ML20205Q948
| ML20205Q948 | |
| Person / Time | |
|---|---|
| Site: | Maine Yankee |
| Issue date: | 01/31/1987 |
| From: | Eng R, Grant D, Ostrow S EBASCO SERVICES, INC. |
| To: | |
| Shared Package | |
| ML20205Q916 | List: |
| References | |
| APTR-42, NUDOCS 8704060139 | |
| Download: ML20205Q948 (124) | |
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APTR-42
- Maine Yankee Atomic Power Station 1
! In-Situ Decommissioning / Low Level Waste Management l
Topical Report i
Prepared For Maine Yankee Atomic Power Company By l j EBASCO SERVICES, INC. l
! 1 S.L. Ostrow I R.Eng l T.J. Grant
- L. Skoski G. Buniak l
January 1987
,EM*28ei82888.
PDR j
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, EXECUTIVE
SUMMARY
This study defines and examines the technical feasibility of a concept which would combine low-level radioactive' waste
$ management and disposal with the ultimate decommissioning of the Maine Yankee nuclear plant at its scheduled end-of-life.
- Currently, all low-level radioactive waste produced in Maine is processed,' packaged and shipped to an out-of-state repository I where it is buried. However, for the purposes of this study, it is assumed that beginning January 1, 1993, this will no longer be possible under the terms of the Federal Low Level .
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Radioactive Waste Policy Act, which will make low-level radioactive waste a state responsibility. ~Many states are
. joining into compacts, but the strong possibility exists that Maine will go it alone and will have to provide a radioactive 7
waste repository in Maine.
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) Low-level radioactive waste is generated in the State of Maine j by several medical, research and commercial facilities that use radioactive isotopes in their work. The largest single. source l
in the state is -the Maine Yankee Atomic Power Plant. This power i plant, located near Wiscasset Maine, uses a pressurized water
, reactor to produce 850 MW of electricity. It has been in J
operation since 1972, and it produces various types of
, ? low-level radioactive waste during operation, such as filters, l
j ion exchange resins, evaporator bottoms and contaminated
] clothing and trash. In addition to the waste produced during
! operation, a large volume of waste will be produced when the
! plant is decommissioned at the end of its licensed lifetime.
i i This study examines a proposed radioactive waste disposal I concept that attempts to make maximum utilization of the 1
existing Maine Yankee site. The proposed concept, called j " low-level waste management and in-situ decommissioning", would I involve storing all of Maine's low-level radioactive waste in a l
L temporary storage facility at the Maine Yankee site. When the i
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- 1 1 the waste would be.. permanently
,p lant is decommissioned, all disposed of (" encapsulated") inside the Containment Building of '
the plant. This structure is a massive concrete building that'
- , .was designed to contain- radioactivity in the- event of a j
a postulated accident.-Waste disposal would proceed as follows in three distinct phasess.
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(1) Operational Period (1993-2008) - During the operating period for the Maine Yankee Atomic plant, all-low-level radioactive waste generated in the State of Maine would be
- shipped to the Maine Yankee site and temporarily stored-along with Maine Yankee operational, low-level waste in an 4
existing Onsite Storage Facility building that was l
- designed to store low-level waste. This' building may have l to be expanded or supplemented to accommodate the additional waste volume. The amount of waste stored would j continue to be added to (although the radioactivity would steadily decay away with time) until the Maine Yankee plant license expires (October 21, 2008).
l d (2) Decommissioning Period (2009-2011) - Decommissioning of the Maine Yankee plant would begin in 2009 and last for an assumed three years. Following removal of the fuel l
f and draining of all f assemblies and reactor- internals, radioactive liquids, all plant structures outside the f
l Containment Building (a massive concrete structure with 4-1/2 foot thick cylindrical sides and a hemispherical l g I dome, which houses the pressurized ' water reactor and pri= cry system piping and component:) would be dismantled.
] That Containment would be prepared by clearing
! non-radioactive equipment out of certain areas, installing 6
0 racks, and reinforcing floors (if needed), and cutting and i
sealing all penetrations through its outer wall, except those necessary to support facility use as an active
, low-level waste repository.
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O O All radioactive waste (" decommissioning waste" plus waste already stored in the Onsite Storage Facility) would be moved into the Containment and stored in various areas.
The most highly radioactive waste (all of the Class C and most of the Class B) would be placed in the 8-5/8 inch thick steel-walled and stainless steel clad reactor vessel, and then the head put back in place. All the' remaining Class B waste would be placed into the steel-lined refueling cavity and capped with a concrete barrier (if required). This, together with the Containment Building, would provide two formidable barriers separating the highest level waste from the environment.
All of the lowest activity waste (Class A) would be placed in the outer, annular areas of the two floors below the operating deck, and on the operating deck itself. A combination of the overhead crane and forklift trucks would move the waste.
(3) Repository Period (2012-2041)
- During this period, the I facility, now licensed as an engineered, low-level waste repository, would be operated by a State low-level waste authority and continue to receive waste from all State of Maine generators on a periodic basis. This period could be extended as far into the future as desired, or until the Containment ran out of storage space, but it is assumed here that it would operate for at least 30 years.
Starting in 2042, the facility would be permanently closed. In accordance with current, 10 CFR 61 regulations, environmental monitoring would be continued for an additional five years. Then, the remaining penetrations would be sealed, and a 100 year period of institutional a control would begin.
Some of the major issues resolved in assessing technical feasibility are:
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(1) The existing Onsite Storage Building can be expanded, or an additional building constructed, to house the waste until the plant's end-of-life.
(2) The Containment can be prepared for waste storage (spaces cleared, floors supported if required, and the building sealed), adequate storage space exists for the all the waste, and the waste can be moved into the Containment and
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put in place.
(3) The Containment can provide adequate, long-term protection for the stored waste, meeting all regulatory requirements protecting the public from exposure.
This study concludes that the concept appears to be technically feasible, and, in addition, has several attractive features which make it worth considering furthers (1) It presents a viable solution to the problem of what to do with Maine low-level waste generated after January 1, 1993 when the State of Maine must assume responsibility for disposal of low-level waste produced within the state.
It is assumed that it can no longer be shipped out-of-state.
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(2) It avoids having to create another licensed nuclear j facility in Maine, in addition to the Maine Yankee site.
(3) It makes maximum utilization of existing facilities, ;
security, technical personnel and monitoring already located at the Maine Yankee site.
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(4) Radiation exposures from waste handling, transportation and storage for workers and the population would be lower.
than if a repository were located at a different site since most of the waste generated in Maine originates at Maine Yankee.
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(5) It would minimize the amount of radioactive waste
! shipments that would go over public roads.
(6) The "in-situ" decommissioning concept of leaving the Containment Building standing should result in a lower occupational exposure during decommissioning than the currently favored immediate dismantlement option.
(7) The long-term protection afforded to the public by the Containment Building should be equal to, if not better than, the shallow land burial method, which is the primary alternative.
