latent cancer-fatalities, that is essentially no health e f fec t.
Temperatures sufficient to produce creep rupture were observed in an analysis of a cask containing more than one fuel element.
For casks containing one fuel element, such as are proposed to be used in the Oconee-McGuire transfer of spent fuel, significant creep rupture of fuel cladding would not be expected for loss of coolant or fire accident conditions.
2235 101
. The NRC staff has recently examined its regulations on packaging and transportation of radioactive materials.U Transportation accidents of all severities were considered to obtain an expectation value for public health and safety consequences.
Assumed accidents invoiving spent fuel casks shipped at the 1975 rate infer an expected value of about 0.00004 latent cancer fatalities from that year's spent fuel shipping.
Another way to express this result is that if the shipping rate is constant at the 1975 value, one would expect accidents to spent fuel casks to result in about 4 latent cancer fatalities in one hundred thousand years of shipping.
These health effects would not be manifest at the time of any given accident, but might occur within significant fractions, say 30 years, of individual lifetimes after the accident.
Applied to the proposed shipping campaign, and assuming the proposed shipping rate is constant, the expected rate of latent cancer fatalities from accidents is roughly a factor ten smaller than the national value for 1975.
Recognizing that the proposed shipments will not continue indefinitely, the expected health effects must be smaller yet.
The discussion above leads to the conclusion that for all but the most severe transportation accidents, the cask integrity will not be reduced.
That is, one would not expect the cask to be breached in an accident so that a significant quantity of radioactive material could be released into the environment. An accident may bring about some reduction in shielding 2235 102
. capability of the cask.
The regulations require that under the package test conditions specified in 10 CFR Part 71, Appendix B, the reduction of shielding shall not be sufficient to increase the external radiation dose rate to more than one rem per hour at three feet from the external surface of the package (10 CFR 71.36(a)(1)). This dose rate includes both gamma radiation and neutron radiation that might emanate from the cask.
Under these conditions the distance at which the dose rate would be the regulatory limit (10 mrem /hr) for routine exposure at six feet from the truck is estimated to be about 30m (100 ft).
It is unlikely that individuals in the general public would acquire significant doses under such circumstances.
In summary, the effects of a transportation accident involving shipments of radioactive materials are not expected to be significantly different from other transportation accidents.
Finally, the contention refers to an unacceptable incremental burden of radiation dose to persons in the vicinity due to a delay in transit.
If the delay is caused by an accident, persons in the vicinity, whether they are delayed in transit or not, have been considered in the analysis of health effects presented above.
If the delay is caused by a stop of the cask vehicle because of a traffic jam in a high density population area, a population dose of about 0.01 person-rem per hour of delay plus about 0.005 rem / person / hour for persons parked in vehicles along side the cask during the delay would be incurred.
Assuming two persons per vehicle and 2235 103
. four cars beside the catk at an average distance of 3 meters from the truck for three hours, the. population dose would be about 0.2 person-rem and the mcximum individual dose would be about 0.015 rem. Those doses would not result in any readily discernible health ef#ects and thus would not be unacceptably large.
Conclusion Spent fuel casks are designed and certified to contain and shield their radioactive contents during all likely transportation accidents.
Testing, accident experience, and intensive review of cask designs assure us that no significant radioactive releases will occur because of transportation accidents involving these packages.
These considera-tions lead us to the conclusion that an unacceptable incremental burden of radiation dose from transportation accidents involving spent fuel casks is not likely.
In the extremely unlikely event of a release of radiosctivity, the release would be limited to noble gases and possibly small quantities of volatile solid radionuclides such as cesium and tellurium; the incremental burden of radiation dose would not be significant.
In the event of a. delay in transit, the incremental burden of radiation dose is small and acceptable. Furthermore, in view of the very small consequences projected from accidents to all 2235 104
. spent fuel shipments made annually, the risk of consequences to public health and safety from the proposed transfer of spent fuel is acceptably small.
We certify that the above statements are true and correct to the best of our knowledge and belief.
M C. Vernon Hodge 9
M William H. Lake, Jr.
~/'
R. Daniel Glenn Subscribed and sworn to before me this to W day of NAy
, 1979.
' ' Notary Public My Commission Expires: M / /97g.
()
0 ~
2235 105
. If Jefferson, Robert M. :
" Statement for the Senate Subcommittees on Science, Technology, and Space and Surface Transport," Sandia Laborator.ies (August 16,1978).
