ML20247L532
| ML20247L532 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee  |
| Issue date: | 05/23/1989 |
| From: | Buchheit J, Grube R, Little P, Marian M, Pizzuti R, Reid D VERMONT YANKEE NUCLEAR POWER CORP. |
| To: | |
| Shared Package | |
| ML20247L519 | List: |
| References | |
| OLA, NUDOCS 8906020195 | |
| Download: ML20247L532 (70) | |
Text
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UNITED STATES OF AMERICA NUCLEAR REGULATORY COMMISSION before the !
ATOMIC SAFETY AND LICENSING BOARD
)
In the Matter of )
)
VERMONTYANKEE NUCLEAR ) Docket No. 50-271 OLA POWER CORPORATION )
) (Spent Fuel Pool (Vermont Yankee Nuclear ) Expansion)
Power Station) )
)
Sworn Written Testimony of Donald A. Reid, Michael J. Marian, Rudolph M. Grube, John M. Buchheit, Richard P. Pizzuti, acd Peter. S. Littlefield, Submitted by Vermont Yankee Nuclear Power Corporation Pursuant to 10 C.F.R. 6 2.1113(a)
Pursuant to 10 C.F.R. 6 2.1113(a) and th s Board's Memorandum and Order of January 12,1989, the Licensee, Vermont Yankee Nuclear Power Corporation, submits the within sworn written testimony of Donald A.
1 Reid, Michael J. Marian, Rudolph M. Grube, John M. Buchheit, Richard P. Pizzuti, and Peter S. Littlefield in support of its position that Environmental Contention 3 in this proceeding is without merit.
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8906020195 890523 PDR ADOCK 05000271 0 PDR
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i I I. Identification of the Witnesses I Mr. Reid is employed by Vermont Yankee Nuclear Power Corporation and is Operations Support Manager. He holds a Bachelor of Science Degree in Mechanical l
Engineering and is licensed as a Professional Engineer in the State of Vermont. Mr.
Reid has 18 years experience in the nuclear industry including 15 years at Vermont Yankee, where his po' ,ns have included Engineering Support Supervisor, Technical Services Superinte sent, and Operations Superintendent.
I Mr. Marian is employed by Yankee Atomic Electric Company and is the Assistant Project Manager for the Vermont Yankee Project. His responsibilities include the direction of Project departments that support plant licensing, design, and operation in
-I the areas of fuel management, environmental engineering, reload, transient and safety analyses, and quality assurance. Mr. Marian holds a Master of Engineering degree in Nuclear Engineering and a Bachelor of Science degree in Nuclear Science and Engineering. He has 15 years experience in the nuclear power industry and 5 years at Yankee Atomic Electric. Prior to joining Yankee Atomic Electric, Mr. Marian was Supervisor of Radiological Engineering at EDS Nuclear (currently Impell Corporation) where he directed a staff of engineers involved with radwaste, health physics and analytical services. Previously, he was Lead Radiation Protection Engineer at Stone &
Webster Engineering Corporation where he was responsible for shielding design and accident analyses for the design and construction of the Millstone Unit 3 Nuclear Power Station.
1 I Mr. Grube is employed by Yankee Atomic Electric Company and is Director of the Fuel Management Department. His responsibilities include directing a staff of approximately 10 engineers and professionals who provide fuel cycle management r - d I
1 I procurement services to Vermont Yankee and three other nuclear power plants in New F
England. Mr. Grube holds a Bachelor of Science degree in Chemical Engineering. He has worked in the nuclear power field for 30 years including 22 years at Yankee Atomic and 8 years with the Atomic Energy Commission.
l Mr. Buchheit is employed by Yankee Atomic Electric Company and is a Senior Nuclear Fuel Engineer in the Fuel Management Department. His responsibilities include the administration and coordination of activities related to the storage and disposal of spent fuel and the procurement of nuclear fuel for Vermont Yankee and three other nuclear power plants in New England. Mr. Buchheit holds a Bachelor of !
Science degree in Biochemistry. He has 23 years experience in the nuclear power field including 17 years at Yankee Atomic Electric Company and 6 years with the Atomic Energy Comission at the former Nuclear Fuel Services Reprecersing Plant at West Valley, New York.
Mr. Pizzuti is employed by Yankee Atomic Electric Company as its Construction Manager. His responsibilities include direction of a staff that is involved with all I construction related activity for the company including cost estimating. He has 22 years experience in the nuclear industry which includes various responsibilities in the construction of several new nuclear plants and backfitting work in operating plants. He ,
holds a Bachelor of Science degree in Civil Engineering and is a Registered Professional Engineer in the Commonwealth of Massachusetts and the State of New Hampshire. l I Mr. Littlefield is employed by Yankee Atomic Electric Company and is the Manager of the Radiological Engineering Group. His responsibilities include directing a staff of hnical personnel in performing the analyses of the radiological impacts of nuclear I ,
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I power station operation. Mr. Littlefield holds a Master of Science degme in Radiation
! Biology (Health Physics) and a Bachelor of Science degree in Chemical Engineering.
He has 24 years experience in health physics including 20 years with Yankee Atomic.
He is certified in health physics by the American Boani of Health Physics.
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II. Purpose of the Testimony The purpose of this testimony is to establish (i) the decision-making process by which Vermont Yankee determined to expand spent fuel storage capacity by the curmnt reracking rather than by storage in dry casks, (ii) the schedular parameters that rendered the dry cask alternative inadequate to meet Vermont Yankee's need either at the time the decision was made or today, (iii) the comparative incremental environmental parameters that demonstrate that dry casks have no environmental advantage over reracking and that, from today's perspective, an election to switch to dry casks would result in I increased environmental costs, and (iv) that dry casks am more costly than reracking.
I III. Vermont Yankee Experience / Decision Process I
During 1977, Vermont Yankee increased the storage capacity of its spent fuel pool by reracking with spent fuel racks having a closer center-to-center spacing. In doing so Vermont Yankee gained more than a three-fold increase in storage capacity. The initial reracking was done because of Vermont Yankee's concern that reprocessing was no longer available as an option nor would off-site storage or disposal be available for spent fuel within the time frame required to suppen continued operation. Development I of the conceptual design and completion of licensing of the original reracking were accomplished in less than two years.
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f I l In 1982, Vermont Yankee commenced reviewing again its spent fuel storage situation. Projections ofits spent fuel generation rate and storage needs were compared
! .E E to its installed spent fuel storage capacity. Storage needs were simply determined because reprocessing had all but disappeared as an option and because there were indications that a permanent disposal facility would not be available until the late 1990s.
It was therefore concluded that all spent fuel generated would remain on-site for the foreseeable future. This was reinforced by the enactment of the Nuclear Waste Policy Act in 1982, wherein the Department of Energy was not mquired to accept spent fuel until January 1998. Based on this assessment, Vermont Yankee concluded that spent fuel storage capacity should be maximized so that full core discharge capability would be maintained and continued operation would not be jeopardized.
(Although Vermont Yankee typically discharges approximately 1/3 of the reactor core at each refueling, a reactor can experience a situation in which a teminrary off I -loading of the full core (368 fuel assemblies at Vermont Yankee) is required to perform some inspection or maintenance inside the reactor vessel. Consequently, it is considered crucial from an operational perspective (albeit not fmm a safety perspective) to maintain sufficient empty spaces in the spent fuel pool to accomodate such a full core off-load. If this full core off-load capacity is lost, then a problem can result in a loss of the ability to repair and/or operate the reactor.)
I Based on the then current operating schedule, Vermont Yankee was expected to fill its spent fuel pool to the point where sufficient spaces to accomodate the off-load of a full reactor core could be lost by 1989. Further,if the rerack alternative were selected, consideration of the specific logistics required to support implementation of the new I
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I l racks indicated that the sooner the project was started, the better. This was intended to I minimize the number of fuel mover necessery and/or to avoid using timely and costly temporary storage capacity before an old style rack could be emptied and then replaced.
i Also, completion of the purchase of additional old style racks (to increase the spent fuel pool capacity to 2000 bandles) could be avoided. )
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i Vermont Yankee therefore began a review of possible methods available for increasing on-site storage capacity. Of the various methods reviewed, only pool
,l storage /reracking had been previously licensed in the United States. By 1984, pool ;
storage ofirradiated uranium elements and nuclear fuel assemblies had a proven track record of use in the United States for more than thirty years. Additionally, the industry had realized improvements over earlier spent fuel rack designs. These improvements made it possible to store more spent fuel assemblies within the same spent fuel pool volume. These factors, when combined with Vermont Yankee's program to maximize the volume of the spent fuel pool allocated to fuel storage by minimizing the volume allocated to non-fuel storage activities, led to Vermont Yankee's determination that it could add significant additional storage capacity to the spent fuel pool with minimum delay and uncertainty and less cost through another reracking. The decision to pursue the rerack alternative was made, after considering all the information available to Vermont Yankee, at a meeting in November,1984, and a letter of intent to purchase the new racks was given to the vendor in February,1985. By April,1986, sufficient infom1ation was developed such that an application proposing to incicase the capacity of the existing spent fuel pool by 870 fuel bundles was submitted to the NRC for review and approval.
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I IV. Factors Affecting Schedular Availability I A. Vermont Yankee Experience / Considerations I information available to Vermont Yankee both at the time the decision was made to increase storage capacity by reracking (1984) and at the time the license amendment was submitted (1986) showed: (1) the activities req" ired to implement a spent fuel pool expansion through reracking were well known and predictable; (2) there existed a solid base of industry and Vermont Yankee experience with fuel racks and reracking; (3) there were no known technical or licensing issues or uncertainties related to reracking; and, (4) the costs of reracking were known with a high degree of certainty while the costs of altemative storage methodologies were much less certain. In particular, the activities to implement the dry cask storage scenario were largely unknown and therefore less predictable (and still are for a specific application at Vermont Yankee).
Additionally, at the time, reracking was the only available fuel storage option that had been previously licensed, and it was the only option that involved no new technology.
Also, as a well understood procedure, it qualified for "no significant hazards" treatment. (More than 100 such applications have been received by the NRC and not one has been denied.)(4) From the perspective of a management required to make decisions (and responsible for the decisions it makes), certainty has a significant value.
The decision would have been the same even if (as was not the case) the costs of reracking appeared to be appmximately equal to the costs of dry casks. Based on these considerations, Vermont Yankee made its decision to expand spent fuel pool storage capacity by reracking.
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l The foregoing conclusions are even more compelling from today's perspective. Of the aggregate projected costs of reracking, most of the engineering costs, most of the procurement costs, all of the costs of disposing of the old racks, and most of the costs l of installing the new racks have already been incurred. In addition, however long licensing will ultimately take (an uncertainty with both options), 5 years of the process j have already been accomplished for reracking. Consequently, a decision to change from reracking to dry casks, if made today, would involve substantial economic and ,
schedular penalties,if not also a shutdown of the reactor.
B. Industry Experience I
Udlity experience to-date for licensing, construction, and initial loading for dry cask storage facilities is as follows:
I The Virginia Power Company's Surry Independent Spent Fuel Storage Installation (ISFSI) is a facility designed for dry storage of spent nuclear fuel in air cooled, metal casks. The casks are loaded in the spent fuel pool and then transferred to an outdoor concrete storage pad where they are stored in a vertical position.
Vircinia Power Company - Surry ISFSI Pre-license application discussions September 23,1981(1)
, License application submittal: October 8,1982(1)
Licenseissued: July 2,1986(1)
Initial fuelloaded: June 24,1987(1)
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I Carolina Power and Light's H. B. Robinson Steam Electric Plant uses a horizontal modular concrete design rather than dry casks for spent fuel storage. Spent fuel is I. sealed in multi-element stainless steel canisters which are then transferred into thick walled, reinforced concrete modules for storage. l Carolina Power and Light Comoany - H. B. Robinson ISFSI Pre-license application discussions July 20,1984(1)
License application submittal: February 4,1985(1) i License issued: August 13, 1986(1) hiitial fuelloaded: March 16,1989(2) 1 When Carolina Power and Light Company's license application for its H. B.
1 Robinson Steam Electric Plant ISFSI was received by the NRC, reviews of the !
Virginia Power Application and the Surry ISFSI cask vendor's topical report were well underway. During the Surry ISFSI licensing review, the NRC had established acceptable review procedures, the means to document design analyses, and the technical issues and their bases. These factors, plus the use of relatively standard materials for the H. B. Robinson ISFSI, allowed for a more straightforward review process and a quicker approval.(6)
Storage in spent fuel pools is included in the license issued for a nuclear power plant by the NRC under 10 CFR 50, " Domestic Licensing of Production and Utilization Facilities." Dry storage in an independent spent fuel storage installation at reactor sites or elsewhere is subject to the regulations contained in 10 CFR 72,
" Licensing Requirements for the Independent Storage of Spent Nuclear Fuel and High l I
I 1 Level Radioactive Waste." Under 10 CFR 72 in its present form (and as it existed when Vermont Yankee made its decision), a new, separate license is required for the independent spent fuel storage installation. A major document to be submitted as a part ofie application for that license is the Safety Analysis Report (SAR), whose purpose is to demonstrate the safety of the dry storage facility design under the conditions i specific to the reactor site in question. In practice, the vendors of dry storage systems have attempted to enhance the marketability of their designs by submitting topical reports on their designs to the NRC in advance of utility submittals of site specific license applications. Although the NRC's review of a Topical Safety Analysis Report (TSAR) is not a licensing action, TSARS reviewed and approved by the NRC may be referenced in the SAR for a site specific licensing application, thereby effering the possibility for reducing the time and expense involved in the licensing process.(3) On the other hand, because the TSAR approval by the NRC Staff is not a licensing action, allissues affecting the TSAR and its approval would appear to be open for litigation in a contested ISFSI license proceeding.
Both Surry and Robinson are pressurized water reactors (PWRs). The NRC Staff has not completed any Safety Evaluations for TSARS for the dry storage of boiling water reactor (BWR) fuel like Vermont Yankee's. Only one such "JSAR (received November 1985) for BWR spent fuel storage is currently under NRC Staff review.0)
C. Estimated Schedule to Implement Dry Storage at Vermont Yankee Design and construction time for a dry cask storage facility is necessarily site dependent and will vary according to cask vendor. Although Vermont Yankee does not -
have a detailed schedule or estimate of the time required to implement dry cask storage l
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I at the Vernon site, we believe that the following discussion represents a fair assessment ;
of the timing required for such implementation.
i We estimate that it would take a minimum of one year, more likely two years, to develop a site specific desig i and to prepare a license application for such a dry cask storage facility. Vermont Yanke has no estimate of the time required to license such a proposal, inasmuch as that duration would be controlled by the regulatory pmcess, but we have used 1.5 years to 3.5 years for estimate purposes (assuming no contested hearings) taking into account the Surry and H. B. Robinson experiences and the fact that no BWR cask has yet been reviewed by the NRC Staff. Once licensing had been accomplished, Vermont Yankee estimates it would take at least one year to complete the construction of the facility. In addition, approximately six months to one year would be required to prepare and obtain NRC Staff review and approval, as required, of the required associated changes in the VYNPS Emergency Plan and Procedures, Station Administrative Procedures, Station Health Physics Procedures, and an ISFSI Training ,
i and Certification Program, all of which requirements were imposed upon Virginia Power Company for its Surry ISFSI following the issuance of its license, but prior to authority to render the facility operational.
In summary, it is projected that the range of time required to design, license, construct, and implement a dry storage facility at Vermont Yankee, exclusive of contested hearings and the time necessary for land acquisition, crane modifications, or other modifications to the existi.ng plant, is as follows:
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Minimum Maximum .
l Design and Prepare Licensing Application 1.0 years 2.0 years l
Process Application and Issue License 1.5 years 3.5 years j Procurement and Construction 1.0 years 1.5 years i
Follow-up Activities Prior to Fuel Loading 0.5 years 1.0 years Total 4.0 years to 8.0 years )
i Therefore, at the time Vennont Yankee was required to make its decision, reracking was the most viable solution since dry casks did not appear to offer a timely or predictable alternative. Not only is there not enough time now to switch to dry cask storage, but there wasn't enough time to opt for dry casks when the decision was made when one considers the additional uncertainties (and associated risks) noted above.
V. Incremental Environmental Comparison A. Radiological Considerations The offsite dose from fuel presently stored in the fuel pool is negligible due to the shielding provided by the fuel pool water. Vermont Yankee's Technical Specifications require that the pool water be maintained at a minimum level of 36 feet to ensure adequate shielding. This existing amount of shielding is sufficient to ensure that the l
planned addition of 870 spent fuel assemblies to the fuel pool inventory will have no measurable impact on offsite dose.
t The direct dose from fuel stored in dry casks is expected to result in a small contribution to the offsite dose. Vermont Yankee is committed to the State of Vermont l' not to exceed a direct radiation dose rate of 20 mrem / year at any location on the site I _
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boundary. (This is significantly more restrictive than the limit imposed by 10 CFR 72.) The design of new facilities must ensure that this limit is not exceeded. Thus, any consideration of dry cask storage would necessitate a design with features that would limit the dose rate offsite to within the existing commitment to the State of Vermont.
Based on the reviews of the experience gained with similar projects at other plants, and based on Vermont Yankee's previously reracking experience, exposures to Vermont Yankee personnel for the entire reracking effort, including the installation of the enhanced spent fuel pool cooling subsystem, were estimated at 33 person-rem.
This dose estimate will not affect Vermont Yankee's ability to traintain individual occupational doses within the limits of 10 CFR 20, and as low as is reasonably achievable (ALARA) in accordance with Regulatory Guide 8.8. Incremental occupational exposures due to long term operation with an additional 870 fuel bundles in the spent fuel pool are negligible due to the shielding provided by the fuel pool water as discussed above. Any dose rate contribution to Vermont Yankee personnel due to l spent fuel stored in the fuel pool is negligible in comparison to the normal dose rates measured in the vicinity of the fuel pool surface.
The occupational exposure to construct and install a dry cask storage system would be insignificant. The implementation of dry cask storage however would involve l greater occupational exposures than would the reracking option. This is due to the additional fuel movement required for dry cask storage (initial move into spent fuel racks and subsequent move into dry casks) and the surveillance requirements that would be associated with a cask storage facility. Over the lifetime of Vermont Yankee, total occupational exposure would be approximately equal for the two options.
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With reference to spent fuel pool releases, the principle source of radiation doses I to individual members of the general public is Kr-85. Any additional Kr-85 release due to the addition of 870 spent fuel assemblies to the spent fuel pool will be small and I j
subsequent doses to the population will be environmentally insignificant. Any spent fuel stored in dry casks must first spend a period of time, nominally 5 to 10 years, stored in the spent fuel pool to allow time for decays in waste heat generation and fission products. The majority of any radioactive gaseous releases would occur during this time period regardless of the fuel storage alternative chosen. As a result, both !
reracking and dry cask storage can be expected to have an equal, but insignificant environmental impact insofar as production of gaseous wastes is concerned.
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No increase in the amount of liquid radioactivity released from Vermont Yankee is expected as a result of the reracking modification. The spent fuel pool cooling and demineralized system operates as a closed system and only generates liquid wastes during the backflushing of the demineralized. Since the demineralized is normally changed out based on chemical contamination (not radioactivity), no increase in backflushing is expected with increased fuel storage. Likewise, dry cask storage is not I expected to increase environmental liquid radioactive releases.
As discussed above, the demineralized resin is normally changed out based on chemical contamination, not radioactivity loading. Therefore it is not expected that the frequency of resin replacement will change following the reracking modification. Since j the resin is the major source of solid low level waste resulting from fuel pool operations, neither reracking nor dry cask storage will impact the quantity of solid low level waste generated.
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B. Non-radiological Considerations I Both reracking the existing spent fuel pool and dry cask storage have minimal' associated environmental costs. However, of the two alternatives, dry cask storage has incremental environmental costs greater than reracking. This occurs mainly because the reracking alternative uses an existing storage facility wherem the dry cask alternative needs to have a storage facility built.