(8) Although cost was never quantified in the study, it is reasonable to expect that onsite storage at Maine Yankee would be less expensive than constructing a facility elsewhere and shipping Maine Yankee waste to it. Also, in-situ decommissioning of the plant should be less expensive than immediate dismantlement (which must decontaminate the Containment, remove all radioactive material, then demolish the very massive building).
1 This study examined just technical feasibility. Further consideration of the concept would call for a detailed engineering study (including a realistic radiation dose assessment), a cost estimate, and a licensing assessment to identify the steps necessary to obtain proper regulatory approval for the various aspects of the concept.
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TABLE OF CONTENTS SECTION CONTENTS PAGE EXECUTIVE
SUMMARY
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1.0 INTRODUCTION
1-1 1.1 Purpose of Study 1-1 1.2 Scope of Study 1-2 1.3 Description of Concept 1-3 1.4 Low Level Waste Management and 1-4 Decommissioning Background 1.4.1 Maine Low-Level Waste Compact 1-4 1.4.2 Regulatory Overview 1-4 1.4.2.1 Storage 1-4 1.4.2.2 Decommissioning 1-6 1.4.2.3 Disposal 1-6 1.4.3 Relevant Decommissioning and Low-Level 1-7 Waste Studies 1.5 Report Contents 1-9 1.6 References 1-10 2.0 MAINE YANKEE ATOMIC POWER PLANT 2-1 2.1 Ownership and Location ~2-1 2.2 Major Structures and Equipment 2-2 2.3 References 2-6 3.0 RADIOACTIVE WASTE CHARACTERIZATION 3-1 3.1 Maine Yankee Operational Waste 3-1 ;
3.1.1 waste Types 3-1 3.1.1.1 Ion Exchange Resins 3-2 3.1.1.2 Concentrated Liquids 3-3 3.1.1.3 Filter Sludge 3-3
, 3.1.1.4 Cartridge Filters 3-4 3.1.1.5 Trash 3-5 3.1.2 Waste Volumes and Classifications 3-5 vi
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TABLE OF CONTENTS (cont'd) i SECTION CONTENTS PAGE 3.1.3 Waste Forms and Containers 3-7 3.2 Other Waste Generators 3-8 3.3 Total Non-Decommissioning Waste Summary 3-11 3.4 Maine Yankee Decommissioning Waste 3-12 3.4.1 Types of Decommissioning Waste 3-13 3.4.2 In-Situ Decommissioning Wasta 3-15 3.5 References 3-18
) 4.0 ENCAPSULATION: IN-SITU DECOMMISSIONING 4-1 AND LOW LEVEL WASTE MANAGEMENT 4.1 Description of Concept 4-1
- 4.1.1 Operating Period 4-2 4.1.2 Decommissioning Period 4-3 4.1.3 Active Repository Period 4-4 4.1.4 Post-Closure Period 4-5 4.2 Encapsulation Boundary 4-5 4.3 Sealing Preparations 4-6 i
4.4 Structures and Equipment Inside 4-7 Containment i
4.5 Reactor Internals 4-7 4.6 Waste Handling 4-8 4.6.1 Operational Waste 4-8 4.6.2 Decommissioning Waste 4-9 4.6.3 Disposal 4-10
'. 4.7 durveillance, Maintenance and decurity 4-11 4.7.1 Storage Period 4-11 4.7.2 Decommissioning Period 4-12 I 4.7.3 Final Closure Period 4-12 4.8 References 4-14 l
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. TABLE OF CONTENTS (cont'd)
SECTION CONTENTS PAGE 5.0 TECHNICAL FEASIBILITY OF CONCEPT 5-1 4 5.1 Low-Level Waste Storage During Normal 1 Operation i 5.2 Encapsulation Boundary 5-2 5.3 Preparation of Containment for Waste 5-2 Storage 5.4- Waste Placement and Storage 5-5 5.4.1 Waste Volumes and Storage Areas 5-6 5.4.2 Waste Handling 5-7 5.5 Containment Integrity 5-8 5.6 Environmental Surveillance 5-10 l
j 5.7 Waste Integrity 5-11 j 5.8 References 5-12 i
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} 6.0 RADIOLOGICAL ASSESSMENT 6-1
! 6.1 Transportation Doses. 6-2 6.1.1 Population Transportation Dose 6-2 6.1.2 Occupational Transportation Dose 6-5 1 6.2 Disposal Doses 6-6 i
6.2.1 Population Disposal Dose 6-6
! 6.2.2 Occupational Disposal Dose 6-7 6.3 Decommissioning Doses 6-7 6.4 Population Dose From Containment Storage 6-9 I
6.5 References 6-11 i
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7.0 CONCLUSION
S AND RECOMMENDATIONS 7-1 1 1
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FIGURES FIGURE TITLE PAGE l Site Plan 2-14 2-1 2-2 Plot Plan 2-15
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2-3 Reactor Containment Plan El. 46'-0" 2-16 2-4 Reactor Containment Plan El. 20'-0" 2-17 2-5 Reactor Containment Plan El. -2'-0" 2-18 2-6 Reactor Containment Elevation - Sh 1 2-19 2-7 Reactor Containment Elevation - Sh 5 2-20 2-8 Reactor Vertical Arrangement .2-21 2-9 Primary Auxiliary Building - Sh 1 2-22 I
2-10 Primary Auxiliary Building - Sh 2 2-23 i
2-11 Fuel Building - Sh 1 2-24 2-12 Fuel Building - Sh 2 2-25 2-13 Onsite Storage Facility Location 2-26 4-1 In-Situ Decommissioning and Low Level 4-15 I
Waste Management Time-Line J 4-2 Decommissioning Boundary 4-16 I
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TABLES TABLE TITLE PAGE 2-1 Maine Yankee Design Characteristics 2-7 3
3-1 Typical PWR Trash 3-19 3-2 Specification For Water Purification 3-20 Systems t
3-3 Maine Yankee Waste Isotopic Activity 3-21
. 3-4 Maine Yankee Low-Level Radioactive 3-22 Waste Generation 3-5 Maine Yankee Waste Classification 3-23 3-6 Maine Yankee Shielded vs. Unshielded 3-24 Shipments 3-7 Maine Yankee Spent Resins 3-25 3-8 Maine Yankee Evaporator Bottoms 3-26 I
3-9 Maine Yankee Filter Cartridges, Sludges 3-27 3-10 Maine Yankee Dry Compressible Waste 3-28 (Contaminated Trash) 3-11 Maine Yankee Irradiated Components 3-29 3-12 Maine's Low Level Waste 3-30 3-13 Characterization of Low-Level Radioactive 3-31 Wastes 3-14 Maine Yankee Decommissioning Waste 3-32 3-15 Isotopic Distribution of Decommissioning 3-33 Waste 5-1 Containerized Low Level Waste Streams 5-13
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5-2 Distribution of Waste in Containment 5-14 4 I
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1.0 INTRODUCTION
- ! 1.1 Purpose of Study i
i This report defines and examines for the first- time the technical feasibility of a concept involving the in-situ, I decommissioning of a nuclear power plant and the encapsulation j within the containment building of all low-level radioactive
! waste generated by the plant and by other facilities in the t
state. This study applies the concept specifically to the Maine Yankee nuclear plant and the ultimate disposition of low-level
! waste generated within the State of Maine. Thus, this concept f addresses the issues of decommissioning method and of low-level waste storage and disposal.