2] " Final Environmental Statement on the Transportation of Radioactive Material by Air and Other Modes," NUREG-0170, U.S. Nuclear Regulatory Commission, Office of Standards Development (December 1977).
3/ " Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants," WASH-1238, U.S. Atomic Energy Commission, Directorate of Regulatory Standards (December 1972),
P. 61.
4] Chandler, John M..
"The Peach Bottom Spent Fuel Element Shipping Cask Accident, December 8,1971," ORNL-TM-3844, Oak Ridge National Laboratory (July 1972).
5/ Best, Ralph E.:
Letter to C. E. MacDonald, U.S. Nuclear Regulatory Commission, with attached memorandum on the subject:
" Transportation Accident Involving NAC Truss-Type Trailer and NAC-1 Cask Serial Number C," Nuclear Assurance Corporation (February 22,1978).
6f " Environmental :mpact Appraisal Related to S ent Fuel Storage of Oconee Spent Fuel at McGuire Nuclear Station - Unit 1 Spent Fuel Pool," (EIA), Docket No. 70-2623, U.S. Nuclear Regulatory Commission, Office of Nuclear Material Safety and Safeguards (December 1978).
2235 106
PROFESSIONAL QUALIFICATIONS C. VERNON H0DGE TRANSPORTATION BRAdCH DIVISION OF FUEL CYCLE AND MATERTAL SAFFTY My name is C. Vernon Hodge.
I have been employed by the U. S. Nuclear Regulatory Commission since February 1975. My position is in the Transportation Branch of the Division of Fuel Cycle and Material Safety of the Office of Nuclear Material Safety and Safeguards. My responsibilities and activities include the following endeavors:
Review of environmental statements dealing with transportation of radioactive materials, including:
" Transportation of Radioactive Materials by Air and Other Modes," NUREG-0170 (1977),
Generic' Environmental Statement on Use of Mixed 0xide Fuel in Light Uater Reactors (GESMO)," NUREG-0002 (1976),
and
" Transport of Radionuclides in Urban Environs: Working Draft Assessment, " SAND 77-1927 (1978);
Writing environmental statements dealing with transportation of radioactive materials, including:
" Calculations of Radiological Consequences from Sabotage of Shipping Casks for Spent Fuel and High Level Waste,"
NUREG-0194 (1977),
" Environmental Survey of Waste Management in the Light Water Reactor Fuel Cycle," NUREG-Oll6 (1976);
Testifying before Atomic Safety and Licensing Board panels on contentions dealing with radioactive materials in transportation to and from nuclear power stations; Serving on study panels dealing with emergency response to radiological accidents, including:
2235 107 O
Technical coordinator for an NRC/D0T Study Group report
" Review and Assessment of Packaging Requirements (Yellow-cake) and Emergency Response to Transportation Accidents,"
(1978).
This study was initiated at the request of Congressman Timothy Wirth of Colorado af ter a serious spill of yellowcake in transit occurred in Colorado in September 1977.
Several recommendations were developed for Federal rulemaking and Federal-State interaction to enhance emergency response capabilities of various parties.
Representative for the Office of Nuclear Materials Safety and Safeguards on the NRC/ EPA Task Force on Emergency Planning.
This Task Force was convened by the NRC Executive Director for Operations in late 1976 to write guidance for States in the matter of defining the light water reactor
~
accident for which States should plan emergency responses.
The Task Force issued a draft report NUREG-0396 in 1978, which is now out for public comment.
Advisory Committee to the International Atomic Energy Agency to consider emergency responses to transportation accidents involving radioactive materials.
I was able to contribute considerable data from recent NRC studies and activities to the draf t document. A Safety Series document may be produced from this effort in about one year.
Source selection panels for NRC contracts designed to survey the capabilities of State and local government agencies to respond to transportation accidents involv'Ing radioactive materials and to develop criteria for evaluating emergency plans for such accidents.
Serving on groups studyingother aspects of transportation of radioactive materials, including:
Revision of the NRC/ DOT Meniarandum of Understanding; Transportation Working Group of the Interagency Review Group on Management of Nuclear Wastes, convened by the President; and Research review group to monitor a recently initiated NRC research program to develop quantitative information on sabotage of spent fuel casks.
2235 108
Prior to employment with the NRC, I was employed by an engineering consulting firm, Holmes and Narver, Inc., Anaheim, California.