The only non-radiological effluent affected by reracking is the waste heat rejected by the Vermont Yankee's cooling systems to the Connecticut River. Spent fuel assemblies freshly unloaded into the spent fuel pool constitute the major heat source to the fuel pool. The rate of heat generation in the assemblies decreases rapidly after removal of the fuel assemblies from the reactor. Any spent fuel stored in dry casks must first spend a period of time, nominally 5 to 10 yeam stored in the spent fuel pool to allow time for decays in waste heat generation and fission products. During this time, the majority of waste heat ever to be generated by the discharged spent fuel would !
be rejected regardless of the fuel storage alternative chosen. As a result, neither
.I reracking nor dry cask storage is expected to have a significant environmental !
advantage insofar as waste heat ejection is concerned.
The potential incremental environmental costs are listed below. All of these costs associated with dry cask storage are similar to or greater than the environmental costs associated with the spent fuel pool reracking.
I 1. Steel mining and manufacturing: Dry metal casks use considerably more steel than reracking. Dry concrete casks use similar or greater amounts of steel than reracking.
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- 2. Concrete manufacture and pad construction: Dry metal and concrete cask storage use considerable amounts of concrete. Reracking uses none.
I 3. Land use: Dry metal and concrete storage casks increase land use.
Reracking maximizes the utilization of existing facilities.
I 4. Air impacts: Both reracking and dry metal and concate storage casks have minimal impact on air, the principal impact being heat rejection to the air.
Both cask alternatives have a marginal environmentalimpact on air due to construction of the casks and storage facility.
- 5. Water impacts: Reracking has negligible impact on water consumption.
Dry metal and concrete storage casks use water resources for the production of concrete. Neither reracking nor dry cask storage will have an environmental advantage insofar as the temperature of the Vermont l
Yankee cooling water discharged to the Connecticut River is concerned as any spent fuel stored in dry casks must first spend a period of time, nominally 5 to 10 years, stored in the spent fuel pool to allow time for decays in waste heat generation and fission products.
I C. Conclusion i
The environmental costs of reracking are negligible. The environmental costs of I. dry metal and concrete storage are small although not as small as reracking. Essentially all environmental costs of reracking have already been incurred while those of any dry I I .
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cask alternative are all incremental. Consequently, a decision to switch fmm reracking to casks, if made from today's perspective, would change dry casks fmm an alternative of no environmental advantage over reracking to an alternative canying with it a greater environmental penalty compared to reracking.
VI. Comparative Economic Costs A. Of the Proposed Reracking I Projected aggregate costs, based on "to-date" actual expenditures, and excluding Vermont Yankee direct costs (Vermont Yankee employees) and hearing costs (legal fees and hearing related additional engineering), are as follows:
Engineering $1,500,000 Materials 2,700,000 I Installation 3.300.000 Sub-Total $7,500,000 Enhanced SFPC Sub-system 3.000.000 I Total $10,500,000 A study recently published by the United States Department of Energy entitled l
" Final Version Dry Cask Storage Study"(3) (hereinafter " Final DOE Report") states the following: Reracking is usually the choice of utilities for increasing storage capacity at reactor sites because the technology for reracking is fully developed and reracking is usually the least expensive of all available options.(5) i I I
I B. Of Dry Casks i Vermont Yankee has not prepared a detailed, site specific study of the cost of a dry cask storage facility. However, the Final DOE Study provides some generic cost estimates to which we have added preliminary estimates of Vermont Yankee site specific costs. Additional costs likely at Vermont Yankee, but not included in these estimates, relate to land acquisition, crane modifications and/or intermediate transfer equipment, all of which are likely to be required but cannot be estimated until a particular dry cask technology wne selected for detailed evaluation.
For a metal, dry cask alternative, it is estimated that 17 storage casks would be required to provide the same ir.cremental increase in storage capacity of 870 bundles as represented by the proposed reracking. These casks are reported to cost about $1 million each. According to the Final DOE Study, a cask transponer to move the large casks to a storage pad or building would cost about $0.3 million. Additionally, using 1
data from the Final DOE Study, Vermont Yankee estimates that a concrete storage pad for 17 casks would cost about $0.9 million. The cost ofloading and handling each j cask would be $20-40 thousand per cask, or about $0.5 million for the 17 casks.
These estimates include engineering costs. The implementation of a dry, metal cask storage pad at Vermont Yankee equivalent in capacity to the present reracking is l estimated to cost:
Casks $17.0 Million 1 Transporter 0.3 Million Concnte Pad & Ancillary Equipment 0.9 Million Cask loading and Handling 0.5 Million
$ 18.7+ Million I !
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If site specific evaluations made during the design process indicated that a shielded structure surrounding the pad is required to meet site boundary dose rates, Vermont Yankee estimates the engineering and construction of such a seismically qualified structure would cost about $6.1 million. The implementation of a dry, metal cask storage building at Vermont Yankee equivalent in capacity to the present reracking is estimated to cost:
I Casks Tmnsporter Shielded Structure
$17.0 Million 0.3 6.1 Million Million Cask Loading and Handling 0.5 Million
$23.9+ Million I
The Final DOE Study also cites limited cost data for other alternative at-reactor I storage technologies. Among these is storage in dry, horizontal concrete modeles. For i
a facility of this type which would be suitable for Vermont Yankee's needs, a cost range of $60 to $80/kg HM (heavy metal)is cited. Assuming that the mid-point of this range applies, and noting that a Vermont Yankee fuel bundle contains about 180 kg uranium (i.e., heavy metal), Vermont Yankee's estimated cost for storage capacity equivalent to 870 fuel bundles is at least ,$11 million. This cost estimate does not include any site specific considerations.
Other data is available in the literature to support these prices. For example, in a recent article by E. R. Johnson of E. R. Johnson, Associates published in the August 1988 issue of Power Magazine, the cost of storage using metal casks is stated as
$40,000 to $50,000 per PWR (pressurized water reactor) assembly. This equates to approximately $100 per kilogram of uranium in spent fuel since a PWR assembly contains approximately 450 kilograms of uranium. Although the storage cost for a BWR assembly is not stated in the report, Johnson notes that metal casks are designed I
i to contain twice as many BWR assemblies. However, because a BWR assembly contains approximately half the uranium fuel of a PWR assembly, the cask storage I cost per kilogram for BWR fuel is approximately the same as for PWR fuel. Each Vermont Yankee fuel bundle contains approximately 180 kilograms of uranium. One !
can therefore calculate a storage cost of approximately $16 million for incremental dry ask storage equivalent to 870 spent fuel bundles. This independcut cost estimate l
compares reasonably well to the estimates based on the Final DOE Report presented above. l I C. Conclusion I The economic costs of reracking (including the costs for design and installation of the enhanced spent fuel pool cooling subsystem) are less than the economic costs of dry metal and concrete cask storage. Further, the costs to implement reracking include t,ite specific considerations and are relatively predictable while those associated with the dry cask alternative are simply projections based on generic studies. Most of the economic costs of reracking have already been incurred while those of any dry cask storage alternative are all incremental. Consequently, a decision to switch from reracking to dry casks, if made from today's perspective, would carry with it a :
significant economic penalty.
I VII. Summary I 1. As of the time Vermont Yankee made its decision, reracking had high probability with high confidence of being available in time to meet the need, while dry casks had doubtful probability, with great uncertainty, of being available to meet the need (and only if changes and accomplishments beyond the power of Vermont Yankee to control LI i o
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occurred). As of today, the time required to implement dry cask storage may be less, but the time available is less and the same conclusions apply.
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- 2. As of the time Vermont Yankee made its decision, reracking had high probability with high confidence of n inimum environmental impacts, while dry casks had high (but lower than reracking) probability with less confidence of small (but larger than rerackiag) environmental impacts. As of today, dry casks have greater envimnmental penalty because more raw matenals would be utilized and property disturbed and because all environmental costs of reracking have already occurred.
- 3. As of the time that Vermont ' Yankee made its decision, reracking had a significant cost advantage with a high degree of certainty. As of today, the cost penalty of dry cask storage has a potential for diminishing, based on uncertain effects of future market acceptance, but is still more costly and more uncertain. In addition, virtually all costs of reracking have already been incurred so that a switch to dry cask storage now effects a significant economic penalty.
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- 4. A preference of dry casks over reracking, whether made then or now, and whether I (if made today) taking into account sunk costs or not, does not effect a significant environmental advantage not outweighed b/ economic penalty.
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Then personally appeared Donald A. gid, who bein first duly swom, made oath that the fomgoing statements en tme, this L of Ma fore me:
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Commonwealth Of Massachusetts:
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Then personally appeared Michael J. Marian, who being first duly sworn, made oath that the foregoing statements are true, this AA%_ f May,1989, before me:
AW Notary Public My Commission expires f/>9/4/
Commonwealth of Massachusetts:
VlDCtMYSL caMTY ss.
Then personally appeared Rudolph M. Grube, who being first duly sworn, made oath that the foregoing statements are true, this 4 V of May,1989, before me: ,
b Notary Public My Commission expires f/79/4/
Commonwealth of Massachusetts:
LAlMc4ST4 CAWiYss.
Then personally appeared John M. Buchhjit, who being first duly sworn, made oath that the foregoing statements are true, this 121. of May,1989, before me:
NM Notary Public My Commission expires //3f/f/
Commonwealth of Massachusetts:
uAb C.GSTFA CbdCN ss.
Then personally appeared Richard P. Pizguti, who being first duly sworn, made oath that the foregoing statements are true, this a of May,1989, befom me:
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Notary Public My Commission expires #29/9/
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l Commonwealth of Massachusetts:
vioWSfGC %)iv ss, Then personally appeared Peter S. Littlefi Id, who being first duly sworn, made oath that the foregoing statements are true, this.:D of May,1989, before me:
Notary Public I My Commission expires M19/9/_
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1
I REFERENCES 1 NRC Staffs Response to NECNP's First Set of Interrogatories and Request for Production of Documents to the NRC Staff on the Staffs Environmental Assessment, December 8,1988.
I 2 Personal communication with Mr.R. Kunita ofCarolina Power and Light.
3 Final Version Dry Cask Storage Study, United States Department of Energy, DOE /RW-0020, February 1989.
4 Environmental Assessment and Finding of No Significant Impact - Spent Fuel Pool Expansion, Vermont Yanke Nuclear Power Station, July 25,1988.
5 Final Version Dry Cask Storage Study, United States Department of Energy, DOE /RW-0020, February 1989, Page 2-19.
6 NRC Staffs Response to NECNP's Third Set ofInterrogatories and Request for Production of Documents h the NRC Staff on the Staffs Environmental Assessment and Finding of No Significant Impact - Spent Fuel Pool Expansion, March 14,1989.
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I I 2ND CASE of Level 1 printed in FULL format.
PAGE 2 i
I In the Matter of GENERAL PUBLIC UTILITIES NUCLEAR CORPORATION, et al.
Three Mile Island Nuclear Statien, Unit 2 Docket No. S0-320-OLA; ASLBP Nc. 87-554-3-OLA (Disposal of Accident-Generated keter) ; LBP-89-07 I WUCLEAR REGULATORY COMMISSION Atomic Safety and Licensing Board 1989 NRC LEXIS 11 February 2, 1989
[*l)
Peter B. Bloch, Chair; Glenn O. Bright; Dr. Oscar H. Paris For General Public Utilities Nuclear Corporation, et al. (Applicants nl): Thomas I L Baxter, Ernest L. Blake, Jr., David R. Lewis, Maurice A. Ross, all of Shaw, Pittman, Potts & Trowbridge.
n1 In prior opinions, referred to as Licensee or GPUN.
I For the Staff of the Nuclear Regulatory Commission (Staff n2): Stephen H.
Lewis, Colleen P. Woodhead.
n2 In prior opinions, referred to as NRC Staff.
For Three Mile Island Alert and Susquehanna Valley Alliance (Interveners n3), I Frances Skolnick. l n3 In prior opinions, referred to as Joint Interveners.
I OPINION:
FINAL INITIAL DECISION '
The issue before us is almost a decade old, n4 It originated during the famous Three Mile Island (Unit 2) accident in 1979. As a result of the I accident, the reactor building basement was covered with about 260,000 gallons of accident-generated water (AGW). Staff's Preliminary Environmental Impact Statement, NUREG-0683, June 1987, Staff Exhibit 1, Supp. 2 at 2.1, Sec. 2.1 and I. Table 2.1. Since the accident, additional water has accumulated.
e.g., id at 2.3, Table 2.2. Water not present at the time of the accident but Ibid; see,
(*2] which has been used for cleanup following the accident is classified as I AGW because it has cecote radioactively contaminated. The final volume of AGW at the end of defueling is expected to be approximately 2.3 million gallons.
Id. at 2.3 (Table 2.2, footnote (c)).
I n4 We used language very similar to the beginning of this opinion in the beginning of our prior opinion, Rulings on Motions for Summary Disposition, LBP-88-23, 28 NRC 178 (1988).
I Several alternative methods for disposing of the AGW have been considered.
See, e.g., id. at v-vii, including Table S.I. After considering the summary I - .
I I
I 1989 NRC LEXIS 11, 2 PAGE 3 I disposition papers before us, we concluded that the principal remaining genuine i.ssues of fact, for which there would be a hearing, was whether the AGW should 2e evaporated (and the solidified evaporator bottoms properly buried), as I
{
3roposed by General Public Utilities, or whether it should be stored in tanks on I site (the "no .ction alternative"), perhaps for 30 years, to allow most of the
- ritium t decay. n5 n5 Applicants and Staff now agree that tritium is the radioisotope of I :ritical concern in assessing the radiological impacts of the proposed evaporation since it will remove radioactive solids, such as strontium 90, from
- he AGW until those solids will be reduced to 1/1000 of their original level; ut the process will not affect the quantity of tritium. See, LBP-88-23, 28 NRC I' _78 at 191. [*3)
We note that Applicant and Staff filings concerning summary disposition I lisagreed about the cost of these two alternatives and that the record did not
- entain detailed information on the cost of on-site storage for 30 years. We
'ere not sure, at the time of our summary disposition decision, why more consideration was not given the no-action alternative, but a possibility we I onsidered was that the Staff : wiled to give it adequate consideration because
.t believed that Commission policy prohibited it. We also noted the following
- enuine issues of fact in the record: (1) the amount of tritium now present in
- ach of the separately stored portions of the AGW, and (2) the seriousness of I :he health effects of the release of tritium through evaporation.
I. Interveners' Procedural Position Thes'e issues we have just discussed, previously set forth in our Summary Disposition decision, were the issues we heard and that we must decid.t. n6 iowever, the Interveners, represented by a non-lawyer, n7 adopted a static theory of their case (see Interveners' Conclusions of Iaw, particularly I ?aragraphs 2, 3, and 3 (the second paragraph numbered 3)), in which they apparently did not accept (*4] our framing of the issues and sub-issues and
- ontinued to argue that they should prevail because the Staff's Preliminary I Environmental Impact Statement, NUREG-0683, June 1987, Staff Exhibit 1 was deficient.
n6 We stated the same principle in our Order, in the following language:
I The primary issue to be heard is whether the no-action alternative is obviously superior to the forced-evaporation proposal because the latter method vill release all of the tritium in the AGW to the atmosphere without any I [further) prior period of natural radioactive decay.
Related sub-issues to be heard are: whether the tritium content of the AGW has been accurately determined; whether tritium is of more critical concern with I respect to our determination on strontium-90; and whether the risk to public health from tritium released by forced evaporation is greater than Applicants and Staff have acknowledged.
Id at 233.
n7 Frances Skolnick worked diligently for Interveners and we appreciate her efforts to inform this Board.
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PAGE 4 1989 NRC LEXIS 11, 4 our ruling, which we set forth as clearly as we could, apparently was not inderstood by the Interveners n8 However, it is based on sound principles of
.aw that recognize (*S) the nature of the hearing process, in which all the j
- arties have an opportunity to introduce evidence that bears on the determinations made in the preliminary environmental impact statement. After
.aving heard that evidence, it is logical that we should weigh it directly in aking our decision, which is itself a public process of weighing environmental ssues. LBP-88-23, supra, at 183-184, citing the following NEPA cases that we 1so stated we would apply to the analogous issues arising under Commission equlations requiring releases to be As Low as Reasonably Achievable (ALARA):
ublic Service Co. of New Hampshire (Seabrook Station, Units 1 and 2), CLI-77-8, NRC 503, 526 (1977), aff'd sub nom. New England Coalition on Nuclear ollution v. NRC, 582 F.2d 87, 95 (1st Cir. 1978), citing Monroe County
- observation Society, Inc. v. Volpe, 472 F.2d 693, 697-98 (2d Cir. 1972).
n8 During the hearing, we advised them further (Tr. 584, 590-592, relating to ow costs were a part of the pending issue, and Or. 608, describing Interveners' roblem and stating that they have to demonstrate the existence of a better lternative; see also Tr. 581-582, where the Board explained the difference etween the burden of going forward and the burden of proof) . However, see ntervenors' statement. Tr. 1680-1681. [*6)
Perhaps some of the Interveners' apparent difficulty in accepting our rulings ind legal interpretations may have arisen out of confusion about the different stages of agency consideration of the AGW issues.
First, there was the preliminary environmental impact statement. Then,
- ntervenors were permitted to intervene by stating contentions together with
'the bases for each contention . . . with reasonable specificity." Texas Jtilities Electric Co. (Comanche Peak Steam Electric Station, Unit 1), ALAB-868, 25 NRC (1987). Memorandum and Order (Memorializing Special Prehearing
- onference; Ruling on Contentions Scheduling)., unpublished, January 5, 1988 at
- 3. By filing contentions, considered under the liberal standards appropriate for this stage of the case, Interveners indicated the issues that concerned them and they gained a ticket of admission to the case, entitling them to access to the discovery process.
At the close of the discovery process, Interveners faced a motion for summary disposition. At that point, they filed affidavits. As we have already discussed, this Board found that they had raised genuine issues of fact that they could take to hearing; and the Board [*7) also applied the law to this
- ase and defined the genuine issues of fact for hearing.
I The evidentiary record now is far more complete than it was at either the contentions admission stage or the summary disposition stage. And it is this Board's obligation, acting for the Commission, to reach a determination based on the entire record of this case.
II. Conclusion After considering all the facts of record, in light of the applicable law, we conclude that the Applicants' request for an amendment to its license should be granted. By a preponderance of the evidence, they have demonstrated that their
} alternative is environmentally acceptable, because of the acceptable level of J occupational exposure and the very low level of atmospheric release, and we
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1 PAGE 5 1989 NRC LEXIS 11, 7
- enclude that the Interveners' no-action alternative is not obviously superior to the evaporation proposcl.
We are convinced that implementation of Applicants' proposal will have I axtremely small radiation exposure consequences, both to workers and the general oublic. Furthermore, using extremely conseristive assumptions, we conclude that
- he maximum savings in radiation dose that could be attributed to the no-action be 36.4 person rem, comprised predominantly of I alternative would [*B) occupational exposure -- with some exposure of the public. Applying the
- ost/ benefit standard of 10 CFR Part 50, Appendix I, Sec. II, D, it would be appropriate to require Applicants to spend at most S 36,400 ($ 1,000 per person I rem) to avoid this dose consequence, but the no-action alternative (including lisposal oi the waste after 30 years) would cost $ 800,000 more. Consequently, applicants correctly conclude that their proporal is entitled to our ap;roval ind that Interveners' no-action alternative is not obviously superior.
Of particular concern to us has baen the proposed radiation exposure of the reneral pgblic, which is comprised of people who will not have consented to this 3xposure.
Applicant and the Staff independently determined the radiological
- consequences to the public from the controlled, atmospheric release of the evaporated AGW by estimating the dose to both the maximally exposed hypothetical Im af f-site person and to the total exposed population. Interveners did not submit any calculations of their own and did r:t seriously challenge the estimates of the magnitude of release made by the otei parties.