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I Currently, all Maine Yankee low level waste is processed, t
packaged and shipped to an out-of-state repository. Other Maine i generators of low level waste follow a similar disposal i procedure. However, under the terms of the Low Level l Radioactive Waste Policy 'Act Amendments of 1985, beginning-
! January 1, 1993, the disposal of Maine-generated low-level j waste will become the responsibility of the State which, it is
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d e addition, Maine Yankee (like all other nuclear facilities) must I
I eventually be decommissioned at the end of its useful life.
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! low-level waste generated by Maine Yankee and by others within
! l the .L.Le will be = Luc.a ' .L Lhe plenL .its bcginning Jnnunry 1, 1993. During decommissioning, all radioactive components I outside the containment would be removed and stored within that t f structure. All low-level waste generated as_ a result of l ,
decommissioning, along with the waste already being stored at I the sito, would be placed within the Containment. The f
! Containment would be appropriately modified and sealed to serve t
as the State low-level waste repository thereafter.
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1.2 Scope of Study The focus of the study is on the technical feasibility of the proposed course of action of combining low level waste storage with nuclear plant decommissioning. It uses specific Maine Yankee and other Maine facilities data, along with generic information and studies, appropriately modified.
Detailed engineering is not performed. Issues such as economics, licensability (state and federal), and ALARA (radiation dose minimization) are not treated quantitatively or explicitly, although they are qualitatively considered in evaluating the feasibility of the concept. Optimization is similarly not performed.
Three principal assumptions are made:
(1) Following the terms of the Low Level Radioactive Waste Policy Act, Maine producers of radioactive waste will no longer be able to shio to an out-of-state repository beginning in 1993. (The federal government, of course, may at any time modify its current position).
(2) The Maine Yankee plant reaches its planned end-of-life in 2008.
(3) The DOE facilities at Hanford, Washington (or elsewhere) will still accept the highly radioactive ( > Class C) reactor internals.
Altheegh nuscrical results may be alterad, thc conclusions of f
the study are relatively insensitive to small changes in actual dates for events to occur.
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s 1.3 Description of Concept Beginning in 1993, it is expected that Maine low level waste will no longer be permitted to be shipped to an out-of-state l repository. Waste generated thereafter (from Maine Yankee and other Maine facilities) would be stored in the Onsite Storage i Facility located on the Maine Yankee site. This would continue until the plant reaches its end-of-life, nominally set in 2008.
Decommissioning would begin in 2009 and last for three years.
Fuel assemblies and highly radioactive reactor internals would be removed and shipped to a DOE facility. All radioactive liquids would be drained and processed as waste. Radioactive components outside the Containment would be dismantled, and all radioactive waste moved into the Containment. Finally, all penetrations, with the exception of those necessary for the use of the structure as a low-level waste repository, would be permanently sealed. The facility would then be turned over to a State low-level waste authority for its continued use as a repository.
t Low-level waste generated in the State for a period of 30 years (assumed) after decommissioning would be moved into the Containment on a periodic basis. Finally, the facility would stop accepting waste, the remaining penetrations would be sealed, and the facility, consisting of a sealed containment containing low-level waste, would be left intact as an inactive low level waste repository. It would be monitored thereafter by the State low-level waste authority for an indefinite period of time, i
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'l.4 Low Level Waste Management and Decommissioning Background g
1.4.1 Maine Low Level Waste Compact Maine is not currently within a regional compact, nor is negotiating with other states to form such a compact. The Federal Low Level Waste Policy Act requires that states assume responsibility for their waste by 1993. Therefore, it is expected that Maine authorities will establish a separate entity to develop and operate a low-level waste disposal facility located in Maine limited to c:dy Maine-generated 'l l
waste.
1.4.2 Regulatory Overview The regulatory framework which currently governs the concept of in-situ decommissioning and low-level waste disposal is composed of those regulations applicable to three separate, yet complimentary elements--storage, decommissioning, and disposal.
l These will be reviewed below.
1.4.2.1 Storage i
Regulations pertaining to the storage of low-level waste are separated into two categories depending on whether or not the waste is generated at power reactor sites. Guidance for the l former is presented in Generic Letter 1 (81-38), " Storage of Low I
Level Waste At Power Reactor Sites", which was updated from a June 1980 draft, " Safety Consideration for Temporary On-Site 2
Storage of Low Level Radioactive Waste". Two time periods for storage are considered: less than five years and greater than I
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five. years. 'The following are the salient. points for less than f
five years storage:' ,
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- q (1) A proposal to increase storage capacity must be evaluated
'by the licensee under the provisions of 10 CFR 50.59.
M (2) The evaluation should follow the guidance provided~in the enclosure in Generic Letter (81-83) - " Radiological Safety Guidance for Onsite Contingency Storage Capacity".
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j (3) The Guidance Document ~ is incorporated into the NRC i Standard Review Plan.
! NRC estimates of the time requirements for permitting of j
F interim storage facilities are: none under 10 CFR 50.59, and l
1 nine months under 10 CFR 30. Construction time for new or expanded facilities is on the order of one year.
j It should be noted that the guidance document (81-83 enclosure) j assumes the possibility of storage of wet radioactive waste, .
l solidified radioactive waste and dry radioactive waste. The f categories are not compatible with disposal classes A, B, and C
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under 10 CFR 61. Any incompatibilities of this sort would be i resolved for the concept proposed in this report.
Regulations on storage of waste on other -than power reactor
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sites (materials licensees) depend on the options pursued. To 4
quote the NRC: " Simple expansion of existing storage capacity i within licensed possession limits often would not require a 1
j license amendment." Generally, a license amendment would be l required if a licensee needs to exceed material possession
! limits authorized by the license to provide for additional l
- storage, or if the storage involves new materials handling-techniques or use of new facilities not previously evaluated.
j The most complex materials licensing case would probably
} involve a proposal'to operate a facility for storage of waste
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. o generated- by others as a commercial enterprise. An environmental impact assessment would likely be required for commercial storage.
t 1.4.2.2 Decommissioning Nuclear power plants are decommissioned at the end of their lives using a method meeting the requirements for license termination of 10 CFR 50 and associated documents, such as I Regulatory Guide 1.86. Three -of the principal decommissioning alternatives recognized by the NRC are immediate dismantlement (DECON), entombment (ENTOMB), and mothballing (SAFSTOR).