I was principal investigator for the study " Transportation Accident Risks in the Nuclear Power Industry, 1975-2020," EPA-520/3-75-023 (1975), and reported it to the Fourth International Symposium on Packaging and Transportation of Radioactive Materials, Miami Seach, Florida, in 1974.
I also was principal investigator for a ccmpanion study on routine exposure from transportation. Other assignments involved a conceptual design of a fuel handling facility in a twin HTGR complex, an environmental statement on cleanup, rehabilitation, and resettlement of the Eniwetok Atoll, and an environmental report for a proposed power plant at Kawaihae, Hawaii.
Prior to my employment with Holmes and Narver, my graduate work involved a post doctoral term in theoretical metallurgy at the School of Engineering and Applied Science, University of California at Los Angeles, and Ph.D. (1972) and M.S. (1969) degrees in physics from the University of Idaho.
Publications during this period include:
"On the Elastic Interaction of Tetragonal Precipitates in Cubic Metals," UCLA-ENG-7392;
" Spin-Orbit Interaction in Model Potential Calculation of Lattic Dynamics of Simple Heavy Metals," Ph.D. thesis (1971);
" Current-Induced Reflectivity Effects in a Semiconductor,"
Phys. Rev. B1, 3347 (1970).
- Prior to Sraduate school, I was employed with Idaho Nuclear, Inc., and Phillips Petroleum Co. at the National Reactor Testing Station, Idaho, as a critical facility supervisor and as a reactor engineer.
My responsibilities with the critical facility included planning, conducting, and reporting reactivity and flus distribution measurements of mockup assemblies of experiment, fuel, and poison simulation loadings for a test reactor.
My responsibilities as a reactor engineer included operating a high pcwer test reactor and associated engineered experiments.
2235 109
p;.
PROFESSICNAL QUALIFICATIONS WILLIAM H. LAKE, JR.
- I-L My name is William H. Lake, Jr.
I have been employed by the U. S.
Nuclear Regulatory Commission (U. S. Atomic Energy Commission) since November 1972.
I am in the Transportation Branch in the Division of Fuel Cycle and Material Safety wnich is in the Office of Nuclear Material Safety and Safeguards.
The Transportation Branch is respon-sible for review and approval for use of shipping packages for fissile material and quantities of other radioactive mat'ri-ls' exceeding Type A quantity limits, in accordance with the requirements of 10 CFR Part 71.
One of my responsibilities is to review the heat transfer and thermal analyses of Safety Analysis Reports provided by applicants in support of approval requests under 10 CFR Part 71.
In addition to my primary technical functions as a heat transfer specialist, my responsibilities also include:
- 1) coordination of the technical avaluations of the various disciplines involved in issuance of a certificate of compliance and preparation of a staff position;
- 2) review of containment performance of packages from the standpoint of fluid dynamics, liquid and particle release; 3) evaluation of operating procedures proposed for the handling of packages (i.e.,
loading, unloading, etc.); 4) evaluation of specific test procedures determined to be significant to safety; and 5) assessment of the operational reliability of systems and components that comprise the containment system.
I had been employed by Grumman Aerospace Corporation from June 1969 until joining the Atomic Energy Commission (USNRC).
There, I was employed as a thermodynamicist in the Thermodynamics Section.
I was assigned to space programs and was involved primarily in developing technical proposals.
My experience involved developing thermal models of space vehicles in space, atmospheric re-entry environments, and thermal aspects of ground servicing.
I developed the computer programs for aerodynamic heating for the Grumman Aerospace Advanced Development Program which are described in the following reports:
Lake, William, Aerodynamic Heating to Swept Cylinders Using the Energy Integral Method of Fleming and Krauss," Grumman Aerospace Corporation, Advanced Developmen: Report No. ADR 01-04-71.2, December 1971 Lake, William, " Aerodynamic Heating Methods for Blunt-Nosed Hypersonic Vehicles," Grumman Aerospace Corporation, Advanced Development Report No. ADR 22-02-73.1, February 1973.
2235 110
My publications in the area of transportation of radioactive materials, which are available in the open literature, are:
Lake, William H.,
" Capabilities and Limitations of Heat Pipes for Use in Radioactive Materials Shipping Casks As An Alternative to Active Cooling Systems," Proc. 4th Int.