The dose to (*9] the maximally exposed hypothetical off-site person is a
- conservative assentsment of the exposura to a member of the public, as required rj Appendix I to 10 C.F.R. Part 50, using Regulatory Guide 1.109 dose I methodology. It io very unlikely that any actual person will receive a dose as great as that of tha maximally exposed hypothetical individual, who is assumed to be a person in the maximum inhalation location who consumes meat, vegetables and milk from each of the other maximum dose pathway locations.
The MIDAS code, whose use by Applicant was accepted by the Staff and not specifically challenged oy Interveners, calculates the estimated doses to the naximally exposed hypothetical off-site person for the duration of the I evaporation process (taking into account, as well, the extent of crocessing/ reprocessing of the AGW). The dose to the bone is estimated to bc 3.4 mrem, while the total body dose is estimated to be 1.3 mrem (1.2 mrem of which is from tritium).
These are not annual doses but rather estimates for the duration of the evaporation process, ar.d they are still well below the annual guideline of 15 I mrem given in Appendix I to 10 C.F.R. Part 50, for exposure from airborne
(*10) releases. Baker, Tr. 638.
In addition to considering the maximally exposed individual, we can consider I
the estimated dose to the total exposed population; that dose is a more representative assessment of the radiological consequences resulting from evaporation of the AGW. Bakar, Tr. 637. MIDAS was again utilized to estimate the dose to the population. In addition to estimating the inhalation and I ingestion doses to the 2.2 million people within a 50-mile radius of TMI-2, the code also estimates the ingestion dose to an additional 13 million people I .-
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i PAGE 6 1339 NRC LEXIS 11, 10 I issumed to be fed agricultural produce exported from within the 50-mile radius.
The total exposure to the population from evaporation of the AGW is estimated to
>e 2.4 person-rem to the bone, and 12 person-rem to the total body.
For simplicity in calculating an average, Applicants applied the total
}
>opulation dose (to 15.2 million) to the 2.2 million people living within 50 tiles. This highly conservative assumption yields an upper bound average exposure to a member of the 50-mile population of 0.001 mrem to the bone and I '.005 mrem to the total body. Since the evaporation process will take more than
- ne year, the annual population (*11]
>aker, Tr. 638.
doses are less than these values.
The Staff independently estimated the off-site doses expected from
.pplicants' evaporation proposal. The Staff predicted that the maximall) 3xposed hypothetical off-site person would receive 0.8 mrem to the bone, and a I :otal body dose of 0.7 mrem. The Staff estimates a dose to the off-site
,opulation of 0.2 person-rem to the bone, and 3 person-rem to the total body.
- unson, Tr. 742, 747. Considering that different models and assumptions were sed (e.g., GPUN's conservative consideration of population ingestion doses I eyond 50 miles), the Board considers the dose estimates of Applicant and the taff to be in general agreement -- both agree that the doses would be very low.
The Board finds that the insignificance of these doses is evident. Compared I :o the 0.01 mrem Applicant-estimated annual bone dose from strontium and the 1.2 trem total tritium dose that Applicant estimates the maximally exposed
_ndividual might receive from the evaporation of the AGW, the average individual
- n the TMI area will receive 300 mrem per year from natural radiation (about 70 I arem from direct radiation from the soil and cosmic rays, 30 mrem from internal natural radioactivity and weapons fallout and 200 mrem whole body aquivalent from radon daughters) each year.
[*12)
The maximum individual organ dose I o the bone therefore is less than 0.003% of the naturally occurring whole body radiation the average member of the population would receive during the 50-year integration period. The whole body dose from tritium is about 0.01% of the Tatural whole body dose. Baker, Tr. 639.
As another illustration, the worst case dose to the maximally exposed individual is on the order of a single day of natural background radiation, and is received over a one to two year period. Munson, Tr. 743.
I III. Expected Radiation Releases Whether the AGW is disposed of through evaporation or through the no-action I siternative, followed by a method such as evaporation, some radiation will be released to the environment. With respect to either alternative, the amount that will be released is the planned release plus the expected release due to an unplanned mishap.
In this portion of our dccision we will estimate and compare both the expected and unplanned releases from the evaporation alternative and from the no-action alternative.
A. Evaporation (*13) Proposal Releases
- 1. Description of Applicants' Proposal I --
I I 1989 NRC LEXIS 11 13 PAGE 7 I On July 31, 1986, GPUN filed with the NRC report on the disposal of the processed AGW, evaluating three dispo W ~ N ns on the basis of relative technical' f easibility, regulatory cunpliance, environmental effects, costs, waste generated, and time required to accomplish. Based on its evaluation, GPUN I asked the NRC to approve a proposal for forced evaporation followed by vaporization and atmospheric release of the product distillate.
I The GPUN proposal also includes the separation and final treatment of the solids removed and collecte d during the evaporation process and the preparation of the resulting waste product for shipment to and burial at a commercial low-level waste facility.
- 2. Implementation GPUN has entered into a contract with Pacific Nuclear Systems, Inc., to I supply the disposal system. In February, 1988, GPUN authorized the vendor to proceed to final design and fabrication of the disposal system for the specific TMI-2 application. A detailed description of the systems and evolutions which Jill accomplish the controlled disposal of the AGW is contained in GPUN's I rechnical Evaluation Report for Processed (*14] Water Disposal System.
Suchanan, Tr. 456-57. That report is in evidence as Applicants' Exhibit No. 1.
Pr. 470.
The processed water disposal program consists of: (a) a dual evaporator system designed to evaporate the processed water at a rate of five gallons per minute; (b) an electric powered vaporizer designed to raise the evaporator I distillate temperature to 240 degrees Fahrenheit and to release the resultant steam to the atmosphere via a flash tank and exhaust stack; (c) a waste concentrator designed to produce the final compact waste form; and (d) a packaging section designed to prepare the resultant waste for shipment I consistent with commercial low-level waste disposal regulations. Buchanan, Tr.
457.
All ACW will be processed through the evaporator prior to release to the I environment via vaporization. The designed flexibility of the disposal system permits the evaporator assembly to be decoupled from the vaporizer assembly.
this configuration, the evaporator operates independently of the vaporizer and In processes the water in a batch cycle method of operation. The distillate from the evaporator is pumped to a separate staging tank, and the feed to the vaporizer is (*15] supplied from an independent staging tank. Conversely, if the vaporizer is coupled to the evaporator during cperations, the water is processed in a continuous flow operation. The distillat; from the evaporator is I fed directly to the vaporizer for atmospheric discharge. Buchanan, Tr. 458.
Average activity levels have been projected for the tocal 2.3 million gallons
~
of AGW assuming further preprocessing of approximately 31 percent of the inventory. See Buchanan, Tr. 465, Columns 1 and 2. This data appears in PEIS Supplement No. 2, Table 2.2, and is identified as " Base Case" Water. These activity levels formed the basis for the Staff's analysis of the environmental I effects of evaporator discharges. The activity releases occurring from evaporator discharges of " Base case" vater result in releases that are a small fraction of the releases permitted by existing regulatory requirements for the operation of a nuclear power plant. Buchanan, Tr. 458-59.
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1989 NRC LEXIS 11, 15
- 3. Operating Limits Since the PEIS analysis assumed processing " Base Case" water with a vaporizer f j
.scharge tc, the atmosphere containing 0.1 percent of the radioactive I .rticulates from the influent, the PEIS values for Base
.ed as the system operating limit. Thus, when operating
.sposal system in the coupled mode (evaporator and vaporizer in continuous the
(*16] Case will be processed water
]
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.eration), the volume of water being processed will be isolated from all I urces of contamination. Its radionuclides content will be verified to be
. thin the Base Case limits so that quarterly average concentrations of all
.ter processed in this mode will be no greater than the concentrations listed Tr. 465, Column 2.
When processing water through the vaporizer in the decoupled mode
.ndependent of the evaporator), the quarterly average vaporizer influent
- ncentrations will be no greater than 0.1 percent of the values in Tr. 465, I alumn 2. These linits equate to an atmospheric release rate for particulate
- dionuclides of 8.23E-5 uCi per second if processing water containing the
.ximum linits at a rate of five gpm. Buchanan, Tr. 459.
Joint Iraervenors' Material Statement of Fact 4 (xiii) under Contention 3 iserted that the NRC's dose calculations are inadequate because the water atering the evaporator in batch cycle will deviate from the concentrations I isted in Table 2.2 of PEIS Supplement No. 2. See (*17]
.R.C. at 199 (1988). For the reasons just discussed, imit will be the same for all methods of operation, and the batch cycle the system LBP-88-23, supra, operating 28 peration of the evaporator will not affect the resulting dose calculations.
I uchanan, Tr. 459.
- 4. Accident Risks Both Applicants and the Staff testified regarding the potential risks ssociated with the evaporation proposal. Applicants noted that the evaporation rocess is estimated to last from 15 to 24 months. During that time, the AGW ill be stored in an approximately 500,000 gallon tank prior to being vaporized.
I .pplicants estimated the probability of an uncontrolled release from the staging
.ank as 0.17% over a 24-month period, with a rerulting dose of 2.50 mrem from
.he liquid pathway and 1.79 mrem from the airborne pathway to the critical
>rgan-the bone. Weaver, Tr. 47S. We consider Applicants' estimates of the I >ercentage risk and resulting dose to be highly conservative.
- 5. Occupational Risks -- Shipment and Burial The transportation of evaporator bottoms to a disposal site involves radiological and non-radiological risks. Radiological of AGW and bottoms, plus dose risks include and handlersThe(*18]
I occupational to members of dose the to driverspopulation.
general general population dose consists of routine dose exposure to bystanders and other vehicular passengers in addition to accident dose due to transportation mishaps. Weaver, Tr. 475.
Applicants estimated conservatively that disposal of evaporator bottoms will require 8 to 12 truck shipments to the burial site. The average activity of each shipment is expected to be less than 0.5 curies total activity. Applicants I assumed the shipments would travel along the least risk route from TMI te Hanford, Washington, which is the proposed burial site.
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,I I 1989 NRC LEXIS 11, 18 PAGE 9 Applicants estimated the incident-free population dose from 12 shipments ould be 10.4 person-rem, and the estimated dose to the driver per shipment ould be 95 mrem. Weaver, Tr. 475-7 6.
Applicants testified that the expected number of traffic accidents and stalities for these shipments would be 0.049 and 0.002, respectively.
- 6. Accidental Risks -- Shipment and Eurial Taking into account the severity and probability of an accident, the
.pulation density along the least risk route, and the resulting release !
I raction of radionuclides produces 0.003 person-rem expected (*19] frcm these hipments. Weaver, Tr. 476.
In addition, the further preprocessing of AGW prior to evaporation will I reduce approximately 40 liners which will require 20 to 40 shipments isposal and represent a disposal volume of 6,200 cubic feet. Applicants for estified that the expected number of traffic accidents and fatalities resulting rom disposal of these liners would be 0.093 and 0.0038, respectively, and the xpected dose to each driver would average approxintely 15 mrem per accident.
pplicants estimated the incident-free dose to the general population from these nipments as 4.8 person-rem, and taking into account the severi y and robability of an accident, the estimated accident dose as 0.56 g rson-rem.
I eaver, Tr. 476.
The Staff assumed that the maximum accident for Applicants' proposal involved
- he rupture of an 11,000 gallon storage tank of AGW. From this accident I scenario, the Staff estimated a dose of 0.015 mrem to the bone and 0.002 mrem to
- he total body for the maximally exposed individual and a population dose of 0.7 serson-rem to the bone and 0.015 person-rem to the total body. Munson, Tr. 748.
In addition, the Staff estimated that Applicants' proposal (*20) would require a total of 68 waste disposal shipments. The Staff estimated that the 68 2hipments would result in 0.6 accidents, 0.5 injuries, and 0.03 fatalities.
Iunson, Tr. 749.
The Interveners did not challenge the Applicants' or the Staff's risk malysis of Applicants' proposal. Both analyses demonstrate that Applicants' I aroposal does not present significant accident risks.
- 7. Occupational Radiation Exposure -- On-site Applicants conservatively estimated the occupational dose attributable to l
evaporation of AGW and the packaging of the evaporator bottoms as 23 person-rem.
This maximum dose was based on 9.6 person-rem from approximately 16,000 person-hours for the evaporation process in a radiation field of 0.6 mrem per 4
l I hour, 8.7 person-rem from approximately 3,500 person-hours for the packaging of the evaporator bottoms in a radiation field of 2.5 mrem per hour, and 2 to 5 person-rem from the pre-processing of water. Tarpinian, Tr. 443-44.
t
- 8. Off-Site Doses to the General Population Applicants and the Staff both presented testimony estimating the radiological 1
I doses to the public from GPUN's evaporation proposal and from the Interveners' alternative. While the Interveners' (*21] witnesses implied some I
" " - - - - - _ . - _ _ _ __m_,
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I 1989 NRC LEXIS 11, 21 PAGE 10 f>
iticisms of these modeling ef forts, which we address below, they did not offer f
[ eir own estimates of off-site doses or indicate in any quantitative wayThe the cent of any perceived error in the estimates of Applicants and Staff. a ard will discuss the modeling issues first, and then consider the dose I :imates.
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a) Dose Modeling I
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The primary environmental dose assessment computer code used by GPUN j vironnental Controls is the Meteorological Information and Dose Assessment (
stem (MIDAS). The MIDAS Code uses atmospheric dispersion calculations based I
f the Pasquill-Gifford method presented in NRC Regulatory Guide 1.111 to derive j a average airborne concentration, deposition rate from a plume, and the ground 9 ncentration of each radionuclides in each sector as a function of time.
The dose due to direct exposure to radioactive material in the plume and posited on the ground is determined by MIDAS directly from these functions, ing published conversion factors such as those in NRC Regulatory Guide 1.109.
MIDAS also accounts for the transfer of radionuclides through I ker, Tr. 628.
e environment using transfer coefficients from Regulatory Guide (*22) 109. It estimates the concentration of radionuclides in each trophic level to rive at estimates of the quantity of each radionuclides ingested or inhaled by mbers of the public.
When the ingestion and inhalation quantities have been calculated, dose anversion factors (DCF's) are applied. The primary sources of these factors I ce Regulatory Guide 1.109 and NUREG-0172, which in turn are based on ICRP International Commission on Radiological Protection) publications,Cooper, ORP Publication 2, 1CRP Publication 10, and ICRP Publication 23.
including Tr.
.g 32-34. The DCF's take into account both the effective half life of radionuclides in the body as well as the quality factor of the radiation from g sch radionuclides. The dose calculated by MIDAS in the manner described above 3 a 50-year dose commitment.
Cooper, Tr. 635.
Although the relevance of assertions in their testimony was poorlyInterven rticulated, n dose modeling. Dr. Huver referred to studies by Koranda and Martin and by irchmann, and his description of the findings of these studies suggested that I Huver, (*23]
ritium might bicaccumulate in plants and animals. Morgan's testimony also suggested that a
.664-65. Both Dr. Huver's and Dr.
reater Ouality Factor for tritium beta radiation should be acknowledged.
Tr.
tuver, tr. 1655, 1658-59, 1665-65A; Morgan at 2-3. These notions were I :envincingly dispelled by the testimony of Applicants' witnenues, Dr. Auxier and Jr. Fabr4kant, and is discussed in the subsequent portion of our opinion dealing Based on their testimony, we accept the Applicants' use of ilth hr.alth effects.
I 1 DCF pf 1.7 for tritium as a conservative value, b) Dose Estint.ces Applicants and the Staff independently determined the radiological consequences to the public from the controlled, atmospheric release of the evaporated AGW by estimating the dose to both the maximally exposed The dose to thehypothetical maximally off-site person and to the total exposed population.
I (over-estimated) exposed hypothetical off-site person is a conservativeassessment of the ex I
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PAGE 11 1989 NRC LEXIS 11, 23 I I to 10 C.F.R. part 50, using Regula*.ory Guide 1.109 dose methodology.
maximally exposed hypothetical individual, who is a concept that does not exist in flesh and blood, is assumed to be a person in (*24)
The the maximum inhalation location who consumes meat, vegetables and milk from each of the other maximum I lose pathway locations. The estimated dose to the total exposed population is a nore representative assessment of the radiological consequences resulting from evaporation of the AGW. Baker, Tr. 637.
The MIDAS code was used by Applicants to calculate the estimated doses to the taximally exposed hypothetical off-site person for the duration of the evaporation process (taking into account, as well, the extent of
.l arocessing/ reprocessing of the AGW). The total dose to the bone is estimated to g >e 0.4 mrem, while the total body dose is estimated to be 1.3 mrem (1.2 mrem of ihich is from tritium). (If the strontium-90 concentration in the AGW were not reduced by evaporation, the strontium would dominate dose calculations. With a decontamination factor of 1,000 achieved by the evaporator, however, tritium is
- he radionuclides that contributes the most to calculated doses -- 1.2 of the 1.3
- rem total body dose to the maximally exposed individual for immediate I evaporation. Baker, Tr. 643.)
These doses, which are not annual doses but rather estimates for the duration
>f the evaporation (*25] process, still are well below the annual guideline
)f 15 mrem given in Appendix I to 10 C.F.R. Part 50, for exposure from airborne releases. Baker, Tr. 638.
MIDAS was again utilized to estimate the dose to the population. In addition o estimating the inhalation and ingestion doses to the 2.2 million people I althin a 50-mile radius of TMI-2, the code also estimates the ingestion dose to an additional 13 million people assumed to be fed agricultural produce exported from within the 50-mile radius.
The total exposure to the population from evaporation of the AGW is estimated to be 2.4 person-rem to the bone, and 12 person-rem to the total body. For simplicity, in calculating an average Applicants applied the total population dose (to 15.2 million) to the 2.2 million people living within 50 miles. This fields a conservative (i.e., upper bound) average exposure to a member of the 50-mile population of 0.001 mrem to the bone and 0.005 mrem to the total body.
since the evaporation process will take more than one year, the annual I population doses are less than these values. Baker, Tr. 638.
Independent Staff Estimate. The Staff independently estimated the off-site
(*26) from Applicants' evaporation proposal. The Staff I
doses expected predicted that the maximally exposed hypothetical off-site person would receive 0.8 mrem to the bone, and a total body dose of 0.7 mrem. The Staff estimates a dose to the off-site population of 0.2 person-rem to the bone, and 3 person-rem to the total body. Munson, Tr. 742, 747. Considering that different models and I assumptions were used (e.g. , GPUN's conservative consideration of population ingestion doses beyond 50 miles), the Board considers the dose estimates of Applicants and the Staff to be in general agreement that the doses would be very low.
In the interest of conservatism, we adopt the higher of the two estimates in each instance. Hence, we consider that the maximally exposed hypothetical l off-site individual vill receive .8 mrem to the bone and 1.3 to the total body. l The total dose to the off-site population is found to be 2.4 person rem to the I
)
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- _ - - _ _ - _ _ - _________-___________-___-_LU
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PAGE 12 1989 NRC LEXIS 11, 26 I Sone and 12 person rem to the total body. We note that the Applicants' estimate af off-site dose is over 20 times larger than the Staff's estimate for exposure
- o the bone and over 4 times larger for total body.
c) Significance of Doses The Board finds that these doses are [*27] insignificant when compared to radiation doses that people receive every day as the result of natural
- henomena . In addition, for reasons we will discuss below, we conclude that the
- ealth consequences of this additional exposure are expected to be negligible or
- on-existent.
Compared to the less-than 0.02 mrem Board-estimated annual bone dase from
- trontium (.8 mrem divided by the number of years life expectancy) and the 1.2 trem total tritium dose that Applicants estimates the maximally exposed I .ndividual might receive from the evaporation of the AGW, the average individual
.n the TMI area will receive 300 mrem per year from natural radiation (about 70
- rem from direct radiation from the soil and cosmic rays, 30 mrem from internal atural radioactivity and weapons fallout and 200 mrem whole body equivalent rom radon daughters) each year. The maximum individual organ dose to the bone
- herefore is less than 0.006% of the naturally occurring whole body radiation
- he av eage member of the population would receive during the 50-year
.ntegration period. The whole body dose from tritium is about 0.01% of the 1atural whole body dose. See Tr. 639 (Baker -- adjusted by the Board [*28] l i
'or its higher estimate of bone dose).