1.4.2.3 Disposal Disposel of low level waste is governed by 10 CFR 61 and related Regulatory Guides or Branch Technical Positions. Of particular importance to this effort, are the regulations and guidance for wakt? form stability and classification. The concept of storage u.3r long-term (as previously discussed) j will be better addressed if the waste is stored in a manner
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which would allow immediate disposition. In essence, the j- reactor containment will, after decommissioning, become
- e,uivalent to an above-ground vault. The performance objectives r
- j. (61.,6) will be an integral part of the above-ground vault j concept. It is expected that more states and compacts will i forbid traditional shallow land burial, and further guidance I will be issued by NRC for this and other concepts.
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1.4.3 Relevant Decommissioning and Low-Level Waste Studies Several different methods of decommissioning nuclear power plants have been proposed and studied, including immediate dismantlement, entombment and mothballing. The definitions of these options and a discussion of earlier studies (both taken from Ref. 3) follows o Immediate Dismantlement (DECON) - Following decontamination, all radioactive materials, equipment, and structures would be dismantled and removed. Resulting termination of nuclear licenses would allow unrestricted access to all areas of the site, o Entombment (ENTOMB) -
The most radioactive components would be entombed in a massive concrete structure (e.g.,
below the sealed operating deck of a PWR containment),
while the remaining radioactive items would be decontaminated, dismantled and removed. The entombment structure would stand until much of the radioactivity decays away, then it too would be dismantled.
o Mothballing (SAFSTOR) - Radioactive areas of the plant would be placed under protective storage, continual
. security and environmental surveillance, until the plant would eventually be decommissioned by one of the other j acceptable methods. This is par'.icularly attractive for multi unit sites where one unit would be decommissioned first.
I Major decommissioning literature dates back to 1976 when Manion and LaGuardia produced a comprehensive report for the National f
Environmental Studies Project of the Atomic Industrial Forum 4
examining and comparing the different decommissioning options.
Battelle Pacific Northwest Laboratory then produced a series of g
studies for the NRC looking at . decommissioning a range of i . !
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' nuclear facilities using different decommissioning methods. Two of the Battelle reports (which have been supplemented several 5
times cover decommissioning of commercial power BWRs and l PWRs, making detailed cost and dose comparisons for reference 7
plants. A DOE handbook was also prepared discussing different decontamination techniques, which would provide useful information for decommissioning.
A fourth decommissioning option, intact decommissioning, had 8
its genesis in a paper prepared by Lewis of Duke Power for the 1982 Health Physics Society Annual Meeting. Under this option, the most highly radioactive components (e.g., the reactor internals and primary coolant system) would be left intact inside the containment, which would then be sealed (all penetrations cut and filled with concrete, then capped with steel cover plates). While the rest of the plant would be decontaminated and dismantled, the intact structure would be left in place into the indefinite future, without the costly requirements of continued maintenance and surveillance. Several recent Canadian studies9,10,11 examined delaying the dismantling of CANDU reactors for long periods of time after fuel has been removed and radioactive fluids drained. Finally, 3
a new AIF/NESP study by Ostrow, et al. of Ebasco Services, Inc. took a detailed look at the feasibility and dose consequences (occupational and public) of the intact decommissioning option for generic PWRs and BWRs, and found it not only feasible, but also lower in occupational dose exposure than immediate dismantlement.
" Solving" the nation's low-level radioactive waste disposal problem has become a high political priority in the last few years, culminating in the passage by Congress of the 12 Low-Level Radioactive Waste Policy Act in 1980. The act differentiates between low-level waste which would become a state responsibility, and high-level waste which would remain a i federal responsibility. States may form compacts to dispose of 1-8 i
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low-level waste and exclude waste generated outside the compact boundaries.
The NRC issued 10 CFR 61, " Licensing Requirements for Land Disposal of Radioactive Waste", in December 1982, giving technical and licensing standards to low-level wasts disposal.
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? Draft and Final Environmental Impact Statements prepared by the NRC staff supported Part 61 with technical analyses using
', the methodology developed by Oztunali, et al. at Dames and 16 Moore. This methodology has been updated and expanded in 1986 17 by Oztunali of Ebasco Services, Inc. and Rolles of the1,2 NRC.
I Finally, as mentioned earlier, two NRC documents form the primary regulatory guidance on storing nuclear waste on-site at nuclear power plants.
1.5 Report Contents Section 2 describes the features of the Maine Yankee Atomic Power Plant, including operating and physical parameters, as well as location, building layout and ownership. Section 3 characterizes all the radioactive waste considered in this study by type, amount and source. Section 4 defines and describes the features of the concept behind this study; i.e.,
combining low-level waste storage and disposal with Maine Yankee decommissioning, Section 5 looks at the technical feasibility of the concept, and Section 6 at the dose consequences. Finally, Section 7 summarizes the findings of the study and makes recommendations for further investigations.
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. o 1.6 References
- 1. U. S. Nuclear Regulatory Commission Generic Letter 81-38,
" Storage of Low-Level Waste at Power Reactor Sites", 1981.
- 2. U. S. Nuclear Regulatory Commission, " Safety Consideration for Temporary On-Site Storage of Low-Level Radioactive Waste", Draft, June 1980.
I 3. S. Ostrow, et al., " Intact Decommissioning of Nuclear Power Plants: A Dose Assessment", AIF/NESP-034, prepared by Ebasco Services, Inc. for the National Environmental Studies Project of the Atomic Industrial Forum, March 1986.
- 4. W. J. Manion, T. S. LaGuardia, "An Engineering Evaluation of Nuclear Power Reactor Decommissioning Alternatives",
AIF/NESP-009, November 1976.
1 5. H. D. Oak, G. M. Holter, W. E. Kennedy, Jr., G. J. Konzek,
" Technology Safety and Costs of Decommissioning a Reference Boiling Water Reactor Power Station", NUREG/CR-0672, prepared by Battelle Pacific Northwest Laboratory for the USNRC, Vols. 1 & 2, June 1980, Addendum 1, July 1983, Addendum 2, September 1984.
- 6. R. I. Smith, G. J. Konzek, W. E. Kennedy, Jr., " Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station", NUREG/CR-0130, prepared by Battelle Pacific Northwest Laboratory for the USNRC, Vols 1
& 2, June 1978, Addendum 1, August 1978, Addendum 2, July 1983, Addendum 3, September 1984.
- 7. " Decommissioning Handbook", DOE /EV/10128-1, prepared by Nuclear Energy Services, Inc. for the USDOE, November 1980.
- 8. L. Lewis, "To Decommission a Nuclear Power Reactor, 'Just Lock the Door'", report presented at the Health Physics Society Annual Meeting, Las Vegas, Nevada, 1982.
- 9. J. M. Liederman, J. I. Saroudis, "The Inherent Advantages of Delayed Dismantling of Decommissioned Nuclear Power Stations", presented at NEA Workshop on Storage with Surveillance vs. Immediate Decommissioning for Nuclear Reactor Components and Buildings, Paris, France, October 22-24, 1984.