Symp. Packaging and Transportation of Radioactive Materials, Miami Beach, FL, Sept. 1974 Lake, William H., " Reliability In Maintenance and Design of
~
Elastomer Sealed Closures," Proc. Sth Int. Symp. Packaging and Transportation of Radioactive Materials, Las Vegas, NV, May 1978.
I earned both my B. S. M. E. and M. S. M. E. from the Polytechnic Institute of Brooklyn (Polytechnic Institute of N. Y.) in 1967 and 1970, respectively.
My graduate school specialization was thermal engineering which included:
heat transfer, thermodynamics, fluid dynamics, and combustion.
My masters thesis was titled, " Simulated Spark Ignition Engine," and involved the study of variations in flame growth of a methane air system, and the resulting engine cylinder peak pressure variations which contribute to performance inefficienty.
In addition, I worked on an experimental investigation of flame propagation in solid rocket fuels which resulted in the following technical note:
Blair, D. W. and Lake, W. H., " Solid-Propellant Flame Zone Radia-tion," A1AAJ, V7, No. 9, Sept. 1969.
I am a member of Sigma Xi, The Scientific Research Society of North America.
I joined, by invitation, as an undergraduate in 1967.
I am also a member of Pi Tau Sigma, which is the mechanical engin-eering honor society.
2235 111
e sa,
Resley D. Glenn, Serior Development Enoineer, Environmental Evaluations
~~
Section, Occupational and Environmental. Safety Department, Battelle Pacific Northwest Laboratory Education B.S.
Enaineering Technology, O!1ahoma State University 1971 M.S.
Environmental Health 5'ciences, University of Michigan 1972 Experience Mr. Glenn specializes in Health Physics and ha~s exoErience in the following areas:
Radioloaical incineerino.
Mr. Glenn has professional experience in radiological engineering which includes evaluation of potential hazards, safeguards, soecial requirements for new and unique R&D programs and training of health physics technicians.
Exposure Evaluation.
fir. Glenn has additional experience as exposure evaluation for the Hanford Project.
Duties as exoosure evaluation included evaluation of bioassay, invivo and TL dosimeter data in routine and non routine situations to determine occupational radiation exposure received in these situations.
Environmental Imoact.
Mr. Glenn assist'ed in the generation of the Generic Environmental Impact Statement on Management of Commerically Generated Radioactive Wastes.
Duties included overseeing the evaluation of doses to various population groups from routine and postulated accidental releases. Additional duties included evaluation of environmental impact associated with trasportation of spent fuel, fuel residues and other radioactive wastes.
Publications RD Glenn, KR Heid, JR Houston; Assessment of a Cerium-Praseodymium - 144 Inhalation Case; accepted for publication in Health Physics Journal.
RD Glenn, PE Bramson; Hanford Critical Radiation Dosimeter; PNL-2276.
LG Faust, RD Glenn, et al; A_ Guide to Good Practices at a Plutonium
~ Facility; BNWL-2086.
KR Heid, RD Glenn; Internal Dosimetry Proaram in a Plutonium Facility, The Plutonium Fuel Cycle, Bal Harbour, Florida, May 1977.
2235 112 W
'7 UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION BEFORE THE ATOMIC SAFETY AND LICENSING BOARD In the Matter of
)
)
DUKE POWER COMPANY
)
Docket No. 70-2623
)
(Amendment to Materials License
)
SNM-1773 for Oconee Nuclear Station
)
Spent Fuel Transportation and Storage )
at McGuire Nuclear Station)
)
AFFIDAVIT OF C. VERNON H0DGE AND R. DANIEL GLENN Introduction Our names are C. Vernon Hodge and R. Daniel Glenn.
Copies of our professional qutlifications are attached.
This affidavit addresses a contention which reads as follows:1/
Transportation of spent nuclear fuel from the Oconee Nuclear Station for storage at the McGuire Nuclear Station will create an unacceptable hazard by significantly increasing the radiation doses to persons in the region near the proposed transpor-tation routes between the two facilities.
Specifically:
(a) There will be an unacceptable incremental burden of radiation dose to persons living in the vicinity of the transportation routes.
(b) There will be an unacceptable incremental burden of radiation cose to persons traveling over the transportation routes concurrently with spent fuel shipment.
- / This contention is raised by both Carolina Environmental Study Group and Carolina Action as Contention 2 of " Stipulations" dated October 18, 1978. Only Parts (a) and (b) are addressed in this testimony.