As another illustration, the worst case dose to the maximally exposed I individual is on the order of a single day of natural background radiation, and is received over a one to two year period. Manson, Tr. 743. The additional 3ose to the maximally exposed individual from evaporation is far below the normal environmental dose variability, and the additional dose to the average off-site individual is thousands of times smaller. Baker, Tr. 640.
Another way of considering these same data is that the dose to the hypothetical individual from evaporation of the AGW would be less than ten percent of an additional dose a person would receive from living in a brick ouilding each year, and is comparable to the whole body dose an average individual in the general population receives from watching color television I each year. The dose to the average individual is many hundreds of times less and thus de minimis. Fabrikant, Tr. 1225.
The NCRP does not even calculate population doses when individual doses are I this low because the NCRP considers them insignificant.
B. Description of the "No-Action Alternative" Munson, Tr. 743.
Since the Interveners (*29] resisted specifying the alternative that they considered to be obviously superior to Applicants' proposal (Tr. 561, 581-590), it is necessary for us to review part of the history of this case to determine which alternative Interveners were entitled to have the Board I evaluate. We then will also briefly discuss some other alternatives the Interveners mentioned and that we briefly inquired into pursuant to our broad powers to require the preparation of an adequate record.
I -
I i I i i
PAGE 13 1989 NRC LEXIS 11, 29 As admitted, Contention 2 states as follows:
The EIS fails to comply with the requirements of the National Environmental
~
olicy Act (42 USC 4332, n.29). The NRC failed to conduct conclusive I isk/-benefit analysis of the "No Action Alternative."
LBP-88-23, supra, 28 N.R.C. at 185 (1988). This contention embodies what we ave identified as the principal issue at the hearing -- whether from a I ost/ benefit standpoint, the alternative is obviously superior to Applicants' roposal. As drafted, Contention 2 would appear to refer to alternative 3.5.1, Liquid Storage in Tanks," evaluated by the Staff in PEIS Supplement No. 2, and hich the Staff alsoThis refers to as the "no action alternative." See Staff Ex. 1 I t [*30) 3.32. alternative involves pre-treatment of the AGW to " base ase" levels (as does Applicants' proposal), id. at 3.2 (Table 3.1), existing nd newly constructed storage tanks at TMI, id. at 3.32, and indefinite storage.
d.
At the special prehearing conference, Interveners clarified Contention 2 by xplaining that "the no action alternative supposes that eventually the5,water ill be disposed of." Tr. 65; Memorandum and Order, supra at 20 (Jan. 1988).
- 1. Procedural Setting It should be noted that the NRC hearing process provides great latitude in I~ ;iscovery and it gives the Applicants for the license amendment the burden of 2 roof. However, an obligation of the Interveners is to state their allegations lith sufficient specificity that the Applicants are put on notice of the issue I an which they have the burden.
The notice requirement is a natural outgrowth of fundamental notions of fairness applied to the party with the burden of proof. As the Atomic Safety I md Licensing Appeal Board has observed:
The Applicants are entitled to a fair chance to defend. It is therefore I
antitled to be told at the outset, with clarity and precision, what argumonts 1re being advanced (*31) and what relief is being asked. . . . So is the l
l Soard below. It should not be necessary to speculate about what a pleading is supposed to mean.
Kansas Gas and Electric Co. (Wolf Creek Generating Station, Unit No. 1),
ALAB-279, 1 N.R.C. 559, 576 (1975) (emphasis added; footnote omitted); see also l Texas Utilities Electric Co. (Comanche Peak Steam Electric Station, Unit 1),
l ALAB-868, 25 N.R.C. 912, 930 (1987). Moreover, the Licensing Board is entitled I to adequate notice of an Intervenor's specific contentions to enable it to guard l against obstructionism of its processes. As noted by the Supreme Court in l
upholding the Commission's requirements for a threshold showing of materiality:
. . . (I]t is still incumbent upon Interveners who wish to participate to structure their participation so that it is meaningful, so that it alerts the agency to the Interveners' position and contentions. . . . Indeed, administrative proceedings should not be a game or forum to engage in unjustified obstructionism by making cryptic and obscure reference to matters that "ought to be" considered. . . . Vermont Yankee Nuclear Power Corp. v.
NRDC, 435 U.S. 519, 553-554 (1978) (emphasis (*32) added).
I . - . . . -
i I
I \
I 1989 NRC LEXIS 11, 32 PAGE 14 Similarly, "an Intervenor is not free to change the focus of its admitted
- entention, at will, as the litigation progresses." Public Service Co. of New iampshire,' et al. (Seabrook Station, Units 1 and 2), ALAB-899, 28 N.R.C. 93, 97 I 1.11 (1988).
- 2. Interveners' Discovery Responses During discovery, Applicants asked Interveners the estimated size of storage
- anks encompassed in the no-action alternative. The Interveners responded
'Those most convenient to the Licensee which ccmply with all regulations and five adequate protection to the workers and the public." (SVA/TMIA's Responses I :o Licensee's Interrogatories and Request for Documents, Feb. 15, 1988, at 13,
- interrogatory 2-2(c).)
Applicants asked the Interveners the location on the TMI site where storage I :anks would be placed. The Interveners responded: "That loc ation which is least tazardous to employees and most accessible to radiation moritoring." (Id. at
.4, Interrogatory 2-2(d).)
The Staff asked the Interveners for what length of time they claimed the cater should remain on site. The Interveners responded: "It is expected that
- he water may remain on-site at least until Unit 1 is decommissioned and for as long as Unit 2 remains in Post Defueling Monitored Storage."
I ;*33]
.SVA/TMIA's Response to NRC's Interrogatories, Feb. 22, 1988, at 4,
- nterrogetory 5.)
b The Interveners also were asked what the ultimate disposal method would be I for their alternative. They responded: "That method which entails the least nealth impact on the surrounding population." (SVA/TMIA's Respenses to Licensee's Interrogatories and Request for Documents, Feb. 15, 1988 at 14, Interrogatory 2.2(f).)
- 3. The Board's Conclusions About the Interveners' contention I Based upon the Contention 2 reference to the PEIS "no action alternative,"
Interveners' statement at the special prehearing conf 9rence, and Interveners' responses to discovery requests, Applicants addressed at the summary disposition stage a Contention 2 alternative which consisted of:
(a) pre-treatment of the AGW to " Base Case" levels of Table 2.2 in PEIS Supplement No. 2; (b) storage of the AGW in existing and newly constructed tanks; (c) a storage period of thirty (30) years; (d) ultimate disposal, but by undefined means. (Licensee's Motion for Summary Disposition on Alternatives (Contentions 1, 2, 3 and 8), May 16, 1988.) l l In response [*34] to Applicants' motion for summary disposition on I Contention 2, the Interveners did not contest Applicants' characterization of E J the Interveners' alternative. In fact, the Interveners explicitly accepted the j 30-year storage period and the use of tanks for storage. (SVA/TMIA's Response j l
' i to Licensee's Motion for Summary Disposition on Contentions 1, and 8, June 20, 1988, at 8.)
2, 3, 4, 5d, 6 l l
l 1
I
, I PAGE 15 1989 NRC LEXIS 11, 34 Further, the Inter.enors' witnesses subsequently endorsed in their direct I testimony the 30-year storage period and the use of tanks. Piccioni, Tr. 127; Aorgan at'3-4 (" Third Set of Comments Relative to Treatment and Disposal of 2,100,000 Gal, of Contaminated Water at TMI-2," by Karl Z. Morgan, September 30, I l988. It appears that Dr. Morgan's five pages of direct testimony were not noved into evidence by the Interveners. See Tr. 1525-26. The direct testimony is not in the transcript of the November 15, 1988 evidentiary session. Since
- ho Board and th7 other parties also apparently assumed the testimony was i
I eceived into evidence, we formally admit the document as Interveners' Exhibit lo. 1 and cite it hereafter as " Morgan.")
Although the parties should be on notice at the contention filing [*35) atage of what issues are to be litigated, at the out-set of the evidentiary learing the Board requested that the Interveners explicitly define their 11ternative so that the Applicants and the Staff would be on notice, without I taking their own assumptions, as to what the Interveners considered to be
' obviously superior" to the Applicants' proposal. (At the summary disposition stage of this proceeding, the Board held that NRC precedent requires us to weept the Applicants' proposal unless an alternative is " obviously superior. "
iee LBP-88-23, supra, 28 N.R.C. at 183-84 (1988).)
The Interveners did not add to the previous understanding of the Board and
- he other parties. Interveners defined their alternative as involving retreatment of the AGW followed by monitored storage in proper tankage for an indefinite period of time prior to final disposition. Tr. 106.
On the third day of hearing, the Board once again requested that the I Interveners define their alternative or alternatives. Tr. 577. The Interveners balked at the notion that they had to set a target for the hearing and argued that they did not consider their alternative a " tight pattern" from which they
- ould not [*36) deviate. Tr. 586. However, after a recess to confer and consider the question, the Interveners stated that they wanted the AGW: (1) pretreated to the level of " achievable case" water in Table 2.2 of PEIS Supplement No. 2, (2) stored in existing locations in the plant for an indefinite period of time not less than 30 years, and (3) subject to continuous research. Tr. 581-90.
- 4. Interveners' First Witness When the Interveners' first witness, Dr. Richard Piccioni, appeared for cross-examination, it became evident that Interveners did not have a single alternative in mind. Although there was no motion to strike portions of his I testimony, the Board would have granted such a motion to the extent that a new alternative was being suggested. Furthermore, we now rule that the thrust of Mr. Piccioni's testimony was irrelevant to the admitted contention.
Dr. Piccioni testified that the AGW should be pretreated until the radioactive content of AGW equals the levels listed in the " achievable" column in Table 2.2 of PEIS Supplement No. 2, rather than to " base case" levels.
, l Piccioni, Tr. 159. Further, Dr. Piccioni testified that he envisioned the l g storage of the AGW in as many as 209 (*37] tanks of 11,000-gallon size on the TMI site. Piccioni, Tr. 120, 145.
I The Board rules that Dr. Piccioni's testimony is irrelevant.
not consider it likely that, in so ruling, we are eliminating an alternative However, we do I _- _ _
I I
E PAGE 16 1989 NRC LEXIS 11, 37 I that strongly recommends itself or that could possibly be " obviously superior" , considering radiation releases and costs.
We strongly suspect that a field of 209 such tanks would be very costly to I
- onstruct and maintain, if there is space available. In addition, immediately '
arecessing to base case would reduce the level of radioactivity in the stored later, but at some cost (including use of the evaporator) and in immediate The I
orker exposure, similar to that occurring in Applicants' proposal.
)rincipal gain from this proposal is the reduction in exposure should there be a aishap and spill in a single tank; however, the Board does not rely on the level radiation from a spill in reaching its conclusions in this case.
- 5. Existing Storage Locations During the Interveners' cross-examination of Applicants' witnesses, I :ntervenors again deviated from the previously understood definition of their titernative. This time, the Interveners had two revelations. First,
- he Interveners announced that they were "looking at a variety of alternatives."
[*38]
Tr. 561. Second, among the variety of alternatives, the Interveners were I ]roposing that the AGW be pre-treated and returned to its existing storage
.ocations (presumably including the Reactor Building basement, system piping, sumps and pools) . This was the first time the Interveners revealed their
' pre-treatment and leave it where it is" alternative, which differs from the I :estimony of the Interveners' own witnesses, who addressed the storage of water in tanks. Piccioni, Tr. 327; Morgan at 3-4.
The Board recognized the inherent unfairness to the other parties presented I oy Interveners' moving definition of the Contention 2 alternative. However, our initial impression was that this alternative might have some advantages, so we asked a few questions of our own -- in the interest of an adequate record --
oefore deciding to abandon this line of inquiry, which is irrelevant under a strict interpretation of the admitted contention.
After the testimony of Applicant's witness buchanan on occupational doses,
- osts, interferences with cleanup completion, accident risks, surveillance and I maintenance difficulties, [*3S) and other impediments to this new alternative, the Board was satisfied that the possibility or simply pretreating the AGW and putting it back in its existing locations was not obviously I superior to the Applicants' proposal. Tr. 946. Therefore, the Board declined to expand the proceeding to consider this option further and it discontinued its inquiry. Tr. 947.
C. Radiation Consequences of the No-Action Alternative Applicants and the Staff both attempted to estimate the off-site dose j consequences of Interveners' alternative. Both assumed that the AGW would be I
evaporated after a 30-year storage period for additional radiological decay. j See Baker, Tr. 642; Munson, Tr. 741, 747. During the 30-year storage period, ]
the tritium would decay to roughly 17 to 19% of its current value, and the strontium-90 would decrease by a factor of about two. Baker, Tr. 642-43; I Munson, Tr. 741, 745.
- Because the models used by Applicants and the Staff both utilize current land I use and population distribution data, the prediction of doses 30 years hence requires assumptions. The witnesses both assumed that off-site parameters I
_ l
I I
I 1989 NRC LEXIS 11, 39 PAGE 17 I renain exactly the same as they are in 1988.
2r, 742-43, 746.
Baker, Tr. 642; Munson, In fact, the projected dose reductions may not even occur.
- hanges in land use and population distribution could actually result in dose
(*40) j l
i 3rojections 30 years from now that are higher than current estimates. In other ords, the effects of decay could easily be offset by other factors.
I Using the Board's conservative (high) estimates of doses from the evaporation )
alternative, as presented above on page (20), we find that:
- As a result of the further storage period the whole body dose to the j
- aximally exposed individual would be reduced from 1.3 mrem from all radionuclides (1.2 mrem of which is from tritium) to about 0.3 mrem, a reduction ,
- o about one-fourth.
- The maximally exposed hypothetical off-site person would receive a bone I ;ose of 0.4 mrem over the individual's life (instead of 0.8 mrem), which epresents an average annual dose of less than 0.01 mrem (instead of less than
.02 mrem from evaporation now). Baker, Tr. 6J9 (adjusted to the Board's bone
.ose estimate).
I
- After 30 years the average exposure to the bone to a member of the
- opulation would be one-half of the currently projected 0.002 mrem, and the
' hole body dose would be one-fourth of the currently (*41) projected 0.01 I trem. Baker, Tr.
titness, Baker, that 643 (adjusted by the Board). We also agree with Applicants' I . . . these dose levels are so low that they are within the range of uncertainty of state-of-the-art dose assessment methodology and radiological monitoring. Baker, Tr. 644.
We note that Staff projections are slightly higher, but at the level of dose I ae are considering we do not consider the projections to be substantively different. Staff states that after the 30-year storage period evaporation would result in doses to the maximally exposed hypothetical off-site person of 0.4 arem to the bone and 0.1 mrem to the total body. The Staff's population dose I astimates are 0.09 person-rem to the bone and 0.6 person-rem to the total body.
Tunson, Tr. 742, 747.
I The Board finds tha'; the doses from evaporation now are already so small that any savings achieved from the Interveners' proposed storage period are unimportant.
IV. Health Effects In discussing the radiation consequences of the Applicants' proposal and comparing them to those of Interveners' alternative, we have considered the
, I magnitude of the expected releases and have compared them to naturally occurring radiation. [*42) In this portion of our decision, we will examine evidence concerning the health effects of radiation releases of the expected magnitude.
This evidence was submitted with respect to issues that survived the summary disposition stage of the proceeding. In a subsequent portion of this opinion, we set forth the contentions that were litigated and the issues that arose under each contention.
I -
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I PAGE 18 1989 NRC LEXIS 11, 42 Before we begin the formal consideration of the issues, however, let us asider an. issue raised in the limited appearance session by Ms. Mary Stamos oorn. Ms. Osborn showed a series of color slides, which she used to lustrate what she believes to be the mutating effects of radiation on plant The presentation was very graphic and emotionally powerful and the Board
'I fe.
s considered its substance.
We understand Ms. Osborn's presentation in the context of the evolving
.cipline of radiation ecology. See Vicent Schultz and F. Ward Whicker, Selected Readings in Radiation Ecology, slogical Aspects of the Nuclear Age: 1972 (not in the record but appropriate gonne National Laboratory, TID-25978, In this respect, we l cause it is being used to consider non-record material).
note that:
43)
All living organisms, always, from the time of their origin on earth until W, have been irradiated.
And in the future, everywhere, they will continue to
- I irradiated.
Some of the radiation, from the sun, is necessary fe; some of it is unnecessary; and some is harmful.
for continued It is extremely difficult sort out the various effects Id at and Buchsbaum, 12 (Ralph to decide what is factual
" Species and of Response to Radiation;
'I gnificance. .
dicecology").
We consider Ms. Osborn's views important because of the depth of commitment I .e attaches to them and because other members of the public obviously support ir.
The difficulty we have in interpreting Ms. Osborn's slides and commentary the lack of information about radiation dose levels, both for background and 3
II-related radiation, and the absence of information about other possible I <planations for the observed phenomena that do not involve radiation as the suse.
For example, NUREG-0738, Investigations of Reported Plant and Animal Health I ffects in the Three Mile Island Area (1980), at 26-28, found that citizen observations about changes in wildlife and vegetation were attributable (*44) o such factors as increased human population, changes in farming methods, rolonged severe winters, chemical pesticides in use throughout the eastern nited States, fire blight (a bacterial disease), red-spider or pear-slug See also Robert Ford eeding, leaf spot disease and the trampling of animals.
mith, "The Vegetation Structure of a Puerto Rican Rain Forest Before and I ain Forest: A Study of Irradiation and Ecology at El Verde, Puerto Rico, U.S.
.tomic Energy Commission, 1970 (finding extensive damage to plant life from from the 3xposure to a 10,000 curies source, with plant damage found 40 meters I source, where the dose was 2,000 rads. Damage was manifested at times of leaf
'all and leaf development and in malformed new leaves. There were no mutation effects reported).
The Board is aware that many citizens will reject our reasons for discounting the pictures of plant mutations, just as they also will reject However, our reasons as judges, it is notfor our
, granting the license that has been requested.Our role is to listen, to consider, to decide and to role to be popular.
explain. [*45)
A. Effects of Tritium on Human Beings I -
1 I
I I 1989 NRC LEXIS 11, 45 PAGE 19 I Interveners have argued, pursuant to a surviving portion of their contention d, that tritium is disproportionately damaging to human beings, thus increasing he risk from the evaporation of tritium as part of Applicants' proposal.
ntervenors have not, however, related their arguments concerning tritium to the i I verall risk they believe will be incurred by the public. Hence, it is up to he Board to relate Interveners' argument that tritium is "more" damaging to the isk that will be experienced in this case.
)
I We note that Interveners' arguments about tritium rest largely on the estimony of Dr. Charles W.'Huver, Tr. 1652-1668 (curriculum vitae at 669-1675), who did not appear before us. We find Dr. Huver's testimony to be ogical and well-presented and credible. We also find that Dr. Joseph Fabricant I nd Dr. John A Auxier, Tr. 1132-1417 (curriculum vitae at 1250-1289, 1237-1249) ppeared before us for cross-examination and analyzed in detail the same sources nalyzed by Dr. Huver (Tr. 1138-1236, 1291-1300), answering questions about the I elevant literature in detail and without estimony of Dr. Fabricant [*46) notes. We are convinced that the and Dr. Auxier is also highly credible and e find it more directly relevant to the issue before us -- how to estimate in n appropriate way the population effects of the Applicants' proposal to vaporate tritium.