- 10. P. Denault, I. Kuperman, "Gentilly-1 Decontamination for Decommissioning Program", presented at the International Conference on Decommissioning, Bethesda, Maryland, July 16-18, 1985.
C. R. Bennett, "The Extension of the SWS Period of CANDU 11.
Reactors with Particular Reference to Douglas Point", j presented at the International Conference on <
Decommissioning, Bethesda, Maryland, July 16-18, 1985. ,
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- 12. " Low-Level Radioactive Waste Policy Act", Public Law 96-573, 1980.
- 13. U. S. Nuclear Regulatory Commission, Code of Federal Regulations Title 10 Part 61, "J.icensing Requirements for l Land Disposal of Ra.!ioactive Waste", December 17, 1982.
- 14. U. S. Nuclear Regulatory Commission, " Draft Environmental
, Impact Statement on 10 CFR Part 61: Licensing Requirements for Land Disposal of Radioactive Waste", NUREG-0782, September 1981.
- 15. U. S. Nuclear Regulatory Commission, " Final Environmental Impact Statement on 10 CFR Part 61: Licensing Requirements for Land Disposal of Radioactive Waste", NUREG-0945, November 1982..
- 16. O. I. Oztunali, et al., " Data Base for Radioactive Waste Management", NUREG/CR-1759, prepared by Dames & Moore for the USNRC, November 1981.
- 17. O. I. Oztunali, G. W. Roles (USNRC), " Update of Part 61 Impacts Analysis Methodology", NUREG/CR-4370, prepared by Ebasco Services, Inc. for the USNRC, January 1986.
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2.0 MAINE YANKEE ATOMIC POWER PLANT l
The plant contains a 2,630 MW(t) Combustion Engineering pressurized water reactor, and produces a net power output of 850 MW(e). It is owned by the Maine Yankee Atomic Power Company, and went into commercial operation in 1972.
2.1 Ownership and Location The Maine Yankee Atomic Power Plant, contains a single unit pressurized water reactor, and is owned and operated by the Maine Yankee Atomic Power Company which, in turn, is owned by ten member electric utility companies in the New England area.
The largest ownership shara, 38%, is held by the Central Maine Power Company.
l The plant is located on a peninsula known as Bailey Point which extends toward the south to Montsweag Bay, about four miles from Wiscasset, Maine. Wiscasset is a small town of 3,000 people, situated near the coast in the southeast part of the state. The plant site consists of 740 acres of land bounded by Back River estuary on the east, mainland on the north, and Birch Point Road on the west. The waters of Back River, Montsweag Bay, and its tributaries are tidal and open to
. boating, both commercial and recreational. The plant and the surrounding area are shown in the site plan, Figure 2-1.
The exclusion radius for the site is in excess of 2,000 feet, and a low population zone (LPZ) has been established with a radius of six miles. Within five miles of the site, land use is largely home sites, small businesses, summer houses, idle farmland and forest. There is one small dairy within the area, with several other locations having a few milk cows for private use. Housing is scattered along principal roads and is concentrated only in the center of Wiscasset.
2-1
_a -
Grade elevation for the plant, which is located on a ridge of bedrock (all major structures are founded directly on hard, crystalline bedrock), is at El. 20' and the normal groundwater level runs about 8 to 10 feet below that. Pumps with automatic level controls are provided in wells to maintain the groundwater level below the bottom of the containment mat and thus prevent uplift of the structure. Water wells in the area are either dug wells, usually less than 25 feet deep, or drilled wells penetrating the bedrock to depths of 100 feet or more. The wells are for domestic or farm use and seldom exceed 5 or 10 gpm for short-term pumping and even less for sustained pumping. There are no municipal or other important well water supply systems in the area, and all wells within the 2,000-foot exclusion radius are controlled by Maine Yankee.
The maximum predicted water level at the power station (based on simultaneous occurrence of the maximum storm surge and other hydrological factors) is at Elevation 19'-6". There is no significant risk of flood at the site since the shore protection Elevation of 22'-0" should preclude water from entering.
2.2 Major Structures and Equipment The layout and dimensions of the structures on the site are shown on the Plot Plen, Figure 2-2. The major structures are:
Reactor Containment, Primary Auxiliary Building, Fuel Building, Onsite Storage Facility, Turbine Building, Service Building and Staff Building, with only the first four housing radioactive l materials, systems or components. These, and the major pieces of equipment they house, will be- discussed briefly in turn.
1 Table 2-1, taken from the Maine Yankee FSAR, summarizes the design characteristics of the plant.
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'The Reactor Containment (illustrated in Figures 2-3 to 2-7) is a steel-lined, reinforced concrete cylinder with a l hemispherical dome and an essentially flat reinforced concrete foundation mat. It has an outside radius of 72' and a cylinder height of 102' above the inside surface of the foundation mat.
The dome has an outer radius of 70'. The wall thickness of the g cylinder is 4'-6", and of the dome 2'-6". The foundation mat is 10' thick resting on bedrock, with a central reactor vessel pit projecting below its top surface.
q The Reactor Containment houses the three-loop nuclear steam I supply system, including the nuclear reactor, the primary system piping and pumps, the three ' steam generators and the pressurizer. The building is a Class I seismic structure, and is designed to contain, with a very low leakage rate, radioactive material released from the core at elevated temperatures and pressures during a design basis accident.
The 472 ton reactor vessel, shown in Figure 2-8, is a cylinder with hemispherical heads, approximately 42' high (to the top of the control rod drive mechanism nozzles) by 14' inside diameter. The wall thickness is a minimum of 8-5/8", and the carbon steel bace material is clad on the inside with 5/16" of stainless steel. Major components inside the reactor vessel are the core, thermal shield, core support structure, in-core instrumentation, core barrel, and control rod guide tube assemblies. The core consists of 217 fuel assemblies with an active fuel length of 137" and has an equivalent diameter of 136". The slightly enriched UO fuel is clad with Zircaloy-4.
2 The reactor vessel sits in the reactor cavity which is shielded on the sides by a 4'-6" thick concrete primary shield wall. An i annular neutron shield tank surrounds the reactor vessel below the coolant nozzles inside the reactor cavity.
The reactor heats primary coolant water under pressure to prevent boiling. This water is circulated through three primary 2-3 e- , ._ , , - - - - - ,
- n. , - - - - , , , ---
coolant loops, each with one reactor coolant pump and one steam generator in which secondary system water is heated and turned into steam. The steam is then used to produce electricity in a
- turbine-generator system. A pressurizer, which maintains correct primary system pressure, is connected to one of the coolant loops. The vertical "U" _ tube steam generators are approximately 58' high, by 15' diameter for the upper shell and 11' for the lower shell. Each weighs about 330 tons dry.