Part (c) is addressed elsewhere by the NRC Staff.
2235 113
. Discussion These contentions claim that the exposure to the population living near the routes of the proposed shipments and traveling on those routes is unacceptable. That the shipments would produce an incremental radiation dose is true, but the expected magnitude of the dose is very small and is acceptable.
Estimates of routine exposure to various groups due to shipment motion and stops, including inadvertent stops such as traffic jams, are given in Section 5.3 of " Environmental Impact Appraisal Related to Spent Fuel Storage of Oconee Spent Fuel at McGuire Nuclear Station -
Unit 1 Spent Fuel Pool" (EIA). Additional estimates are available from recent calculations. All the estimates are summarized in Table 1 below.
That Table also presents data to clarify the significance of these dose estimates. A comparison of the dose related to the spent fuel shipments (assumed to be receised in one year) to annual dose received from background radiation is presented and estimates of health effects of the dose are discussed.
The rauiation dose to persons living in the vicinity of the transpor-tation route was calculated in accordance with data presented in
" Environmental Survey of Transportation of Radioactive Materials to and 1
from Nuclear Power Plants," WASH 1238. Additional analysis based on "The Transportation of Radioactive Material by Air and Other Modes,"
NUREG 01702, corroborated the analysis based on WASH 1238. This 2235 114
. additional analysis indicates that the average dose to a person along the route would be about 0.000003 rem per year with a maximun dose to any individual of about 0.00002 rem per year under normal conditions, or about 0.02 percent of the dose received annually from background radiation.
Transportation doses to persons traveling on the route concurrently with the spent fuel shipment were calculated based on NUREG 0170, Appendix D.
For travel in the direction opposite to that of the sh-ipments, the cumulative population dose for the assumed 300 shipments in one year was calculated to be about 0.04 person-rem.
The average dose to an individ-ual per shipment would be 0.00000009 rem.
The dose to a hypothetical individual who passes each of the 300 shipments would be about 0.00003 rem.
This dose represents about 0.03 percent of the background dose received by such an individual during one year.
The cumulative dose to persons traveling in the same direction and at the same speed as the shipment was calculated to be about 0.08 person-rem.
The average individual dose was calculated to be about 0.00001 rem per shipment and about 0.003 rem if the same individual traveled concurrently with all 300 shipments.
This dose represents about three percent of the background dose received by an individual in one year.
The dose to the public from a delay in transit (in this case a traffic jam in a high density population area) would result in a population dose of about 0.01 person-rem per hour of delay plus about 0.005 ren/ person /
2235 115
. hour for persons parked in vehicles along side the cask during the delay.
Assuming two persons per vehicle and four cars beside the cask at an average distance of 3 meters from the truck for three hours, the popu-lation dose would be about 0.2 person-rem and the maximum individual dose would be about 0.015 rem.
As shown in Table I, the dose estimates (assumed '.o be received in one year) from the described hypothetical exposures are significantly smaller than the dose commitment from background radiation in one year. The basis used for these comparisons is 0.1 rem per year for each individual.
The background dose has been estimated higher, but this number is used for conservatism, that is, to obtain the highest value of the ratio of estimated annual dose commitment due to the proposed spent fuel shipments to the annual dose from background radiation.
The variability of life style contributes greatly to the variation in background dose.
For example, living in a concrete house rather than a wooden one, taking an occasional commercial air trip, eating different foods, and using different consumer products could add doses to individuals ranging from 0.001 to 0.01 rem per year.
Medical and dental radiation exposure would 3
be expected to add about 0.01 to 0.1 rem per year -
Further explanation of the contributing factors in estimation of background dose are given in Table II.
2235 116
d Table I.
Estimates of Individual and Population Doses from Routine Exposure to Proposed 6'9 Spent Fuel Shipments f a
Individual fraction of Si tua tion Individual Group Dose Background 00se Population Group (Assumptions)
Oose (rem)bf jperson-rem)bf (percent)
Health Effen
- 1. Public at truck
- 1. A person standing
- 1. 0.0013
- 1. 1. 3 stops Im (3 ft) from cask for 3 min; not The nun.bers of health normally expected effects are too small to estimate. For e5 ample,
- 2. Public living near 2a. 300 shipments passing 2a. 0.00002 2a. 0.02 the BEIR ReportE/
the Oconee-McGuire must exposed individ-estimates 50-165 latent route ual 30m (100 ft) cancer fatalities per une million person-rem 2b. 300 shipments passing 2b. 0.1 2b. 0.0024 during the ti,rst 25-27 42,000 people living yearsatterigadiation.