We note that the Staff's testimony by James A. Martin, Jr. and Dr. Schlomo aniv was also highly impressive, particularly Dr. Yaniv's ability to cite the I iterature in detail and without notes. Tr. 809-902 (curriculum vitae at 24-831). The Staff's testimony also was helpful because of its focus on the
.ssues before us. In particular, we note that Dr. Huver's conclusions about the affect of tritiated thymidine in animal experiments is of very little value here I )ecause almost all of the tritium is in the form of water. Fabrikant, Tr. 1174,
The Board acknowledges its indebtedness to counsel for Applicants for filing zery complete findings that we have used as the basis for our own findings, naking alterations and explanations whenever we felt necessary. Applicantc' l 3roposed findings were particularly good in stating the strong points of g :ntervenors' evidence, permitting us to (*47] accept those findings when appropriate and to f airly evaluate their impact on this case.
- 1. Tritium in the Environment The movement of tritium through the environment has been well studied and is addressed in detail by NCRP (National Council on Radiation Protection and I '4 measurements) Report No. 62. Tritium is an isotope of hydrogen, and its chemical properties and distribution in nature are essenti' illy the same as hydrogen. A tritium atom may combine with hydrogen and oxygen to form tritiated Water (HTO). All water in the environment has some tritium in it. The tritium I in the AGW at TMI-2 is in the form of tritiated water and would be released in that form by evaporation. The dispersal of tritiated water from the evaporation of AGW will follow the same pathways as natural water in the environment.
Auxier, Tr. 1155-56.
When tritiated water is released to the environment, some of it will eventually become part of other molecules, including organic molecules in plants I and animals. Tritium which becomes incorporated into such molecules is referred to as organically bound. There are several ways by which tritium can become I _
l B
l PAGE 20 1989 NRC LEXIS 11, 47
<ay is art of an organic molecule. The simplest and most prevalent (*48]
trough the natural exchange cf hydrogen ions bonded to oxygen, nitrogen, 21 fur, or phosphorous. In living tissues, about 80 percent of organically
>und hydrogen exists as exchangeable hydrogen which readily assumes equilibrium
.th tritium. Auxier, Tr. 1156.
The remaining 20 percent of organically bound hydrogen is non-exchangeable.
an-exchangeable hydrogen is primarily bound to carbon. Tritium can become I incorporated into an organic molecule as non-exchangeable hydrogen by the lotosynthetic conversion by plants of carbon dioxide and HTO to form hexose, id the ingestion of plants and subsequent cellular synthesis can introduce this an-exchangeable tritium into animal and human tissues. As the organic I alecules containing non-exchangeable tritium undergo biological turnover, these slecules and the tritium are degraded and eliminated as metabolic waste.
2xier, Tr. 1156.
- 2. Tritium Uptake in plants and Animals While tritiated water, like regular water, is taken up in planto and animals, le evidence shows that there is no significant concentration of tritium in I ither plants or animals. In transpiring plants with leaves having large arface areas, tritium (*49] levels may exceed environmental levels through preferential transpiration of non-tritiated water from the surface of leaves to I ne atmosphere. This preferential transpiration is attributable to the ifference in mass between H(2)O and HTO (18 vs. 20), which reduces the vapor ressure of tritiated water to 90 to 92 percent of that of normal water. Under the tritium xtreme conditions of low atmospheric humidity (such as in deserts),
B .ontent in plants may be increased by as much as a This factor of three over phenomenon is the pecific activity of the environmentalAuxier, soil water.
Tr. 1157.
significant in temperate climates.
Tnere may also be a discrimination factor under non-equilibrium conditions.
' hen plants or animals are exposed to tritiated water, some of the tritiated ater will become organically bound. Under equilibrium conditions, the
- ercentage of tritium in organically bound pools in the tissuer, of the body will 8 se equal to the percentage of tritium in freely available body water. If the tmount of tritium in the environment is then reduced or eliminated, the tritium
.n freely available body water will be eliminated and the percentage lowered at I i faster rate (*50) than tritium that ic organically bound. Therefore, luring the transition period, there may be a higher concentration of tritium in organically bound pools than in the body water. Eventually, however, the
- ritium in organically bound pools will also be eliminated, and the organically I sound and freely available tritium will equalize. Auxier, Tr. 1157-58.
- 3. Expert opinion About Tritium Uptake Each of the articles cited by Dr. Huver were considered by the NCRP. NCRP Report 62 subsequently concluded:
No apparent enrichment or concentration effect for tritium has been found in lI aquatic or terrestrial food chains. In fact, dilution in larger hydrogen or organic pools is the general rule, as tritium moves to consumer populations.
There is no evidence for a significant concentration process for tritium in either plants or animals.
u PAGE 21 1989 NRC LEXIS 11, 50 Auxier, Tr. 1157-58; See also Fabricant, Tr. 1174 (no significant
- t 101; Martin /Yaniv, Tr. 819, citing NCRP Report No. 89 (Staff Ex. 3: "[I)n
- ensideration of the experimental evidence, tritium in the form of tritiated as low levels of ater should be considered to bei twice as effective (*51)
[
1xposure to gamma rays for genet c damage. ").
- 4. Individual Studies Koranda and Martin (1969), cited by Dr. Huver at Tr. 1664-64A, rely on a tudy that involved the unique characteristics of the kangaroo rats at a desert eapons test site and is of little relevance. Koranda and Martin observed a atio of organically bound tritium to body water tritium of about 1.2, but NCRP
- encluded that the result might simply reflect the discrimination phenomena
- nder non-equilibrium conditions. Auxier, Tr. 1158-59. Since Koranda was one
.f the authors of NCRP Report No. 62 (see Auxier, Tr. 1409), the conclusion in
- hat Report is entitled to special weight with respect to interpreting the oranda and Martin study.
Another Koranda and Martin paper (1973) was a study of a small p39t of
.rrigated farm land in California. Tritiated water was applied directly to the
-oil, and the movement of tritium in the plants during the growing season was
- bserved. The transfer factors observed by Koranda and Martin were smaller than
- hose conservative (high) values used in GPUN's dose modeling (a soil-to-vegetation transfer factor (Biv) of aboutAuxier, 0.8 for Tr.
Koranda and Martin and 1159-62.
1 Biv (*52) of 4.8 for Applicants' model).
Kirchmann, et al. (1971) provided dataThe on Kirchmann the uptakedataof tritium into the does show higher organically bound hydrogen pool in cows.
aptake with tritium labeled forage than with tritiated water, but the uptake from vegetation is actually consistent with the relative masses of the two I components (water and milk solids). In the case of ingestion of tritium as sater, from 3% to 4% of the tritiumFor silk solids, inrgely in milk fat.
activity in the tritiated milk was vegetation found tofrom ingestion, be in10%
the Since the o 16% of the tritium in the milk was found to be in the milk solids.
I actual solids content of milk is on the order of 10%, the result for the vegetation ingestion is about as expected, while the result for tritium ingestion at water is lower than could be expected from the simple ratios of the mass of the milk components. Auxier, Tr. 1163.
- 5. Tritium in Humans If a human ingests tritiated water, some of the tritium can become organically bound. Tritium may enter organic compounds by exchanging with the hydrogen at any of the labile sites in the molecule. In addition, tritium Fabrikant, may Tr.
incorporated into stable molecular configurations.
I be [*53) 1164. Hydrogen bonded to carbon, however, is usually not exchangeable during some enzyme-mediated reactions. The only mechanism bond to carbon is apparently by de novo biosynthesis.
except by which tritium can Thus, tritium will generally not become incorporated into molecules as non-exchangeableFabrikant, hydrogen, I except to a very limited extent during enzyme-mediated reactions.
Tr. 1165.
- 6. Multiple Compartment Model Not Necessary I --
PAGE 22 1989 NRC LEXIS 11, 53 When warranted a multiple compartment model may be used to account for i 1 corporation of the tritium into organically bound pools, and the eliminat on f tritium ~can be determined by the combined elimination rate of the three and
- mpartments (tritiated water, exchangeable organically However, bound the cells tritium, or molecules
- n-exchangeable organically bound tritium).
.th 1cng retention of tritium because of their biological turnover rates also re slow to incorporate tritium; and at any time only a small portion of these alls are in a biological development stage that permits them to react with the ritium in the body. Therefore, the body eliminates most of the tritium Accordingly, for
<54] B* fore these pools can respond to its presence.
ractical purposes, the tritium in organically-bound compartments usually may be aglected. Inclusion of these compartments complicates calculations and results In essence, 1
a minor change to the committed dose equivalent to body tissues.
- ological elimination of all but a small portion from the body of an average in occurs at a rate of 50 percent every 10 days. Auxier/Fabrikant, Tr.
I .65-66; Cooper, Tr. 636-37.
B. Relative Biological Effectiveness (RBE) and Q Because different types of radiation may produce different amounts of damage a exposed tissue, many studies have been undertaken to compare the extent of amage by a particular type of radiation against This the degree of damage comparison from aas is expressed I aference ne radiation Relative Biological (usually Effectiveness nich is an experimentally derived ratio of the dose of the reference radiation a the dose of the investigated type of radiation that produces the same 200 kev(RBE) x-rays).of a particular type of radiation.
iological effect. Note, however, that an RBE for a particular type of I adiation depends not only on the Linear Energy adiation but also on the particular biological system (e.g. , different tissues)
.nd biological endpoint (e.g., cell Auxier, death, biochemical
[*55) Transfer (LET) of the Tr. 1152-S3;alteration, genetic Tr.
Auxier/Fabrikant, I .lteration, etc.) being examined.
.197-98.
Because different types of radiation can produce differing degrees of damage
.n exposed tissues, absorbed dose in rads is not by itself a sufficient measure I or radiation protection purposes.
lifference in the degree of damage caused by In order to account for the overall the different derived in part typesexperimental from of radiation, lose is multiplied by a Quality Factor (Q), The Quality Factor represents the I IBE values, to arrive at a dose equivalent.
aest scientific judgment based upon consideration systems, effects, and experimental RBE values.
of thedoses Absorbed studies' biological in rads may then Auxier, Tr.
ce multiplied by Q to calculate dose equivalents measured in rem.
I 1153.
- 1. Expert Opinion on Q including the Based on extensive consideration of many studies and andthe factors, NCRP (NCRP Report reported RBE's, both the ICRP (ICRP Publication 26)
No. 91) have adopted a Q value of one for tritium (*56] beta 635.
radiation.
GPUN's MIDAS Auxier, Tr. 1153; Auxier, Fabrikant, Tr. 1200; Cooper, Tr. Cooper, Tr. 635-36, 5 Code, however, conservatively uses a Q factor of 1.7.
- 2. Individual Studies I - _ - - - - _ - - _ _ _ _ _ _ . _ _ _ _ _ _ _ _ _ __ _ _ _ _ _ _
PAGE 23 1989 NRC LEXIS 11, 56 Testimony by Interveners' witr.ess Huver referred to several RBE studies. In only one experiment referenced by Dr. Huver (Dobson and Kwan, 1976) was an RBE Huver,
.n excess of two repf ted; and the RBE reported in that study was three.
?r. 1655. In the otters, reported RBE's ranged from one to two. See Huver, Tr.
.658-59, 1665A.
- 3. Importance of Reference Radiation Huver failed to take into account the reference radiation that was used in ertain of the studies -- particularly Dobson and Kwan. The reference radiation or determining the RBE should have a confirmed R3E value of one. X-rays with
.n effective photon energy of 200 kev are the recommended reference radiation.
ompared to x-rays, the RBE for tritium is about one in most experiments.
- owever, some investigators use Co-60 as the reference radiation. The RBE for
- o-60 gamma radiation relative to the standard reference x-rays has been eported to range from 0.85 to 0.29 with decreasing dose rate.
uxier/ Fabrikant , Tr. (*S7) 1197.
The reference radiation used by Dobson and Kwan was Co-60 gamma radiation at ery low dose rates (1 to 3.2 rads / day). NCRP 63 considered Dobson and Kwan 1976), as well as Dobson and Cooper (1974) which is also cited by Dr. Huver,
.nd concluded that "an analysis of the dose-response curve from 3H and from
'amma rays indicated that the increase in RBE of 3H in this experiment can be iscribed to a reduced effectiveness of gamma rays in the low dose region (Bond,
.978)." Fabrikant, Tr. 1294. The same is true of Furchner (1957), which Huver 11sc referenced. Auxier/Fabrikant, Tr. 1198.
- 4. High Dose Studies The few other studies cited by Dr. Huver pertaining to RBE's e.g. , Worman
- 1954), Moskalev et al. (1973), all involved acute effects at high doses 1295-96. and I lave little relevance. See Auxier/Fabrikant, Tr.1198; Fabrikant, Tr.
loreover, none of the RBE's reported in these studies would bring into doubt the 2 value of 1.7 conservatively used in GPUN's MIDAS Code.
- 5. Theoretical Importance of " Stopping Power" Both Dr. Huver and Dr. Morgan advanced the theory that tritium beta rays I should have a greater Q factor because the specific ionization or stopping
[*58) of electrons increases at the end of the electrons' tracks. Huver, Tr.
1665; Morgan at 2-3. However, the converse is also true -- that beta particles power nave less ionizing ability at the beginning of their tracks. Morgan, Tr. 1625.
I Further, Dr. Morgan admitted during cross-examination that one can integrate Q as a function of linear energy transfer or linear energy over the entire range of energies dissipated along the track of a beta particle to arrive at an effective quality factor, Q, and that when this calculation is done Morgan, in the I manner described in ICRU-40 (Staf f Ex. 5) , Q is approximately two.
1625-27. See also Auxier, Tr. 1343-44. We see no merit, therefore, to Dr.
Morgan's suggestion that tritium beta particles should be likened to alpha and Tr.
fast neutron particles with RBE's of 20. See Morgan at 2.
- 6. Studies of Tritiated Thymadine Finally, Dr. Huver's testimony referred to the incorporation of tritium into I DNA molecules and remarked on several studies, most of which dealt with mice I -
= _
PAGE 24 1989 NRC LEXIS 11, 58 See Tr. 1654-57, 1665A-66A. Dr. Morgan's posed to tritiated thymidine.
stimony mentioned in passing that tritium in DNA can transmute to helium.Neit rgan at servations, (*59) 3.
but we infer that they would have us find that there will be some ightened effect and greater genetic risk than with other forms of radiation.
We address this hypothesis at the outset by observing that tritiated ymidine is a DNA precursor selected in experiments specifically because itand e 11 become incorporated into DNA, imals to tritiated thymidine results in a greater Fabrikant, effect than Tr. 1174-75, exposure 1179-80, 1188.to milar levels of tritiated water.
- 7. Conclusion is tritiated water, not tritiated thymidine, and the The AGW, however, stimony of Dr. Fabrikant shows that while some Fabrikant,tritiated water in the Tr. 1175.
ken rther, up into DNA would be extremely small and negligible. scientific studies hav certain organisms when tritium is incorporated into one of three key
.sitions in the DNA. Fabrikant, Tr. 1173-74.
For all these reasons, the Board agrees with the NCRP that it is reasonably to assume, for the purposes of practical hazards
- nservative ansiderations, [*60) that there is no significant transmutation effect for tritium ad that one may estimate hazards solely on the basis of absorbed Similarly, beta we dose.
agree abrikant, Tr. 1174 (citing NCRP Report No. 63 at 101).
ith the ICRP that the formation of tritiated organic compounds in the body ollowing inhalation orId. ingestion of tritiated (citing ICRP water is Publication too 30, small Part I, atto65-67).
have any ffect on total dose.
We find that the testimony of Dr. Huver and Cr. Morgan is insufficient GPUN usedto a Q ustify a rejection of the dose modeling of Applicants and Staff.
actor of 1.7 in its dose modeling; and while the Staff used a Q factor of onein its di n its calculations in the PEIS, B reviously calculated population dose to demonstrate Martin that
/Yaniv, Tr. use of 820. a Q factor we Accordingly, of wo would not affect its conclusions.
- ind that the dose-modeling used by GPUN and the Staff is acceptable and we eject the view urged on us by Interveners.
C. Radiation Health Effects
~*61]
Although radiation doses of 9 rads or more have, to varying degrees,bee (For ampirical evidence linking lower levels of radiation to health effects.Indeed, linear or 1646-1648.)
affects of 9 rads, see Yaniv, Tr. 837; Morgan, Tr.
I quadratic projections from the high level radiation dose curves expected effects of radiation at low doses are so sparse, natural suggest to in comparison that the Auxier, variations Tr. 1362; in radiat I empirical evidence concerning health effects at those levels.Furthermore, Yaniv, Tr. 815, 835-836.
the effects of doses of less than 20 rads by a factor of 2 to 10.
Fabricant, linear interpola 1210.
PAGE 25 1989 NRC LEXIS 11, 61 We repeat that at very low doses, sucn as those calculated for evaporation, adverse health effects have not been observed and the probability of occurrence
- ould be zero. Fabrikant, Tr. 1201, 1203; Auxier, Tr. 1304; Martin /Yaniv, Tr.
315.
Nevertheless, for radiation protection purposes, advisory agencies and I
- ommittees such as The National Academy of Science's Committee on the Biological 2ffects of Ionizing Radiation ("BEIR") extrapolate from observed effects at high doses to arrive at [*62) risk estimates for low dose exposures. Fabrikant, Tr. 1203-04; Auxier, Tr. 1304; Martin /Yaniv, Tr. 815. The only effects that
- ould be expressed statistically at doses as low as those predicted for evaporation are cancers and genetic ill health. Yaniv, Tr. 815, 842. We Iddress each in turn.
- 1. Carcinogenic Risk Based on the risk estimates calculated by the BEIR III Committee and the recently publisned NIH-NCI Report (sce Tr. 1211) , Applicants' witness Fabrikant ipplied a cancer risk estimate of 1 to 2 x 10<-4> per person-rem to the total aopulation doses estimated both by the Staff (3 person-rem) and by GPUN (12
)erson-rem) for evaporation. By this method, Dr. Fabrikant arrived at estimates ranging from 0.0003 to 0.0024 total excess fatal cancers among the 2.2 million 3eople living within 50 miles of TMI-2. Fabrikant, Tr. 1226.
The Staff's estimate in PEIS Supplement No. 2, based on a risk estimate of 1.35 x 10<-4> per person-rem and its calculated 3 person-rem population dose, was 0.0004. Martin /Yaniv, Tr. 816, 820. ,
The Board adopts Dr. Fabrikant's conclusion that the upper limit probability of even one fatal cancer among the 2.2 million people living within 50 miles
[*63) of the plant as the result of the evaporation of AGW would be less than one chance in 400. In addition, we find that the upper limit probability of a fatal cancer for the maximally exposed individual is less than one chance in 5 million using the NRC's calculated dose, and less than one chance in 2.5 million i using GPUN's calculated dose.
Dr. Fabrikant added that while an excess value can be estimated, in fact no excess cancer will result from tritium and the other radionuclides released during the evaporation process. Fabrikant, Tr. 1226. The Staff's witnesses also perceived very little risk of any cancer incidence and would expect no health effects. Martin /Yaniv, Tr. 815, 816, 844-45. The Board also does not expect any health effects.
a) Testimony of Dr. Morgan Testimony by Interveners' witness Morgan asserted that recent studies of the Hiroshima and Nagasaki survivors show cancer risk is greater by a factor of three than previously published values. Morgan at 2. (This issue was accepted as a litigable issue of material f act in the Board's ruling on contention 3 as well as being include in its discussion of issues pertaining to Contention 5d.