Other major features of the Reactor Containment include the i large, 38' deep, refueling cavity located north of the reactor, and the 360 ton capacity bridge crane riding on rails located on the top of the containment crane wall. The Reactor Containment has numerous penetrations for fluid, electrical and HVAC systems, as well as a 7' inside diameter personnel hatch and a 22' inside diameter equipment hatch.
The Primary Auxiliary Building (illustrated in Figures 2-9 and
-10) is a rectangular, Class I seismic structure approximately 120'x96' feet located north of the Reactor Containment Building. It contains reactor coolant purification equipment, ventilation equipment for the Primary Auxiliary and Reactor Containment buildings, radwaste processing equipment, reactor coolant boric acid control equipment, and equipment for certain other auxiliary systems. Outside walls are concrete up to Elevation 36'-0", and are steel above that, up to the top of roof Elevation 54'-0".
The Fuel Building (shown in Figures 2-11 and 2-12) is a rectangular Class I seismic building approximately lll'x49' located north of the Reactor Containment and west of the Primary Auxiliary Building. It houses both new and used fuel, and also contains waste disposal equipment. The spent fuel is stored in racks in the water-filled spent fuel pool which is 41'6"x49' and is 36'-6" deep. The spent fuel pool is shielded on all sides by 6'-thick concrete walls. ;
i 2-4 )
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.The Onsite Storage Facility (OSF1- also known as the Low Level Waste and Equipment. Temporary Storage Building) is located on ,
the plant site outside the protected area. It'is 154'x68' by l about 40' high above a floor Elevation of 26'-0". The location is illustrated in Figure 2-13. The. outer walls of the building consists of steel siding, with l'-thick concrete shield walls inside the siding to a height of 16'. A truck bay allows access for the shipping in or out of. waste containers.
l The OSF is intended to provide interim storage for radioactive l waste until a disposal site or facility is available for it to be shipped for final disposition. The radioactive waste will be l packaged in shipping containers that are ready for final-shipment. It will alco store clean or contaminated equipment l~ for reuse.
l 1 The Turbine Building houses the turbine-generator and associated power generation equipment, and the two diesel
~
generators. It is located east of the Reactor Containment-and is 275'x135'. A 125 ton crane. runs overhead along the length of the building. The building is constructed of
+
structural steel framing with a reinforced concrete floor.
Since the secondary water side of a PWR is at the most only-slightly radioactive, the Turbine Building is consid. red to be nonradioactive for the purpose of decommissioning.
The Service Building is located east of the Reactor Containment and contains the main control room, switchgear rooms, shops and employee facilities. It contains no radioactive systems or equipment.
I Other . structures on the plant site includes containment spray pump area adjacent to the Containment on the south side, which contains Emergency Core Cooling System (ECCS) equipment;-
transformer area; circulating water pump house; warehouses; I shops; and, varic as tanks.
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2.3 References
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i I TABLE 2-1 i MAINE YANKEE DESIGN CHARACTERISTICS
' Plant Net Electrical Power Output, MW(e) 9 2,630 MW(t) 850 Gross Electrical Power Output, MW(e) 92,630 MW(t) 864 Maximum Expected Gross Electrical Output, MW(e) 864 I Nuclear Steam Supply System Core thermal Output, MW(t) 2,630 Operating Pressure, psig 2,235 Design Pressure, psig 2,485 g
i Reactor Coolant Inlet Temp. (nomincl op.), F 550 Reactor Coolant Outlet Temp. (nominal op.), F 600 Pipe Size: Outlet - ID, in. 33-1/2
" - Wall Thickness, in. 3-1/4 Inlet - ID, in. 33-1/2
" - Wall Thickness, in. 3-1/4 6
Flow per Loop, lb/hr 44.87x10 Number of Loops 3 Number of Pumps 3 Type Vertical, Centrifugal, Mechanical Seals Design Flow, gpm 120,000 Core Total Heat Output, Btu /hr 8.981xlO Heat Generated in Fuel, t 97.5 DNB Ratio at Nominal Conditions 1.87 Minimum DNBR for Design Transients 1.30 Core Power Density, ki/ liter 80.86 1 Number of Fuel Assemblies 217 l 1
Number of Fuel / Poison Rods per Assembly J i6 Fuel Rod Pitch, in. 0.580 2-7
)
I :
TABLE 2-1 (cont'd)
MAINE YANKEE DESIGN CHARACTERISTICS Fuel Clad Material Zircaloy-4 Fuel Clad Nominal Thickness, in. - CE 0.028
- EXXON 0.031 8;5 Number of Control Rod Locations (maximum)
CEA Pitch, in. 11.57 Poison Materials B C/ stainless steel i
Control Rod Drive Type Magnetic Jack Equivalent Core Diameter, in. 136 Total Uranium, MTU 80-83 Reactor Vessel Inside Diameter, in. 172 Overall Height, Including CEDM Nozzles 42' 3/8" Wall Thickness, Minimum, in. 8-5/8 j
1 Wall Material A-533 Grade B Class 1 Steel Cladding Thickness, in. 5/16 Cladding Material 304 SS Design Temperature, F 650 Design Pressure, psig 2485 Total Weight, tons 472 Steam Generators Number of Units 3 Type Vertical "U". Tube Upper Shell Outside Diameter 15' 1/2" i Lower Shell outside Diameter 11' 3/4" Overall Height 58' - 10" Number of rubes 5703 Tube OD, in. 3/4 Tube Material Ni-Cr-Fe Alloy 2WB o
e s 1
TABLE 2-1 (cont'd)
MAINE YANKEE DESIGN CHARACTERISTICS i
Primary Side:
-t Tube Side Design Pressure, psig 2485 o
. Tube Side Design Temperature, F 650 Tube Side Operating Pressure, psig 2235 o
Coolant Inlet Temperature, F 532-600 o
Coolant Outlet Temperature, F 532-550 Bottom Head Clad Material 304 SS
- Secondary Side:
Shell Side Design Pressure, psig 985 o
Shell Side Design Temperature, F 550 Operating Pressure, Steam Generator Outlet at Plant Rating, psig 860 o
Operating Temperature, F 520.3 Quality, t 99.8 6
Steam Flow / Steam Generator, 10 lb/hr 3.575 n
Turbine Cycle Turbine Design Tandem-Compound, 1 HP, 2LP Turbines i
Exhaust Pressure, Hg absolute 1.5 Steam Atmospheric Dump, % rated steam flow 3 i Steam Bypass to Condenser, 4 normal steam flow to condenser 45 Feedwater Heater Stages 6
! Condensate Pumps - number 3 Half-Capacity Design Flow, gpm 9,060 Design Head, ft 960
- Feedwater Pumps - Electrical - Number 2 Half-Capacity Design Flow, gpm 14,000 Design Head, ft 2,038 Feedwater Pump - Steam Driven - Number 1 Full-Capacity Design Flow, gpm 28,000 Design Head, ft 2,200 f
2H9
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7 TABLE 2-1 (cont'd)
MAINE YANKEE DESIGN CHARACTERISTICS Circulating Water Pumps - Number 4 Quarter-1 Capacity Design Flow, gpm 106,500 Design Head, ft 26 Generator Design Rating, Mva 900' s
Power Factor .90 Terminal Voltage, kV 22 NSSS Auxiliary Systems (a) Chemical and Volume Control System Normal Letdown Flow Rate, gpm 80 Maximum Letdown Flow Rate, gpm 200 Charging Pumps - Number 3 Fixed-Capacity Design Flow, gpm 150 Design Pressure, psig 2,850 Auxiliary Charging Pump - Number i Variable Speed Design Flow, gpm -10 to 30 Design Pressure, psig 3,700 Regenerative Heat Exchanger - Number 1 Full-Capacity 6
Design Heat Transfer, Btu /hr 10.0x10 Letdown Heat Exchanger - Number 1 Full-Capacity 4
6 Design Heat Transfer, Btu /hr 7.82xlO Demineralizers - Number 3 Purification 1 Deborating l Nominal Rating, gpm 80 Maximum Flow, gpm 20 0 3
Resin Volume, ft 32 Filter - Number 2 Type Cartridge Design Rating, gpm 200 Filter Size, microns 2 2-10 1
?'