near soute In NUREG-0170, estimates of 120 latent cancer 4
- 3. People traveling on 3a. Traffic jam,10 3a. <0.2 3a. 0.02 assuming fatalities per million a
route with truck people /mi2 for negligible dose person-rem and 170 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> outside 1 m12 genetic effects per e
million person-rem 3b. tar following truck 3b. 0.00016 3b. 0.16 were used. Thus the at 30m (100 ft) per occupant annual background for 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> population dose estimated for the N
3c. Same way traffic 3c. 0.08 3c. 0.01 population exposed in g
12 cars exposed max, Item 2b, which is 4200 2 occupants / car, person-rem, would result 300 shipments in less than one latent W
cancer fatality per 3d. Opposi te traf fic.
3d. 0.04 3d. 0.0000003 year and less than one 4200 cars exposed genetic ef fect per year.
max. 2 occupants /
q car. 300 shipments
'/ ssuming 300 shipments in one year.
A b/ or comparison with dose received from annual background radiation, the estimated dose is assumed to be received in one year.
F The individual dose from annual background radiation is taken to be 0.1 rem.
S/See Reference 4.
9/ tee Reference 2.
. Table II.
Composition of Background Dose Estimates for Individuals Dose to Whole Body Source (rem)
United States average from naturally occurring sources, per year
- Cosmic 0.045 Terrestrial
.060 Internal
.025 Total 0.130 North Carolina average from naturally occurring sources, per year Total 0.145
- Source:
Klement, A. W., Jr., Miller, C. P., Minx, R. P., and Shleien, B.,
Estimates of Ionizing Radiation Doses in the United States:
1960-2000 ORP/CSD 72-1.
U. S. Environmental Protection Agency, Office of Radiation Programs, Washington, D. C.
(August 1972).
See also Reference 3.
2235 118
. Of most value for understanding the impact of radiation on human beings is the cumulative population dose estimate. On a well documented 4
statistical basis, a million person-rem of population dose is estimated to result in about 50-165 latent cancer fatalities.
The population dose estimates in Table I are orders of magnitude smaller than this estimate.
Even if the maximum annual doses estimated for the hypothetical situations described in Table I were to be realized, they would not produce readily discernible health effects.
Conclusion We conclude that the routine exposure from these proposed shipments would not be unacceptably large.
Since the proposed shipments will not cause a significant increase in radiation dose to persons near them, they will not pose an unacceptable hazard.
We hereby certify that the above statements are true and correct to the best of our knowledge and belief.
A8 G ' 7 l a s e ri t C. Vernon Hodge P. Daniei Glenn Subscribed and svorn to before me this fo Td day of MAY
, 1979.
2235 119 9pdaa. d.
'x Notary Public My Commission Expires:
_ _ _/jff/,
. REFERENCES 1.
" Environmental Survey of Transportation of Radioactive Materials to and from Nuclear Power Plants," WASH-1238, U. S. Atomic Energy Commission, Directorate of Regulatory Standards (December 1972).
2.
" Final Environmental Statement on the Transportation of Radioactive Material by Air and Other Modes," NUREG-0170, U. S. Nuclear Regulatory Commission, Office of Standards Development (December 1977).
3.
" Radiological Quality of the Environment," EPA 520/1-77-009, U. S.
Environmental Protection Agency, Office of Radiation Programs (September 1977).
4.
"The Effects on Populations of Exposure to Low Levels of Ionizing Radiation (BEIR)," National Academy of Sciences - National Research Council (November 1972), p.91 2235 120
l.ible !.
1stiuutes of Individual and population (kises from Routine (sposure to proposed
', pen t f ue l
',h i pnen t s?'/
Individual Iroction of 5ituation 1ndividuai 6'""P Ih'5" IId'E9'""nd Dose populatnon Group (Assumptions)
Ousejeem)b/
(person-rem)I/
__(pert _ent_) _
ocalth {ftecL
- 1. l'uhlit at tru k
- 1. A person,t.inding 1. 0. 0'11 3
- 1. l.3 stops im (3 f t ) f rom task for 3 min; not t he numiser s o f hea l t h normally espet.ted effects are too snull to estimate, f or example, publii livioy near 2a. 300 shigmients pass ing Za. 0. o(H H i!