LBP-88-23, supra, 28 (*64) N.R.C. at 200, 214 (1988).) In a table on page 2 of his testimony, he indicated that the absolute model risk value from "Recent Japan Studies-1988" was 4 to 8 x 10<-4> and the relative model risk value was I 1.6 to 3.2 x 10<-3>. (The absolute-risk model assumes that the dose-related excess risk per year observed during the period of epidemiologic study
PAGE 26 1989 NRC LEXIS 11, 64
- ontinues throughout the lifetime of the studied population. The relative-risk model assumes that from the end of the period of study through to the end of the lifetime of the studied population this risk increases or decreases as the normal age-specific risk varies. Morgan, Tr. 1563.)
b) Credibility of Opposing Witnesses We reject Dr. Morgan's testimony and accept the testimony of Applicants and
- he Staff. Our conclusion is based in part on our assessment of the credibility af these witnesses. The qualifications of Applicants' witnesses and the quality 3f their oral responses, for example. were superb. Dr. Fabrikant is a radiologist, a member of the NCRP, a member of the ICRP, and the only scientist
- o have served on all five of the BEIR Committees. While servingfor on the BEIR
- II Committee, he was Chairman of the Ad Hoc Committee (*65] Estimating
- he Total Cancer Risk of Low-Dose, Low-LET, Whole Body Radiation. Fabrikant,
?r. 1143-45, 1148.
Dr. Auxier is a Certified Health Physicist, a member of the NCRP, and until recently was a consultant to the Radiation Effects Research Foundation ("RERF"),
thich is the organization reassessing the dosimetry from Nagasaki and Hiroshima.
suxier, Tr. 1139-41. Further, Applicants' and the Staff's witnesses were fully
- enversant with the subject matter of their testimony and the studies and data anderlying it.
c) Lack of Credibility of Dr. Morgan Dr. Morgan is a past-president of the NCRP but presented no statement of qualifications with his testimony. His testimony generally lacks credibility oecause he cited a document that purported to be a British public document but was not; and he continued to present this visibly incredible document, without remorse, even after its lack of authenticity was alleged by the Applicants. He also was unable to show the source of some of the most significant portions of the testimony he presented to us.
I Dr. Morgan testified that he does not hold himself out as an expert in epidemiology or biostatistics, and that he has not been a member (*66]
of the BEIR Committees or an employee of RERF. Tr. 1564-65. Dr. Morgan testified that he does not have access to raw epidemiologic data on radiation of any also associated cancer and has performed no computer analyses of such data to arrive I at risk values. Tr. 1566. And Dr. Morgan was repeatedly unable to explain or support assertions in his written testimony.
We note that Dr. Morgan was an Intervenor witness who appeared without remuneration. His desire to serve the public interest, as he sees it, is admirable. However, he has been careless in the way he prepared this testimony and in having available information with which to substantiate his views.
We find that Dr. Morgan submitted a document to this Licensing Board as Appendix C to his prefiled testimony in order to show that the United Kingdom had reduced occupational radiation exposure limits in light of the new Japanese I dosimetry studies. Dr. Morgan's characterization of the United Kingdom's actions was inaccurate. See Fabrikant, Tr. 1299. The submitted " document" consisted of four pages. The first two pages consisted of a cover page of a I publication of the British National Radiological Protection Board (NRPB)
(*67] numbered GS9 and entitled, " Interim Guidance on the Implications of
PAGE 27 1989 NRC LEXIS 11, 67 Recent Revisions of Risk Estimates and the ICRP 1987 Como Statement," and an abstract. The last two pages were a " Summary of Main Conclusions and Recommendations." '
Applicants moved to strike this Appendix asserting that the last two pages vere not part of NRPB-GS9. Applicants provided a complete copy of NRPB-GS9 (a .
- en page document), and Dr. Fabrikant testified that he believed that the last l
- wo pages of Appendix C to Dr. Morgan's prefiled testimony were part of a i Friends of the Earth petition previously submitted to the ICRP. Fabrikant, Tr.
1299. l At the hearing, Dr. Morgan testified that the document he provided as j
\ppendix C to his profiled testimony had come to him that way through the mail. j 2r. 1527-28, 1531. He testified that he did not know who provided the document i aecause it came in the mail without any identification. Tr. 1528. Upon I further cross-examination, however, Dr. Morgan produced what he had used to
)repare Appendix C. Dr. Morgan in fact used two separate documents: the full
- ext in pamphlet form of NRPB-GS9 and a separate two-sided page in different
- ype entitled " Summary of Main Conclusions [*68) and Recommendations." This
.atter page was paper-clipped to NRPB-GS9, but Dr. Morgan admitted he had supplied the paper clip. Tr. 1535.
Thus, we find that Dr. Morgan's prior testimony that he had simp l y provided Appendix C to his prefiled testimony in the same manner he had received it was false. Further. Dr. Morgan showed no concern for the authenticity and accuracy of the documents he had provided with his testimony. See Tr. 1528-29, 1553-54; see also Tr. 1531, 1552-54.
We conclude that Dr. Morgan was careless about the accuracy of his testimony, d) Dr. Morgan's Lack of Documentation Dr. Morgan's testimony also lacks credibility because of his inability to produce documentation or supporting explanations for his statements on risk
/alues.
Returning to Dr. Morgan's testimony on comparative risk values, the recent Japanese studies to which Dr. Morgan is presumably referring are the work of the RERF (Radiation Effects Research Foundation, the successor to the Atomic Bomb Casualty Commission). Since 1981, the RERF has been reevaluating the dosimetry used to estimate the doses of the Nagasaki and Hiroshima survivors. The new system that is being developed is designated (*69] DS86 (for Dosimetry System of 1986).
Previous dose estimates were based on the T65D (Tentative 1965 Dosimetry) system. Auxier/Fabrikant, Tr. 1213-14; Martin /Yaniv, Tr. 817.
) The first indication of the effect this reevaluation might have on risk estimates was the publication of a paper, RERF TR 9-87, by two members of the RERF, Preston and Pierce, in 1987. Martin /Yaniv, Tr. 817. Dr. Morgan's
(
J assertion that the Japanese studies will increase risk estimates by a factor of three is based on the ICRP's review of this paper at its 1987 annual meeting in como, Italy. See Tr. 1559. Dr. Morgan, however, appears to be misinterpreting the ICRP's statement.
) -
PAGE 28 1989 NRC LEXIS 11, 69 In the Statement by the International Commission on Radiological Protection issued after the 1987 Como meeting), the ICRP observed that "under the new DS86 dosimetry' this increase in risk is reported as being by a factor of 1.4
- ompared with the risks that would have been estimated by the former 'T65D'
.osimetry." Fabrikant, Tr. 1298; see Applicants Ex. 4 at Tr. 1688A. The ICRP urther observed, "This inclusion (of a longer follow-up period of the opulation sample) and other factors cited in the paper raise the risk2." estimate
- 70) for the exposed population by a total factor of the order of Morgan,
- r. 1559; Lic. Ex. 4 at 1688A (emphasis added).
Dr. Morgan interprets this statement to reflect a "further" increase by two, r a total increase of 2.8 (2 x 1.4). Morgan, Tr. 1559. We find this interpretation inconsistent with the plain meaning of the words and we are not urprised, therefore that others, such as the British National Radiological rotection Board, interpret the ICRP's statement as indicative of a possible nerease by a total factor of two. Morgan, Tr. 1562.
To provide a setting in which to concider Dr. Morgan's testimony, we cite Dr.
abrikant's testimony that at the time of this meeting, the ICRP considered the nformation to be too sparse and preliminary to warrant any consideration for an
~i mmediate change in dose limits. Fabrikant, Tr. 1298. Dr. Fabrikant explained hat Preston and Pierce was a preliminary presentation of the effects of the S86 dosimetry on cancer risk estimates in the Japanese atomic bomb survivors,
.nd the discussion of risk estimates in this paper was based on a number of I : road assumptions to permit the authors to discuss some of the possible
_ implications of the (*71) emerging new data. They were not based on the actual individual radiation dose revisions in Hiroshima and Nagasaki, nor on the I 1ew data of cancer mortality now available. Accordingly, Preston and Pierce's saper does not contain precisely quantified risk estimates upon which radiation
)rotection guidance can be based. Fabrikant, Tr. 1296-97, 1358.
Applicants questioned Dr. Morgan to ascertain precisely how he derived the ralues listed on page 2 of his testimony and attributed to "Recent Japan Itudies." Dr. Morgan replied that he had copied the numbers off of some table, I aut he did not have the table with him, and could not recollect the source.
1571-72, 1574, 1576.
Tr.
Applicants also asked Dr. Morgan for identification of the source of the I /alues attributed to "NRC-1981" on page 2 of Dr. Morgan's testimony.
Morgan could not explain where these numbers came from.
Again, It appears, however, that the numbers came from the 1987 Supplement No. 2 to the PEIS, which uses an Dr.
absolute model cancer risk estimate of 135 per million person rem (1.35 x I 10<-4>/ person rem) and states "Use of the relative risk model would produce risk values up to four times greater than used in this st 5.4 (emphasis added); Tr. 1570-71.
(*72) report." Staff Ex. 1 The NRC regards this range of between zero and four as reflecting the limits of uncertainty in the risk values.
I Martin /Yaniv, Tr. 816, citing Staff Ex. 1 at 5.4 ("The Staff regards the use of the ' relative risk' model values as a reasonable upper limit of the range of uncertainty").
Dr. Morgan's inability to explain the derivation of the risk estimates in his testimony renders his testimony of no probative value. See Virginia Electric and Power Co. (North Anna Nuclear Power Station, Units 1 and 2), ALAB-555, 10 I N.R.C. 23, at 26 (expert witness may not state ultimate conclusions and then profess an inability to provide the foundation for them); at 27 (where the I _ _
~
1 PAGE 29 1989 NRC LEXIS 11, 72
- enclusion rests upon an analysis, witness must make available sufficient
- information pertaining to the details of the analysis to permit the correctness of the can'clusion to be evaluated) (1979).
-- We further note that Dr. Morgan's values are inconsistent with Preston and Pierce (1988) (Applicants Ex. 3), which Dr. Morgan admitted was one of the 3ources for his testimony. See Morgan, Tr. 1588-89. Dr. Morgan admitted during
- ;ross-examination that using (*73] UNSCEAR's method of extrapolating to low mm loses and relative model methodology, Preston and Pierce arrived at a lon-leukemia risk estimate of 3.7 to 7.3 x 10-4 per person-rem. Morgan, Tr.
.620, 1622-24. Again, using UNSCEAR's method of extrapolating to low doses,
)reston and Pierce arrived at a leukemia risk estimate of 0.4 to 0.8 x 10<-4>
I 2er person-rem. Morgan, Tr. 1621.
Thus, the total lifetime risk for all cancers calculated by Preston and aierce using the relative model method and UNSCEAR's method of extrapolating to I .ow doses is approximately 4 to E x 10<-4> per person-rem. Tr. 1621. Dr.
- organ lists this range of relative risk as the absolute model risk. We note
.u:t the range of values attributed to "Recent Japan Studies-1988" and listed by I 3r. Morgan under the heading " Relative Model" is exactly four times the range of alues he has listed under the " Absolute Model" heading. We surmise that Dr.
- organ has simply applied the Staff's statement in the PEIS that the relative nodel produces values of'up to four times greater than the absolute model. We I further observe that if Dr. Morgan had listed the 4 to 8 x 10<-4> per person-rem astimate under the " Relative Model" heading, (*74] as it appears he should, and had applied the Staff's statement the other way around, he would have arrived at absolute model risk estimates four times smaller -- or about 1 to 2 x I 10<-4> per person-rem, which is exactly the range of values used by Applicants in its evaluation of doses calculated for the evaporation of AGW.
I During cross-examination, Dr. Morgan expressed preference for the relative
.nodel risk values rather than the absolute (The relative model does not apply to leukemia, and hence leukemia risk is expressed as an absolute average excess risk. Morgan, Tr. 1624.) Dr. Morgan, however, admitted testifying in a previous I proceeding in August 1982 that the absolute risk model was preferable. Morgan, fr. 1568. He attributed his change of position to a discussion he had with Dr.
Stewart, Dr. Radford and others on the day before his testimony in this proceeding. Morgan, Tr. 1567-69. Later, he attempted to explain his sudden I change of position by asserting that in reality he had gradually changed his mind after publication of the Mancuso, Stewart and Kneale study on the Hanford workers, Dr. Modan's paper on Tinea capitis, and the (1981] GAO report on cancer cited in App. A to Dr. Morgan's prefiled testimony).
I risk (previously (*75)
Tr. 1632-33.
The Mancuso, Stewart and Kneale and the Modan publications, however, were available prior to 1980 (Tr. 1639), well before Dr. Morgan's August 1982 I testimony. Further, Dr. Morgan's discussion of these studies relates entirely to his preference for a supra-linear dose response curve and not to the use of the relative versus absolute risk model. De are left with the impression that I Dr. Morgan chose to advocate the relative risk model for the first time in this proceeding. We need not speculate on when or how he came to this position.
We also find Dr. Morgan's subscription to a supra-linear theory to be highly I questionable. The Mancuso, Stewart and Kneale publication on the Hanford workers and the Modan publication on the Tinea Capitis patients are both I . .
PAGE 30 1989 NRC LEXIS 11, 75 studies with significant experimental weaknesses. The data from the Hanford Jorkers is flawed by the fact that the Hanford workers came from the chemical industry during World War II. Fabrikant, Tr. 1388-89. The types of excess
- ancer found in this population are those associated with the chemical industry.
- onversely, no excess leukemia, which is almost a signature (*76) of radiation-associated cancer, was observed. Id.
The Tinea Capitis studies are complicated by somewhat unreliable dosimetry which was, however, more likely to be accurate than the Hiroshima and Nagasaki lata) for the Tinea Capitis (ringworm) patients treated by irradiation of the scalp and -- what we consider more important -- by delivery of dose not only to
- he thyroid of these pacients but tr *he pituitary gland as well. Fabrikant,
?r. 1385-86; Yaniv, Tr. 837-38.
I Dr. Morgan's reliance on a GAO report is equally infirm. Dr. Morgan claims that the GAO found that the best fit of the curve for ankylosing spondylitis was I aroduced by a supra-linear model. Tr. 1634. Dr. Morgan admitted on ross-examination that the GAO had examined several sets of data, not just
.nkylosing spondylitis, and found that each set of data could be fit acceptably I y more than one model. Tr. 1639-40. When asked specifically whether the GAO eport found that the linear and quadratic models also fit the ankylosing spondylitis data well, Dr. Morgan responded "not at low doses." Tr. 1642. When specifically referred to a paragraph of the GAO report and asked the same I {uestion, Dr. Morgan again replied [*77) "No." Tr. 1643. The paragraph
- hen read aloud by the Licensing Board, which considered the meaning to be obvious:
was In summary, the data for ankylosing spondylitis patients are fairly well fitted by models that grow as a square root, linearly, and quadratica11y at low doses. Rather than end the controversy over the hazards of lowlevel x-rays,
-hey suggest why the coloroversy exists. Note that the above analysis does not I address the issue of the total cancer risk from x-rays, only the chance of getting leukemia. Tr. 1643 (emphasis added).
I Dr. Morgan then indicated he had replied no to Applicants' questions because in his mind fitting " fairly well" means "it doesn't fit too good." Tr. 1644. We find Dr. Morgan's responses inconsistent and we note that the paragraph quoted above refers not only to the linear and quadratic models but also to the square I root model which Dr. Morgan claims the GAO was advocating. We find Dr.
Morgan's testimony on this point an example of intellectual carelessness.
e) Opposing Testimony I In contrast to Dr. Morgan's testimony, Dr. Fabrikant addressed the various dose-response curves for low-level, low-LET radiation doses examined by BEIR and other scientific organizations. Fabrikant, Tr. 1204-08. The 1980 BEIR I [*78)
III Committee found that the linear-quadratic relationship was the preferred model for estimating response based on the existing experimental evidence and microdosimetric theory. Fabrikant, Tr. 1206-07. Both the NCRP and ICRP hold I the same view. In contrast, the supra-linear model is not used by any recognized national or international radiological protection organization for risk estimation for low-dose, low-LET radiation exposure.
The Board rejects the supra-linear model.
Fabrikant, Tr. 1205.
I I _ _
PAGE 31 1989 NRC LEXIS 11, 78 The Staff's witness, Dr. Yaniv, and Applicants' witnesses, Dr. Auxier and Dr.
abrikant, further testified that the new DS86 dosimetry will not raise risks to ny great extent. Yaniv, Tr. 817-18; Yaniv, Tr. 870; Auxier/Fabrikant, Tr.
215; Fabrikant, Tr. 1297-98. Th, i witnesses discussed the implications of a ubsequent RERF publication. RERf TR 12-87 authored by Shimizu et ri., which ontains more precisely quantified data. Yaniv, Tr. 870; Fabrikant, Tr.
297-98. Even as Dr. Morgan admitted, this report supports an increase in risk stimates of about 1.4 in terms of shielded kerma, but finds that using stimated (*79] organ-absorbed doses, the risk coefficients derived from the wo dosimetries are very similar. Morgan, Tr. 1624-25; Auxier, Tr. 1353-55, 391. (Kerma dose is the energy imparted by the radiation to air and is not It is the organ absorbed doses that are i g irectly relevant to risk coefficients.
elevant. Yaniv, Tr. 866-67, 880-82.)
3 The more precisely quantified data in RERF TR 12-87 and other current I epidemiological data are currently being evaluated by UNSCEAR and the present EIR V Committee. UNSCEAR has derived a lifetime risk estimate for high-dose, igh-dose-rate radiation, and if it uses its previous method for extrapolating a low doses, it will arrive at a lifetime cancer risk estimate of 2.25 x 10<-4>
I er person-rem using the absolute model and 3.5 x 10<-4> per person-rem using ne relative model. The BEIR V reassessment is due to be published at the end f the year, but it too is concluding that the new Japanese dosimetry increases isk estimates only slightly and much less than a factor of two. Fabrikant, Tr.
I 297-98.
Both the Staff and Applicants added further perspective on the ramifications Dr. Fabrikant testified that even if the cancer (*80]
I
- f the DS86 dosimetry.
risks were revised upward by a factor of two, and there is no evidence for this it this time, it would still not result in any additional ill-healthFabrikant, from evaporation of the AGW, because the doses would still be too low.
The Staf f actually applied an increased risk estimate of 4 x 10<-4>
I fr. 1215.
>er person-rem in its testimony (an increase by a factor of three over that used
_n the PEIS), doubled its calculated population dose to reflect use of a Q
- actor of two (increasing the calculated population dose from 3 to 6 2erson-rem), and still only arrived at a 0.0024 population risk of cancer.
I :artin/Yaniv, Tr. 816. Even with this adjustment, the Staff's cancer risk istimate would be no greater than the upper bound that Dr. Fabrikant calculated Compare 1 sing Applicants' very conservative 12 person-rem population dose.
I Fabrikant, Tr. 1226.
- 2. Conclusion In conc'.usion, the Board finds that Applicants and the Staff have adequately svaluated the radiation-associated cancer risk. We further find on the basis of the record before us that the new DSB6 dosimetry is unlikely to have any significant effect on Applicants' and the Staff's independent evaluations.
D. Genetic (*81] Risk During summary disposition, based on statements by Applicants and the Staff I that appeared inconsistent, we questioned whether the cancer risk was greater than the genetic risk. LBP-88-23, supra, 28 N.R.C. at 210, 216Interveners, (1988). Both Applicants and the Staff presented testimony on genetic risk.
I however, presented no testimony and conducted no cross-examination on this question. Accordingly, we find Applicants' and the Staff's testimony to be I __ _
E PAGE 32 1989 NRC LEXIS 11, 81 l
undisputed.
l The Staff explained that the apparent discrepancy at summary disposition oetween Applicants and the Staff's estimates of genetic detriment was theMunson, result of the Staff's inclusion of occupational dosos in their risk estimat s.
Tr. 753-55; Martin /Yaniv, Tr. 816. The risk of genetic disorders in the offspring of irradiated individuals is 0.3 x 10<-4> genetic disorders per person-rem in the first generation. The risk in all future generations (the I equilibrium value) is about 2 x 10<-4> genetic disorders.per person-rem. Id.
! These estimates are taken from NUREG/CR-4214, Health Effects Model for Nuclear
?ower Plant Accident Consequences (see id.) based on the BEIR III report (see 3taff Ex. 2; testimony (*82) of Dr. Auxier and Dr. Fabrikant at Tr. 1220).