TABLE 2-1 (cont'd)-
MAINE YANKEE DESIGN CHARACTERISTICS (b) Safety Injection System Safety Injection Tanks - Number 3 3
Volume - Total, ft 3,500 3
Borated Water, ft 1,500 3
2,000 I Nitrogen @ 225 psig, ft Design Pressure, psig 250 Design Temperature, F 200 Low Pressure Pumps - Number 2 Full-Capacity Rating, Each, gpm 3,000 Head,'ft 350 (c) Containment Spray System
- Spray Pumps - Number 3 Rating, Each,gpm 3,700 Head, ft 305 (d) . Refueling Water Tank
! Fluid Volume, gal 375,600 Boron Concentration, ppm 1,720 (e) Auxiliary Feed System Steam Generator Emergency Feed Pumps Motor-Driven - Number 2 Full-Capacity Rating, gpm 500 Head, ft 2,525 Steam Generator Auxiliary Feed Pump Turbine-driven - Number 1 Full-Capacity Rating, gpm 500 l Head, ft 2,525 I
(f) Component Cooling System Component Cooling Pumps - Number 2 Primary 2 Secondary 2-11
l l
TABLE 2-1 (cont'd)
MAINS YANKEE DESIGN CHARACTERISTICS Rating, Each, gpm 6,000 Head, ft 190 Heat Exchangers - Number 2 Primary 2 Secondary 6
Rating, Each, Btu /hr 51.3x10 i
(g) Spent Fuel Cooling System Spent Fuel Pool Capacity 953 assys.
3 Volume, ft 59,116 Pumps - Number 2 Rating, Each, gpm 772 Head, ft 120-Heat Exchanger - Number 1 6
Rating, Btu /hr 22.3x10 Filter - Number 1 i Type Cartridge Rating, gpm 200 Size, microns 2 Demineralizer - Number 1 Resin Type Mixed Bed Bed Size, ft 32 Nominal Flow, gpm 200 l
! Conventional Plant Auxiliary Systems (a) Service Water System Service Water Pumps - Number 4 Full-Capacity Rating, gpm 10,000 4 Head, ft 66 i
ll (b) Compressed Air System l Compressors - Number 3 Rating, scfm 300 Discharge Pressure, psig 100 2-12 t
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.I TABLE 2-1 (cont'd)
}
- , MAINE YANKEE DESIGN CHARACTERISTICS f
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! Containment i
Type Reinforced Concr.
} f Diameter 135'-0" t
Height 169'-6" Liner - Material ASTM A516 Grade 60 Thickness, in. - Wall 3/8 jl , - Dome 1/2 j4 i - Floor 1/4 Design Pressure, psig 55 o
- Design Temperature, F 280
} Leak Rate,'%/ day 0.1 l
1' Electrical Equipment I Main Transformer - Number 2 Capacity, MVa 430 l
i Voltage, kV 345 l Diesel Generators - Number 2 Full-Capacity i Rating, kva 3,560
, Fuel Oil Capacity 1 Week Station Battery - Number 4 Rating, ampere-hours 1,800/480 Chargers - Number 4 i Inverters - Number 4 AC/DC Voltage, volts 120/125 i Rating, kva 10 i
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3.0 RADIOACTIVE WASTE CHARACTERIZATION This section characterizes the LLW generated in the State of Maine which currently has 81 licensees to possess radioactive - 1 material. Only a small number of licensed firms, however, generate LLW requiring disposal. The principal sources 3 (greater than 100 ft /yr) are: Maine Yankee Atomic Power Plant, Portsmouth Naval Shipyard, Mount Desert Biological Laboratory, 3 and Jackson Laboratory. Minor amounts (less than 100 ft /yr) are generated by various other institutions. Because Maine Yankee generates almost 90% of the State's LLW volume, the waste description is divided into Maine Yankee waste and ochers. The Maine Yankee waste is further divided into operational waste (LLW generated during normal plant operation)- and decommissioning waste (LLW generated during decommissioning). 3.1 MAINE YANKEE OPERATIONAL WASTE Maine Yankee, an 850 MW(e) Pressurized Water Reactor (PWR), generates two types of LLW during normal operations of the plant: process waste and trash. A detailed generic description of this waste is provided in Section 3.1.1. The historical waste volumes and projected volumes are presented in Section 3.1.2. Waste forms and containers are discussed in Section 3.1.3. 3.1.1 Waste Types (
, The two types of LLW generated by Maine Yankee are process i waste and trash. Process waste results from treatment and cleaning the various water systems suppcrting the nuclear 3-1 g --a .,w -- -
reactor, and consists of liquid evaporator concentrates, spent ion-exchange resins, and cartridge filters. Trash results from various plant maintenance activities, and consists of dry compressible waste such as plastic sheeting, paper, cloth and plastic gloves, booties and other items of protective clothing, and small pieces of contaminated equipment such as valve and motor parts. The NRC, as part of rulemaking on 10 CFR 61,2 has formulated
- generic descriptions of these PWR waste streams:
o PWR Ion Exchange Resins o PWR Concentrated Liquids o PWR Filter Sludges o PWR Filter Cartridges o PWR Compactible Trash o PCR Non-compactible Trash 3.1.1.1 Ion Exchange Resins Processes involving ion exchange media are frequently used in PWRs to remove dissolvad radioactivity from liquid streams. Ion exchange media usually consist of organic resins, which can be cation or anion resins, or a mixture of both. Inorganic zeolite ion exchange media have also been used in some cases. The resins (or other ion exchange media) are usually packed in beds into cylindrical vessels called ion exchange columns or demineralizers. The liquid containing the specific contaminant is passed through the resin column. In this process, dissolved radiocontaminants chemically displace ions in the resin and become physically bound to the resin. When an ion exchange bed can no longer perform its function (following depletion) it is 3-2 , 1 a
. replaced. The depleted bed material is typically transferred as a slurry out of the column into a storage tank and ultimately to a shipping container (generally referred to as a liner),
where excess water-is removed prior to transfer to a disposal facility. ' Removal of free water is termed dewatering; however, dewatered ion exchange media, can still contain between 42 and 55% water by weight, in addition to interstitial liquid. In general, the liners are transported in casks that are shielded for radiation protection. 4 3.1.1.2 Concentrated Liquids Concentrated liquid waste may be produced by the evaporation of a wide variety of PWR liquid streams. The waste consists of i liquids with an elevated, suspended and dissolved solids content, as well as sludge resulting from supersaturation during evaporation. PWRs recycle and concentrate boric acid waste solutions from the reactor primary system. Other solutions with low boric acid concentration (e.g., cleaning liquids, floor and laboratory drains) are concentrated to about 25%~ solid by weight. These concentratad liquids are currently solidified in concrete within a liner prior to transfer to a disposal facility. 4 i 3.1.1.3 Filter sludge Filter sludge is waste produced by precoat filters and consists of filter aid and waste solids retained by the filter aid. Diatomaceous earth, powdered mixtures of cation and anion exchange resins, and high purity cellulose fibers are common filter aids. These materials are slurried and deposited (precoated) as a thin cake on the initial filter medium (wire
~
~ mesh, cloth, etc.). The filter cake removes suspended solids from liquid streams. Precoat filtration may be used in f 3-3 a
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' conjunction with ion exchange columns and evaporation, or-it may be the only form of treatment removing suspended solids from a particular liquid stream. It'should be noted that Maine Yankee currently does not use precoat filters.