24, 0.02 the lit IR Repor tf/
t he Ot once-Mi Gu i e e most erposeit i rid i v i d-es t istia tes f,0-16S Ia tt*nt route ual ton (100 ft) tariter to ta li t ies per one miilion person-rem 2b. 300 shipnents passing Ph. 0.1 2b. 0.0024 during the first 25-2/
42,000 people l i v irej years after ir adiation.
nea r rou t e in fiURf G-Jllu', estimates o f 120 la tent t.ariter 1.
People traveling on la. Ira f t :( 1am, 10
3a.
0.2 3a. 0.02 assusnin.)
fatalities per siiillion route with truth people /mi', for ne91i tble dose per son-r em and 1/0 9
3 hour3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br />, outside I mi?
9enetic efterts per mii! ion person-rem Jb. tar following tautk 3b. 0.00016, 3b. 0.16 were used.
lhus the a t 30m (100 f t )
per oc c upan t antiua l bac k.ji'ound ior 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> lopulation diest' estinu ted f or the N
h '.nne wa y t ra t f i t it. 0.08 30, 0.01 populat ion e-s post d ist N
12 iaes eepo'.ed niax, Ita ni 2b, whic h is 4200 2 o(tupants/s.ar, person-r em, would resul t 300 shipments in less than one latent C
cancer fatality per 1.t. Opposite tratti(,
3d. 0.04 3d. u. 000000 3 year and less than one 4/00 tars esposed 9ene t ic e f f ec t per yea r.
g nu x, 2 ott.upants/
< a r, 300 sh ipmen t s
" Assuming (00 ',hipments in one year.
b/ or.ou.pasison with dose rec eived f r om annual backgr ound radiation, the estimated dose is assumed to lie r et elved in one year.
l Ihr indi vi.f ua l dose f r om annua l bad yround radi.it ion is talen to be 0.1 retit.
- / ',et - !?eferenee 4.
d/ 9g
,fg g g, pg, g f,
PROFESSIONAL QUALIFICATIONS C. VERNON HODGE TRANSPORTATION BRANCH DIVISION OF FUEL CYCLE AND MATERIAL SAFETY My name is C. Vernon Hodge.
I have been employed by the U. S. Nuclear Regulatory Commission since February 1975.
My position is in the Transportation Branch of the Division of Fuel Cycle and Material Safety of the Office of Nuclear Material Safety and Safeguards. My responsibilities and activities include the following endeavors:
Review of environmental statements dealing with transportation of radioactive materials, including:
" Transportation of Radioactive Materials by Air and Other Modes," NUREG-0170 (1977),
Generic Environmental Statement on Use of Mixed Oxide Fuel in Light Water Rcactors (GESMO)," NUREG-0002 (1976),
and
" Transport of Radionuclides in Urban Environs: Working Draft Assessment, " SAND 77-1927 (1978);
Writing environmental statements dealing with transportation of radioactive materials, including:
" Calculations of Radiological Consequences from Sabotage of Shipping Casks for Spent Fuel and High Level Waste,"
NUREG-0194 (1977),
" Environmental Survey of Waste Management in the Light Water Reactor Fuel Cycle," NUREG-Oll6 (1976);
Testifying before Atomic Safety and Licensing Board panels on contentions dealing with radioactive materials in transportation to and frem nuclear po er stations; Serving on study panels dealing with emergency response to radiological accidents, including:
2235 122
_2_
Technical coordinator for an NRC/00T Study Group report
" Review and Assessment of Packaging Requirements (Yellow-cake) and Emergency Response to Transportation Accidents,"
(1978).
This study was initiated at the request of Congressman Timothy Wirth of Colorado af ter a serious spill of yellowcake in transit occurred in Colorado in September 1977.
Several reccomendations were developed for Federal rulemaking and Federal-State interaction to enhance emergency response capabilities of various parties.
Representative for the Office of Nuclear Materials Safety and Safeguards on the NRC/ EPA Task Force on Emergency Planning. This Task Force was convened by the NRC Executive Director for Operations in late 1976 to write guidance for States in the matter of defining the light water reactor accident for which States should plan emergency responses.
The Task Force issued a draf t report NUREG-0396 in 1978, which is now out for public comment.
Advisory Committee to the International Atomic Energy Agency to consider emergency responses to transportation accidents involving radioactive materials.