The Staff did not multiply their calculated off-site population doses by
-hese risk estimates in their testimony to arrive at a prediction of effect, but I :his calculation is simple. Three person-rems (the Staff's calculated total sody dose to the off-site population) multiplied by 2 x 10<-4> total genetic lisorders per person-rem equals 0.0006 predicted genetic disorders, or less than 2ne chance in 1,600 of a single genetic disorder occurring in all future Generations as a consequence of the evaporation of the AGW. Applying this risk
.I astimate to Applicants' population dose of 12 person-rem results in a projection of 0.0024 (2 x 10<-4> total disorder per person-rem multiplied by 12 person-rem) genetic disorders, or less than one chance in 400 of a single genetic disorder.
I Dr. Fabrikant took this process one step further and calculatet the possibility of a single genetic disorder using the maximum equilibrium value (1,100 cases per million live births per rem of parental exposure) reported in the BEIR III Report (rather than the geometric mean used in NUREG/CR-4214 as reflected in I Staff Ex. 2). Even then, there is less than one chance (*83) single genetic disorder during all future existence.
in 200 of a Fabrikant, Tr. 1229.
The Staff also discussed NCRP Report 89, which suggested that using a Q I factor of two for tritium might be appropriate when considering genetic damage.
Martin /Yaniv, Tr. 819. This has little bearing on Applicants' dose calculations, since Applicants' MIDAS Code uses a Q factor of 1.7, but the Staff I doubled its previously calculated population dose (which had been calculated using the normally recommended Q factor of one). Using a Q factor of two increases the Staff's calculated population dose to 6 person-rem and the total number of projected genetic disorders to 0.0012 (i.e., less than one chance in I 800 of a single genetic disorder occurring in all future generations as a consequence of the evaporation of the AGW). In sum, no matter how one manipulates these numbers, the doses are simply too low to predict or expect any genetic detriment. Moreover, the reassessment of the Nagasaki and Hiroshima I survivors suggests that the current genetic risk estimates are perhaps four times too high. Fabrikant, Tr. 1218-19.
E. Conclusion on Health Effects Based on the record described above, we find (*84) that, contrary to Interveners' Contention 5d, Applicants and the Staff have properly evaluated the risks of radiation-associated health effects. Applying the generally accepted I risk estimates used by GPUN and the Staff to the population doses calculated for the evaporation proposal leads to the conclusion that, using estimates that are intentionally high and conservative, there is less than one chance in 800 of a I genetic disorder and less than one chance in 400 of one cancer death from implementation of the evaporation proposal.
I - - -
PAGE 33 1989 NRC LEXIS 11, 84 Costs -- Including Board Conclusions About Total Radiation Costs V.
'~
In this section of our opinion, we review the cost estimates provided to us, then we review the radiation estimates provided to us, and we make a preliminary
=
calance of both kinds of estimates for the two alternatives.
A. Dollar Cost Estimates E! The most accurate testimony on dollar costs was provided by Applicants, who
!! made detailed cost estinates that stood up under cross-examination. Interveners 3 resented no estimates on cost and have not challenged Applicants' estimate of I
- ost. The Staff's cost estimate, while adequate and similar to Applicants' astimate, relied on other environmental (*85) impact statements to estimate the cost of new tankage, and not on market prices, and therefore is not as acceptable as Applicants' estimate. Munson, Tr. 759-61.
Based on our acceptance of Applicants' dollar cost findings, we find that further preprocessing of 31% of the water volume by demineralization prior to evaporation is estimated to cost S 2.1 million. This estimate is based on I ictual 1987 processing costs and includes all handling, loading, and processing speration costs, including the cost of the resia and liners, transportation to aurial, and disposal at Kanford, Washington. Based upon vendor price quotes, the evaporation and vaporization of 2.3 million gallons of processed water and I the packaging of the resulting evaporator bottoms is estimated to cost an additional S 1.7 million. The transportation and disposal of the packaged svaporator bottoms will csst an estimated S 293,700. The total cost for the evaporation proposal is estimated to be approximately $ 4.1 million. Buchanan, I Tr. 460; Munson, Tr. 747.
Interveners did not estimate the cost of the no-action alternative. Hence, I
we accept Applicants' estimate and find that the cost of Interveners' alternative will (*86) depend upon the assumptions made for the design criteria used for the storage facilities. The construction of two additional 500,000 gallon tanks to be co-located with and designed to the same standards as I the existing PWSTs is estimated to cost S 1.3 million, excluding piping, monitoring, and pumps. Buchanan, Tr. 462; see also Munson, Tr. 747.
The cost of Interveners' alternative would also include the $ 2.1 million or I more for AGW preprocessing, and the costs of ultimate disposal. Making the reasonable assumption that AGW disposal and preprocessing costs of the proposal and the alternative cancel each other out (involving processing to " base case" levels), one is left with the additional $ 1.3 million cost for the storage I portion of Interveners' alternative (assuming that they do not need to meet the standards of Generic Letter 81-38) n9, whose scope and applicability were not argued before us).
n9 At one point, Interveners suggested the application of Generic Letter 81-38, setting standards for temporary waste storage, to this case. If we adopted that standard, we would find that the cost of the no-action alternative I
would be S 9.1 million, far in excess of the amount we assume for the purposes of our opinion. Buchanan, Tr. 462-63. (*87)
Since the Board finds that Pennsylvania is likely to have a disposal site, I saving the long travel distance to Washington state, we use our judgment to reduce the transportation costs for the no-action alternative by $ 220,000 of I . . . . . . .
L
_- 34 ZZ PAGE 1989 NRC LEXIS 11, 87 the $ 293,700 transportation estimate made by Applicants. Munsen, Tr. 746,
-- 748-49 (assumption of Pennsylvania rencaitory); Buchanan, Tr. 460; Munson, Tr.
747. Hence, we find the net incremental cost of the no-action alternative, primarily for buying tankage, to be about $ 1 million.
We also are faced with Intervenor's argument that there may be technological improvement during the next 30 years, bringing the cost of ultimate disposal --
l through evaporation or other alternatives -- down. Against that, we weigh 3 staff's argument that the cost of disposal of low-level radioactive waste has oeen increasing in recent years, indicating a possible long-term trend. Munson, fr. 748. On balance, the Board's judgment is that it is conservative to assume I a 10 percent saving in the cost of evaporation ($ '.7 million), or $ 170,000 (treated as current dollar savings). Also, we wi2A conservatively assume that there is no long term trend in waste disposal costs because [*88] the data to ahich Ms. Munson refers are too short term to indicate what will happen over 30
/ ears.
Finally, after reducing the $ 1 million net incremental cost we previously
- alculated by S 170,000 assumed technological savings, we find that the I so-action alternative will cost at least S 830,000 more than Applicants' aroposal.
B. Total Radiation Consequences of the Two Proposals After reviewing the testimony and bringing it tocather in one place, we have constructed the following table, summarizing the radiation consequences of I Applicants' proposal:
TOTAL RADIATION CONSEQUENCES OF EVAPORATION PROPOSAL STAGE OF PROCESS POPULATION AFFECTED DOSE IN PERSON REM EVIDENTIARY SOURCE I Evaporation and vaporization General public (total body =TB) 2.4 bone Baker, Tr. 638 12 TB Occupational 23 TB Tarpinian, I Shipping bottoms General public 10.4 TB Tr. 443-44 Weaver, Tr. 475-76 Drivers 1 TB Weaver, I Shipping liners General public 4.8 TB Tr. 475-76 Weaver, Tr. 476 TB Weaver, I Drivers 1 Tr. 476 Transit accidents General public 0.5 TB Weaver, Tr. 476
[ Board conservatism)
I TOTAL ALL 2.4 BONE 52.7 TB
(*89)
To simplify, let us say that the person rem consequences of Applicants' proposal is 55.1 person rem. We assume: (1) that the no-action alternative will have none of these consequences resulting from shipping, and (2) that the thirty I year storage period will reduce the radioactivity of the bottoms by one-half, thus halving the consequences from evaporation and vaporization at the end of I
- PAGE 36 1989 NRC LEXIS 11, 92 100 percent, the standards of 10 CFR, Part 50, Appendix I, Sec. II, D, would be
-- met.
We conclude that, under the operable legal standards, the Applicants'
'I proposal is itself obviously superior.
could not be obviously superior.
Necessarily, the no-action proposal VI. Uncertainties There are a variety of miscellaneous issues we have yet to address. These include disagreements about the tritium content of the AGW and about the accuracy with which Applicants measure the radioactive content of the AGW.
In writing about these subjects, our attention was called to the question of the relevance of these subjects to the litigation before us. To our mind, the I only relevance to the litigation is if these allegations about uncertainties nave an impact on our basic conclusions about the alternative to adopt
- encerning the disposal of the AGW. To be relevant, therefore, these subjects aould have to bear on the risk from the evaporation of the water or on the I relative risk in person rem from the two alternatives.
We note [*93] that Interveners have not made the connection between these
- encerns and the ultimate question before us. Nevertheless, we have decided to sift the evidence ourself with this question before us. Again, we are indebted to Applicants for their care in addressing these issues in their findings and we have decided to use their findings as the basis for this portion of our opinion.
A. Tritium Content of AGW Since 1981, the total inventory of AGW has increased to a volume of approximately 2.2 million gallons due to continued additions from support
.I systems and condensation from the reactor building air coolers during the summer months. Care has been exercised to minimize the additions of new water and to ensure that the commingling of non-contaminated water with the AGW is restricted. Even with care, the final volume of water will increase to a limit not expected to exceed 2.3 million gallons. Buchanan, Tr. 456. Approximately 31% of the 2.3 million gallon inventory will be considered for further pre-processing in order to meet the " base case" levels of Table 2.2 in PEIS Supplement No. 2. Buchanan, Tr. 460.
Applicants estimate that the AGW contains no more than 1,020 curies of tritium, which represents an average concentration of 1.2E<-l> uCi/ml in I (*94]
the projected 2.3 million gallons of AGW. Harner, Tr. 166. The Staff, in PEIS Supplement No. 2, utilized a tritium source term of 1,020 curies.
349; Munson, Tr. 741.
Thonus, Tr.
Applicants' estimate was derived during the preparation of GPUN's July 1986 proposal. The most recent sample data from 25 bodies of water was used and the concentration of each body of water was then multiplied by its corresponding tank volume to yield the amount of tritium present in each tank. The total inventories of tritium in each tank were then added to obtain the total curies of tritium in the AGW. The result was a total of 1,180 curies of tritium in the AGW. Correcting the data from July, 1986, to October, 1988, for radioactive I decay, a conservative total tritium curie content of 1,020 was estimated. This estimate is conservative because reductions for normal evaporative losses of I _ -
=_E
=
1989 NRC LEXIS 11, 89
" that time. Tarpinian, Tr. 444. On net, we therefore find that the no-action
!! alternative would save at most 36.4 person rem.
For the purpose of this calculation, we do not consider accident scenarios, cecause they favor Applicants' proposal. The most important accident risk is total breach of a 500,000 gallon storage container, which must maintain integrity for 30 years. Since Applicants' proposal only requires a maximum of
- wo years of storage in similar containers, it is clear that t~ accident risk is greater for the no-action alternative. Keaver, Tr. 479-80; Munson, Tr. 744.
Indeed, testimony indicated a 3.75% risk of release for 30 years from a 500,000 gallon tank (Weaver, Tr. 477), with dose consequences in the first year of storage of 30.7 person rem (bone) and 0.9 person rem (total [*90)
I body).
Jowever, we shall (conservatively) ignore these possible releases during storage in comparing proposals.
I The question is, How are we to weigh the 36.4 person rem savings through the
,o-action alternative? Our answer is that the appropriate legal standard is the one found in Part 50, Appendix I, for interpreting the Commission's ALARA
)rinciple.
Appendix I to 10 CFR Part 50 is entitled, " NUMERICAL GUIDES FOR DESIGN JBJECTIVES AND LIMITING CONDITIONS FOR OPERATION TO MEET THE CRITERION "AS LOW AS IS REASONABLY ACHIEVABLE' FOR RADIOACTIVE MATERIAL IN LIGHT-WATER-COOLED JUCLEAR POWER REACTOR EFFLUENTS." The relevant portion is Sec II, D, which states (in relevant part):
(T]he Applicants shall include in the radwaste system all items of reasonably I demonstrated technology that, when added to the system sequentially and in order of diminishing cost-benefit return, can for a favorable cost-benefit ratio effect reductions in dose to the population reasonably expected to be within 50 miles of the reactor. As an interim measure (still in effect) and until establishment and adoption of better values (or other appropriate criteria) the values $ 1000 per total body man-rem and $ 1000 (*91] per man-thyroid rem (or such lesser values as may be demonstrated to be suitable in a particular case) shall be used in this cost-benefit analysis.
This standard is not, of course, directly applicable; but it provides a useful standard by analogy. Designers of new plants are required to meet this I criterion. Here, we have a volume of radioactive material to dispose of. The same cost-benefit concerns seem to be applicable.
We conclude, therefore, that Applicants would have to spend as much as S I 36,400 for hardware and technology needed to save 36.4 person rem of exposure.
However, the net incremental cost to Applicants of the no-action alternative would be over $ 800,000. Consequently, we conclude that Applicants need not incur this extra expense and that the evaporation proposal is therefore
.I appropriate and should be approved.
We note that there have been a variety of sub-issues argued fully and I carefully, on our record, concerning the accuracy with which tritium and other radioactive content can be measured. In our review of those materials, we became convinced that accurate measurement is being conducted by Applicants, with Staff oversight. We are satisfied that, during (*92] the life of the I project, measurement errors would at most produce a ten percent increase in radiation release. Even if we thought that the release would be increased by
E ll 1989 NRC LEXIS 11, 94 PAGE 37 12.5 curies per calendar quarter were not included. Harner, Tr. 338.
In addition to this 1986 sampling effort, GPUN has since analyzed about 5,000 routine samples of the AGW, including measurements of tritium; these I measurements confirm the [*95) 1986 data. In conjunction with the routine samples analyzed by the GPUN laboratory, periodic independent Quality Control analyses are also performed. The QO techniques include round robin, blind, duplicate, replicate, spiked and split samples. In this way, the accuracy and I precision of the entire analytical process is verified frequently. In addition, a sample was analyzed independently by GPUN's chemistry department and by the U.S. Department of Energy's Radiological and Environmental Sciences Laboratory I
("RESL"), Idaho Falls, Idaho, on behalf of the NRC. This analysis, as discussed further below, is consistent with the GPUN data. Harner, Tr. 167-68.
- 1. Empirical Evidence Challenged With Models The Interveners asserted in Material Statement of Fact 4 (xii), under Contention 3, that the alleged changing source term of the AGW is particularly relevant to tritium. As an example, the Joint Interveners stated that I Applicants relied on data from the PEIS (1981) and EGG-PBS-6798 to get 3,161 curies and 4,231 curies, respectively, while TPO/TMI-043 Rev. 6 (1986) shows that the cover inventory of tritium at t).. time of the accident was 8,794 curies.
All of the estimates on which (*96) Interveners rely are derived from model predictions. Based on these estimates, the Interveners challenged Applicants' ability to accurately determine the tritium source term. See I LBP-88-23, supra, 28 N.R.C. at 199 (1988).
- 2. Empirical Evidence More Reliable Witnesses for each of the parties, however, urged the Board to rely upon sampling data, rather than model predictions, to assess the content of the AGW.
Licensee witness Hofstetter explained that there are many computational methods available in the literature which attempt to model what might occur at the time of a reactor accident having failed fuel exposed to its cooling medium or the atmosphere. Each uses a set of assumptions to define the accident.
Assumptions are made defining core conditions, plant configuration, responses to the event, and duration. As would be expected, each estimate is different due to the assumptions made and computational method utilized. Hofstetter, Tr. 168.
I Two of the early estimates of the tritium source term at the time of the TMI-2 accident are 8,794 curies (TPO/TMI-043, Rev. 6, Data Report - Radioactive Waste Management Summary Review) and 4,231 curies (EGG-PBS-6798, TMI-2 Isotopic
[*97) Inventory Calculations). Both of these estimates are based on I postulated conditions. Hofstetter, Tr. 169.
During cross-examination, Dr. Hofstetter specifically addressed ORIGEN-2's calculation of a tritium inventory of 8,794 curies at the time of TMI-2 accident. Dr. Hofstetter noted that the accuracy of ORIGEN-2, like any code, depends upon the accuracy of the input parameters used to run the code. The applicable codes for ORIGEN-2 include radiation history, length of isotopic enrichments, various core components, length of power operation in the reactor, and mode of power operation.
I _ .
PAGE 38 1989 NRC LEXIS 11, 97 Dr. Hofstetter observed that the number of input parameters is particularly important when estimating the tritium inventory because tritium is not produced in high yield. Thus, the inherent errors in the model are magnified when one is l attempting to observe a relatively infrequent occurrence such as prcduction of i tritium. Hofstetter, Tr. 275. ,
l Similarly, Staff witness Thonus expressed little confidence in the ability of a computer code to accurately estimate the tritium content of the AGW. Mr. j Thonus testified that he believed the estimate of 8,794 curies resulted from a i gross overestimate of (*98) the amount of lithium contained in the uranium dioxide fuel. Thonus, Tr. 403. Mr. Thonus noted that the authors of the ORIGEN-2 analysis assumed the maximum allowable lithium as an impurity in the manufacturing process, when there was no evidence to support such an assumption.
Id. Therefore, Mr. Thonus would not endorse the estimate of 8,794 curies of tritium. Thonus, Tr. 402.
Finally, even Joint Interveners' witness Morgan testified that he always gives more credence to properly conducted sampling than to theoretical estimates. Morgan at 2. Consequently, it is clear that for a radionuclides like
- ritium, Applicants' actual measurements should be relied upon rather than model predictions. Hofstetter, Tr. 277; Thonus, Tr. 349.
Dr. Hofstetter also addressed the tritium source term of 2,910 curies (NUREG-0683. PEIS, 1981). This estimate was based on limited sample data and source volumes in September, 1980. By contrast, the GPUN tritium source term of approximately 1,180 curies, reported in July 1986, was based on much more sample data and source volumes from early 1986. Hofstetter, Tr. 169.
- 3. Use of Upper-Bound Tritium Source Term The estimated upper-bound tritium source (*99] term of 1,020 curies, presented by GPUN in July 1986, and by the Staff in Table 2.2, PEIS Supplement No. 2, is a projection to October 1, 1988, taking into account only radioactive decay. Since tritium is not produced in the fuel after the fissioning procese has stopped, there has been no tritium added since March, 1979. The amount of tritium present continues to decrease through radioactive decay, while some also is released to the atmosphere through normal evaporation. Therefore, the tritium source term estimate of 1,020 curies, which is not corrected for evaporative losses, is an upper-bound value based on actual laboratory analytical results and storage location volumes of virtually all of the AGW at TMI-2. Hofstetter, Tr. 169-70; Thonus, Tr. 349 ("The source term can most accurately be characterized by the series of samples and measures taken by j GPUN.").
Applicants' witness Harner explained the significance of the conservatism in
! the Licensee's estimate of 1,020 total curies of tritium in the AGW. Mr. Harner l noted that GPUN monitors the air leaving the station for its tritium content.
Harner, Tr. 338-39. Applicants' data show that from 1979 through 1986, except for the (*100) first two quarters of 1980 where the Licensee was doing
) specific processes, there has been an average release of 12.5 curies of tritium j per quarter through evaporative losses. Harner, Tr. 338. This conservatism more than compensates for the theoretical possibilities put forward by the Interveners in their arguments that the AGW could contain more than 1,020 curies of tritium.
I.