I was able to contribute considerable data from recent NRC studies and activities to the draf t document.
A Safety Series document may be produced from this effort in about one year.
Source selection panels for NRC contracts designed to survey the capabilities of State and Tocal government agencies to respond to transportation accidents involving radioactive materials and to develop criteria for evaluating emergency plans for such accidents.
Serving on groups studyingother aspects of transportation of radioactive materials, including:
Revision of the NRC/ DOT Memorandum of Understanding; Transportation Working Group of the Interagency Review Group on Management of Nuclear Wastes, convened by the President; and Research review grcup to monitor a recently initiated NRC research progran to develop quantitative information on sabotage of spent fuel casks.
2235 123
Prior to employment with the NRC, I was employed by an engineering consul ting firm, Holmes and Narver, Inc., Anaheim, Cali fornia.
I was principal investigator for the study " Transportation Accident Risks in the Nuclear Power Industry, 1975-2020," EPA-520/3-75-023 (1975), and reported it to the Fourth International Symposium on Packaging and Transportation of Radioactive Materials, Miami Beach, Florida, in 1974.
I also was principal investigator for a companion study on routine exposure from transportation.
Other assignments involved a conceptual design of a fuel handling facility in a twin HTGR complex, an environmental statement on cleanup, rehabilitation, and resettlement of the Eniwetok Atoll, and an environmental report for a proposed power plant at Kawaihae, Hawaii.
Prior to my employment with Holmes and Narver, my graduate work involved a post doctoral term in theoretical metallurgy at the School of Engineering and Applied Science, University of California at Los Angeles, and Ph.D. (1972) and M.S. (1969) degrees in physics from the University of Idaho.
Publications during this period include:
"On the Elastic Interaction of Tetragonal Precipitates in Cubic Metals," UCLA-ENG-7392;
" Spin-Orbit Interaction in Model Potential Calculation of Lattic Dynamics of Simple Heavy Metals," Ph.D. thesis (1971);
" Current-Induced Reflectivity Effects in a Semiconductor,"
Phys. Rev. 81, 3347 (1970).
- Prior to graduate school, I was employed with Idaho Nuclear, Inc., and Phillips Petroleum Co. at tha National Reactor Testing Station, Idaho, as a critical facility supervisor and as a reactor engineer.
My responsibilities with the critical facility included planning, conducting, and reporting reactivity and flus distribution measurements of mockup assemblies of experiment, fuel, and poison sir;ulation loadings for a test reactor.
My respcasibilities as a reactor engineer included operating a high power test reactor and associated engineered experiments.
I 2;235 124
Resley D. Glenn, Senior Development Encineer, Environmental Evaluations Section, Occupational and Environmental. Safety Department, Battelle Pacific Northwest Laboratory Education B.S.
Enaineering Technology, Oklahoma State University 1971
~
M.S.
Environmental Health Sciences, University of Michigan 1972 T
Exoerience Mr. Glenn specializes in Health Physics and ha's expbrience in the following areas:
Radioloaical Encineerina.
Mr. Glenn has professional experience in radiological enaineering which includes evaluation of potential hazards, safeguards, special requirements for new and unique R&D programs and training of health physics technicians.
Exposure Evaluation.
Mr. Glenn has additional experience as exposure evaluation for the Hanford Project.
Duties as exposure evaluation included evdluation of bioassay, invivo and TL dosimeter data in routine and non routine situations to determine occupational radiation exposure received in these situations.
Environmental Incact.
Mr. Glenn assisted in the generation of the Generic Environmental Impact Statement on Management of Commerically Generated Radioactive Wastes.
Duties included overseeing the evaluation of doses to various population groups from routine and postulated accidental releases. Additional duties included evaluation of environmental impact associated with trasportation of spent fuel, fuel residues and other radioactive wastes.
Publications., _
RD Glenn, KR Heid, JR Houston; Assessment of a Cerium-Praseodymium - 144 Inhalation Case; accepted for publication in Health Physics Journal.
RD Glenn, PE Bramson; Hanford Critical Radiation Dosimeter; PNL-2276.
LG Faust, RD Glenn, et al; A Guide to Good Practices at a Plutonium Facility; BNWL-2086.
KR Heid, RD Glenn; Internal Dosimetry Procram in a Plutonium Facility, The Plutonium Fuel Cycle, Eal Harbour, Florica, May 1977.
2235 125
.