I LI l
PAGE 39 1989 NRC LEXIS 11, 100
- 4. Program to Overcome Measurement Problems Joint' Interveners' Material Statement of Fact 4 (iv) under Contention 3 asserts in part that it is difficult to acquire an accurate assessment of the tritium concentration of water. LBP-88-23, supra, 28 N.R.C. at 195 (1988). As a general matter, determining the concentration of any element on a liquid sample is an intricate process. Tritium analyses are particularly difficult because tritium emits only very low energy beta particles. When determining tritium concentrations in liquids, two major interferences are (1) the presence of other beta emitting radionuclides in the sample and (2) the variations in
- ounting efficiencies caused by chemical impurities in the sample. both of I these interferences are addressed in TMI-2 Chemistry Procedure 4212-CHM-3013.81,
[*101) which was used to analyze all tritium samples taken by GPUN.
Hofstetter, Tr. 170.
The presence of other radionuclides which also emit beta particles (and/or gamma rays) can give a falsely Lign indication of the amount of tritium present in a sample. Therefore, Chemistry Procedure 4212-CHM-3013.81 provides methods to remove the interfering radionuclides using various methods -- filtration, ion exchange, or flocculation. Hofstetter, Tr. 170.
The second major interference comes from the presence of chemical impurities either in the scintillation cocktail or the water sample itself which can produce a falsely low indication of the amount of tritium present in the sample.
Chemistry Procedure 4212-CEM-3013.81 provides two methods which control these
- hemical interferences. In one method, ascorbic acid is added to the sample to react with any organic material and minimize its effect on the detection efficiency. In the other method, a known amount of tritium is added to a second aliquot of the sample. This spiked aliquot is used to determine the exact counting efficiency fc; the detector in the sample matrix. Once this is known, the tritium concentration of the unspiked sample can [*102] be determined more accurately. Hofstetter, Tr. 171.
- 5. Reasons for Different Tritium Measurements The Interveners contended in Material Statements of Fact 4(viii) and 4(x) under Contention 3 that the Staff invalidly assumes that the average concentrations of radionuclides as shown in Table 2.2 of PEIS Supplement No. 2 can be reasonably considered a maximum. Joint Interveners cite as an example the average concentration of tritium in Table 2.2 as 1.3E<-1> uCi/ml, while a document obtained from GPUN during discovery reports a measured tritium concentration in the PWST-2 storage location of 2.1 uCi/ml. This was also the I only issue remaining under Contention 4(b). See LBP-88-23, supra. 28 N.R.C. at 198-99, 204 (1988).
The record shows that the values for radionuclides listed in Table 2.2 of I PEIS Supplement No. 2 are reasonable projections of concentrations expected to exist in the influent to the evaporator. Harner, Tr. 172; Manson, Tr. 782.
Licensee compared the projections of concentrations for processed water in Table 2.2 against the RESL analyses of SDS and/or EPICOR II processed water in the I PWST-2 tank. The actual analytical results for PWST-2 processed water (*103) were lower than the projections of Table 2.2 for all radionuclides except tritium. This demonstrates that the data from Table 2.2 is a reasonable estimate which can be used for dose calculations.
lI .. -
PAGE 40 1989 NRC LEXIS 11, 103 Since the amount of tritium is not reduced by water processing, the final concentration in each storage location after processing is the same as the tritium influent concentration. Therefore, depending on which of these locations is being processed, the tritium concentration will vary. The average tritium concentration is then a mathematical calculation and must be done in place of the capability to undertake a physical mixing of the entire AGW volume in a single, homogenous batch. Harner, Tr. 172.
Second, the tritium concentrations at issue cannot be compared directly to each other. The values are a mixture of actual sample results, as well as estimates calculated using averaged numbers, actual numbers and other input at various points in time. The tritium value of 1.3E<-1> uCi/ml was a mathematical projection of the average tritium concentration of all the AGW on October 1, 1988, and was based on 2.1 million gallons of AGW. The tritium value of 2.3E<-1> uCi/ml was an actual (*104] analytical result obtained by GPUN for a PWST 2 sample (GPUN sample number 86-15668) on October 27, 1986. It was also the result obtained for a tritium analysis performed for a sample of PWGT-2 (GPUN sample number 86-17062) on November 21, 1986. Harner, Tr. 172-73.
We find that one of the tritium values supplied to Interveners by Applicants was an error. The tritium value of 2.1 uCi/ml was an error on page 41, step 7.3 of a GPUN document t.tled, " Technical Specification for Processed Water Disposal for GPU Nuclear Corporation Three Mile Island - Unit 2 Nuclear Power Plant."
This document, produced during discovery, is a bid specification sent to vendors and provides information necessary for them to evaluate the task and return a proposal for construction. Based on Mr. Harner's testimony that he checked the lab results, we find that no sample from TMI-2 ever contained tritium as high as 2.1 uCi/ml. The correct value for the sample in question is 2.3E<-1> uCi/ml.
Harner, Tr. 173; Thonus, Tr. 349; Munson, Tr. 751-52. (We considered whether this error had an adverse effect and we conclude that since tritium is not removed by the proposed AGW disposal system, the error in the (*105) bid specification had no effect on the evaporator proposal desigr.. Harner, Tr.
215.)
B. Sampling Accuracy Joint Interveners' Material Statement of Fact 4(vi) and (vii) under Contention 3 stated that the analyses of the PWST-2 samples by RESL for the NRC and by GPUN gave differing results for Co-60, Cs-137, and Sr-90, and that neither detected C-14, whereas an analysis by Westinghouse found C-14 at a concentration of 3.0E-4, greater than the average concentration listed in Table
> 2.2 of PEIS Supplement No. 2 by a factor of three. See LBP-88-23, supra, 28 N.R.C. at 198 (1988).
A sample of PWST-2 was obtained on February 23, 1987 (GPUN) sample nurber 87-02569). A portion of the sample was transferred to the NRC for analysis by their contracted laboratory, RESL. A comparison of the GPUN data and the RESL data can be done on positive (greater than lower-level-of detection) results for individual radionuclides. One method for comparison is outlined in the NRC
) " Inspection and Enforcement Manual," Inspection Procedure 84725. This method is
) used to evaluate a licensee's analytical capability to make consistently accurate radioactivity measurements. The licensee's measurement (*106] is compared to a NRC or RESL measurement and a determination is made whether the two values are close enough to be in agreement. The "egreement" criteria are based on an empirical relationship which combines prior experience and the l,
1
PAGE 41 1989 NRC LEXIS 11, 106 accuracy needs of the program. Harner, Tr. 175-76.
Applying the procedure to the RESL and GPUN data for the PWST-2 sample shows agreement for the tritium, Co-60 and Cs-137 results. The differences observed between these two sets of data for tritium, Co-60 and Cs-137 are not major and, in fact, are within the range of normal differences observed when comparing radiochemistry data from two separate laboratories in accordance with NRC standards. Harner, Tr. 176, 259-60; Thonus, Tr. 348.
The difference between the GPUN data and RESL data for Sr-90 is explainable from the procedures used by the two laboratories. In most cases, strontium-90 analyses on liquid samples are done through the use of the 14-day method, where one separates the strontium and the yttrium from er.ch other and then watches the yttrium, which is the daughter of strontium, grow back over time.
The more-accurate 14-day method is not practical for process control use at TMI. ,cner, (*107] Tr. 176, 261. Thus, GPUN uses either a 72-hour or a rapid method to analyze for Sr-90 in liquid samples. GPUN quality control data for the liquid Sr-90 analysis has shown the 72-hour method to be more accurata than the rapid method; however, both are inherently conservative. The 72-hour analysis results average 40% higher than the actual, while the rapid method results average 100% higher then the actual concentrations, at the 1E-4 uci/ml Sr-90 level, when compared to standards or the results of split samples reported by off-site laboratories using more traditional methods. Harner, Tr. 176, 261-62.
In any event, both the GPUN and the NRC results for this Sr-90 analysis are lower than the value used in the PEIS to evaluate the environmental impact of Sr-90. Thonus, Tr. 348, 370. In addition, the environmental impact analysis on the PEIS was not calculated from these single sample analytical results. The average or " base case" data was conservatively calculated from many sample results and assumptions made from knowledge of a long history of water processing through the SDS and EPICOR II systems. Harner, Tr. 177.
The C-14 concentrations at issue vary because the values are (*108) a mixture of actual sample results, averaged numbers and estimates or projections calculated from various input. The C-14 value of 3.0E-4 uCi/ml is an actual sample result obtained from a PWST-2 sample (GPUN sample number 85-16198) analyzed by Westinghouse. The sample date is December 23, 1985. The C-14 value of <2.0 E<-7) uCi/ml is an actual sample result obtained from a PWST-2 sample (GPUN sample number 87-02569), taken in February 24, 1987 and analyzed by RESL.
The C-14 value of 1.0E-4 uCi/ml found in Table 2.2 of NUREG-0683, Supp. 2, is a calculated estimate based on operational experience of SDS and EPICOR II Systems and on the average of four processed water analyses performed by Westinghouse in early 1986.
A calculated projection of concentration for the total volume is not the same as an actual sample result of a single portion of water. The C-14 values listed vary, but the NRC and GPUN appropriately used the 1.0E-4 Westinghouse average concentration for dose estimates, even though RESL measured a much lower value.
Hofstetter, Tr. 177-78; Thonus, Tr. 348.
- 1. Staff Checks on Accuracy of Measurement
)
~~ - - - - ~~ - -~
PAGE 42 1989 NRC LEXIS 11, 108 53 In Material Statement of Fact 4(11) under Contention 3, the Interveners us (*109] argued that the NRC's sampling of the AGW was inadequate because the NRC took'a four liter sample from a tank which contains one-fifth of the AGW.
LBP-88-23, supra, 28 N.R.C. at 198 (1988). The testimony during the hearing demonstrated that the NRC has analyzed a representative sample of water available for evaporation.
In February, 1987, the water in the PWST-2 was sampled. The sample of PWST-2 I was divided between the NRC and GPUN. Harner, Tr. 178; Thonus, Tr. 368.
the PWST water has undergone treatment through EPICOR II or through SDS and EPICOR II. Harner, Tr. 304-06. While some variations in concentrations of All of I radionuclides occur among batches of AGW processed through SDS and EPICOR II, the average concentrations resulting from this processing have not been significantly different. Therefore, the PWST-2 vater is representative of the TMI-2 water available for evaporation, and it has been evaluated by the Staff.
Harner, Tr. 178.
Moreover, the NRC did not intend to use the PWST-2 sample to provide bounding radionuclides limits on the AGW. The NRC does not have to develop its own data for environmental statements. GPUN's more extensive sampling provided the
[*110) basis for the estimated influent to the processed water disposal system. The NRC sample in question here was taken only to audit GPUN data. The results of this sample do support the information supplied by GPUN. See Thonus, I Tr. 347, 356, 358.
At the hearing, the Staff also explained that as part of the NRC inspection program, NRC inspectors come on site, split samples with the Applicants, and
'I independently confirm Applicants' laboratory analysis method.
The results of the NRC inspection program are published in the NRC's inspection Thonus, Tr. 359.
The Board, sua sponte, requested and reviewed these inspection reports. Id.
- reports dating back to 1981. Tr. 359-60. The Board was satisfied that the NRC had adequately audited the Licensee's laboratory analysis program. Tr. 1630.
Joint Interveners' Material Statement of Fact 4(111) and (iv) claimed that I procedures presented in ASTM Method 3370 and 4212-CHM-3013.81, p.
were not followed when the AGW was sampled in February, 1987. See LBP-88-23, supra, 28 N.R.C. at 195 (1988).
5.0, 6.1.7 Applicants' and Staff's testimony showed that the Interveners' assertion is incorrect. Harner, Tr. 179; Thonus, Tr. 348.
In [*111) February, 1987, a sample was removed from the recirculation flow path of PWST-2 in accordance with TMI-2 Chemistry Procedure 4212-CHM-3011.05, Rev. 0 (5/23/84) entitled, " Sample procedure." This procedure I explains how one obtains a representative sample from a plant system. The procedure does not concern analysis or testing of the water. It is used simply to obtain a bottle full of water from a well-mixed larger volume. Harner, Tr.
Step 2.4 of 4212-CHM-3011.05, Section 2.0, References, includes " ASTM I 179.
Section D 3370-82" as one of four ASTM standards followed and incorporated into 4212-CHM-3011.05. Harner, Tr. 179; Thonus, Tr. 348.
I TMI-2 Chemistry Procedure 4212-CHM-3013.81, entitled " Determination of Tritium By Liquid Scintillation Counting," provides instructions on how to determine the tritium concentration of the sample. The then-current version of this procedure was followed by GPUN when the PWST-2 sample was transported to the laboratory for analysis. Harner, Tr. 179; Thonus, Tr. 348.
'I I l PAGE 43 1989 NRC LEXIS 11, 111 l The Joint Interveners stated in Material Statement of Fact 4(iv) under .
Contention 3 that Chemistry Procedure 4212-CHM-3013, 81-P 5.0, 6.1.7 (10/27/87) was an updated version of Chemistry (*112] Procedure 4212-CHM-3011.05, Rev. 0 -
(5/23/83). See LBP-88-23, supra, 28 N.R.C. at 395 (1988). The evidence shows, however, that Chemistry Procedure 4212-CHM-3013.81 is not an updated version of -
4212-CHM-3011.05. They are distinct procedures that give directions for two entirely different tasks. Chemistry Procedure 4212-CHM-3011.05 explains how a l3 liquid sample is drawn from a plant system. Chemistry Procedure l lg 4212-CHM-3013.D1 provides step-by-step instructions which must be followed to I produce an accurate and reproducible determination of the tritium concentration in a given aliquot of sample. Harner, Tr. 180.
C. Microorganisms l
In Material Statement of Fact 9 under Contention 3, Joint Interveners ll l
lE contended that an evaluation of microorganisms in the AGW must be undertaken.
LBP-88-23, supra, 28 N.R.C. at 200 (1988). The completely uncontroverted testimony in the record demonst ates, and the Board finds: that the microorganisms associated with the AGW are typical environmental microbes and
! not primary pathogens; that in any case the vaporizer section will heat the
- distillate to approximately 240 degrees F; and that any microorganisms which can survive this temperature (*113) are not pitthogenic to humans. Baker, Tr.
645-46; Masnik, Tr. 349-51.
D. Conclusion 3 In sum, the Board finds that the tritium content of the AGW has been g conservatively determined from the results of actual measurements, and that the accuracy of GPUN's sampling and analysis program has been verified independently. Contrary to the assertions remaining in Joint Interveners' Contentions 3 and 4b, the radionuclides content of the AGW has been adequately characterized to support the environmental comparison of disposal options.
VII. Conclusions of Law Although we believe that we have already presented all the findings of fact and conclusions of law required of us, we have decided to adopt Conclusions of Law, based on those provided to us by the Staff, in order to include the principal conclusions of law all in a convenient place.
- 1. All Issues Resolved This is a contested proceeding on an application for an amendment to an operating license for a utilization facility. The Board has decided all of the admitted matters in controversy raised by the Interveners wfthin the scope of I NEPA and the Commission's regulations implementing NEPA in 10 C.F.R. Part 51.
See 10 C.F.R. Sec. (*114] 51.104(a)(3). The Board has also decided the public health and safety matters raised by the Interveners. ,
- 2. Absence of a Prehearing Conference Since this case involves an amendment of an existing operating license and is not a construction or operating license proceeding, a prehearing conference was I permissive and not mandatory. 10 CFR Sec. 2.752(a). Furthermore, telephone conferences were held prior to the evidentiary hearing and there was no I _ -. .
l PAGE 44 l 1989 NRC LZXIS 11, 114 prejudica, alleged or actual, to Interveners from the absence of a prehearing conference.
- 3. Witnesses All of the witnesses et the evidentiary hearing were qualified for the subjects on which they testified and we have accorded each the evidentiary weight that we found their testimony to be worth.
- 4. The Record The record of decision on this application includes the PEIS, Supplement No.
2 (Staff Ex. 1), as supplemented and amended by this Initial Decision, but we admitted into evidence only those portions of the PEIS specifically referenced by witnesses as part of their testimony. See 10 C.F.R. Secs. 51.102(c),
51.103 (c) ; Philadelphia Electric Co. (Limerick Generating Station, Units 1 and 2), ALAB-819, 22 NRC 681 at 706-707 (*115] (1985), review declined, CLI-86-5, 23 NRC 125 (1986).
- 5. Health, Safety and Common Defense Pursuant to 10 C.F.R. Secs. 50.92 and 50.57, the Board concludes, with respect to the matters in controversy, that:
(a) there is a reasonable assurance that the activities that would be authorized by the amendment can be conducted without endangering the health and safety of the public and would be in compliance with Commission regulations; and (b) issuance of the amendment would not be inimical to the common defense and security or to the health and safety of the public.
- 6. Consideration of Entire Record In reaching this Decision; the Board has considered the entire record of the proceeding. All proposed findings presented by the parties and not addressed in this Decision are considered to be without merit or unnecessary to the Decision.
The Board's findings of fact are supported by reliable, probative and substantial evidence in the record.
- 7. Conclusion All relevant matters have been decided in favor of the Applicants and, I therefore, the requested amendment should be authorized.
VIII. ORDER l
) For all the foregoing reasons and upon consideration of the entire record in
(*116] this matter, it is, this 2nd day of February 1989, ORDERED, in accordance with 10 C.F.R. Secs. 2.760 and 2.762:
- 1. That this Initial Decision shall constitute the final action of the Commission forty-five (45) days after its date unless an appeal is taken in accordance with Sec. 2.762 or the Commission directs that the record be certified to it for final decision.
PAGE 45 1989 NRC IIXIS 11, 116
- 2. Any notice of appeal from the Decision must be flied within ten (10) days after service of the Decision. A brief in support of the appeal must be filed within th~1rty (30) days (forty (40) days if the Staff is the appellant) after the filing of the notice of appeal. Any party not an appellant may file a brief in support of, or in opposition to, the appeal within thirty (30) days (forty (40) days in the case of the Staff) after the period has expired for the filing and service of the briefs of all appellants.
- 3. In accordance with the Commission order in this proceeding, dated December 3, 1987, the parties to the proceeding shall have ten (10) days after service of this Decision to file written comments with the Commission regarding whether the Decision should be effective during the pendency of administrative
(*117) appellate review. .
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$9 t%Y 26 P4 30 Certificate of Service ,,n q . _
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I, R. K. Gad III, hereby certify that on May 23,1989, I made service ofE ^"
the within document by depositing a copy thereof with Federal Express
- (except where indicated by an asterisk, in which case by mailing a copy ther or "o't==e are"*id) a' ro""*':
I p' Charles Bechhoefer, Chairman Gustave A. Linenberger, Jr. /
j Administrative Judge Administrative Judge Atomic Safety and Licensing Panel E Atomic Safety and Licensing Panel U.S.N.R.C. U.S.N.R.C.
l East West Towers Building East West Towers Building -
4350 East West Highway 4350 East West Highway Bethesda, Maryland 20814 Bethesda, Maryland 20814 I James H. Carpenter Administrative Judge Atomic Safety and Licensing Panel Anne Spielberg, Esquire Harmon, Curran & Tousley Suite 430 U.S.N.R.C. 2001 S Street, N.W.
East West Towers Building Washington, D.C. 20009 4350 East West Highway Bethesda, Maryland 20814 Adjudicatory File
- John Traficonte, Esquire Atomic Safety and Licensing Board Assistant Attorney General Panel One Ashburton Place U.S.N.R.C. Boston, Massachusetts 02108 Washington, D.C. 20555 Geoffrey M. Huntington, Esquire Samuel H. Press, Esquire Environmental Protection Bureau George E. Young, Esquire State House Annex Vermont Department of Public -
25 Capitol Street Service Concord, New Hampshire 03301 120 State Street .
Ann P. Hodgdon, Esquire Patricia A. Jehle, Esquire I U.S.N.R.C.
One White Flint North 11555 Rockville Pike '
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l p7 Rockville, Maryland 20852 I
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R. K. Gad til
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