< \
).
b i 3.1.1.4 Cartridge Filters Cartridge filters contain one or more disposable filter ^ elements. These elements may be typically constructed of woven fabric, wound fabric, or pleated paper supported internally by i a stainless steel mesh, as well as pleated or matted paper supported by an external stainless steel basket. Paper filter elements are often impregnated with epoxy. Woven fabric , filters are typically constructed of cotton and nylon. Cartridge filters are effective in removing suspended solids, but do not have the ion exchange capability of precoat filters t or demineralizers. Cartridge filters are used in conjunction with ion exchange columns, evaporators, and precoat filters. These cartridge filters are packed in either 55-gallon drums or
; liners prior to transfer to a disposal facility. The waste can 4
be immobilized by filling the void space of the disposal containers with concrete. t 3.1.1.5 Trash , l Trash waste streams can be divided into compactible and noncompactible trash. Trash is the most varied waste stream l generated and can contain everything from paper towels-to ! t 4 irradiated equipment parts. Typical trash materials are listed ! in Table 3-1. Compactible and noncompactible items are frequently shipped in the same container. In addition, packaging small' pieces of i i I 3-4 l l 3- ,
activated metal with relatively innocuous materials is common. Such factors make a generalized description of trash difficult. In general, compactible trash contains more. combustible material (e.g., paper, plastic), and noncompactible trash contains more metallic components (e.g., pipes and failed equipment). It is usually assumed that .the volume percentage of compactible trash and combustible trash are the same. Similarly, the volume percentages of noncompactible trash and noncombustible trash ^are assumed to be the same. Maine Yankee currently uses a hydraulic box compactor to reduce the final volume of trash wastes. Each metal box is approximately 4' high by 4' wide and 7-1/2' long, and ranges in weight from 2,500 to 3,500 lb. l l, The Maine Yankee primary coolant purification system has three ion exchange demineralizers (in parallel) and a particulate filter. A separate water purification system (filter and f l demineralizer) is designed to clean the spent fuel pool water. The plant specificatiens for these two systems are presented on Table 3-2. Due to extensive volume reduction efforts- at Maine Yankee, the isotopic concentration of its waste is higher than the generic 3 concentrations which assume no volume reduction. A survey of l the principal radionuclides in the Maine- Yankee waste is presented in Table 3-3 (June 1986). It should be recognized that the waste quantity and composition is a function of-the specific plant operations performed at Maine Yankee. Since the 4 operations vary in nature and duration from year to year, the waste quantity and activity will vary accordingly. i l 1 I I 1 i
]
j > 3.1.2 Waste Volumes and Classifications The historical and projected annual LLW volume generatian rate at Maine Yankee is presented in Table 3-4. Historically, the 3-5 Il
.----e - , - - . , . , . . , , , - , - - , , , . ~. - - - - - ,-- ---
Maine Yankee facility has been producing significantly less LLW 3 volume than the generic PWR generation rate of 18,250 ft /yr 3 for a 850 MW(e) salt water cooled PWR. Only once (1978) did Maine Yankee produce more than the generic average volume. As shown in Table 3-4, Maine Yankee is projecting a much lower waste generation rate in the future. The Low-Level Waste Policy Amendments Act of 1985 imposes an annual specificallg average volume disposal limit of 10,452 ft /yr (1986-1989) and 3 8,220 ft /yr (1990-1992) for Maine Yankee. Maine Yankee currently meets and plans to continue to meet-the Congressional limits with the implementation of an ambitious waste volume reduction program. The goal is to reduce the average annual waste volume to no greater than 6,000 ft by 1990. Present j indications are that the goal will be attained. Based on this volume reduction program it is expected that Maine Yankee will generate 5,000 ft /yr of LLW during the proposed storage period, from 1993-2008. Note that total activity of the waste remains unchanged at 300 Ci/yr. As part of the rulemaking of Part 61, NRC developed a waste classification based on relative radionuclide activity: Class A corresponds to low activity, Class B corresponds to medium activity, and Class C corresponds to high activity. The LLRW Policy Amendments Act of 1985 specifically states that disposal of " greater than Class C waste" is the responsibility of the Federal Government, i.e., DOE. i l This LLW classification came into being in 1982 with the implementation of 10 CFR 61. Maine Yankee disposal statistics l (1982-1985) indicate that most of its LLW is Class A waste. Average values (for 1982-85) are 98.6% Class A, 1.0% Class B, and 0.4% Class C, as shown in Table 3-5. Under Part 61, Class B and C wastes are required to be stabilized. This Maine Yankee waste distribution is typical of other PWRs. For this 3+6
1
. . l I
l reason, the waste classification distribution of class A, B, C waste is assumed to remain the same (i.e., 98.6%, 1.0%, 0.4%, respectively) in projecting future Maine Yankee requirements. 3.1.3 Waste Forms and Containers Under NRC Part 61 waste form regulations, all waste must be in l solid form, i.e., no free liquids are allowed. In addition, Class B and C waste forms must be stabilized, i.e., have structural integrity. This requirement may be satisfied with a waste stabilization agent (e.g., cement, polymer, bitumen), or placement in High Integrity Containers (HICs with 300 year life), or engineered structures. Only the small amount, about < l.4%, of the total Maine Yankee waste which is greater than Class A is required to satisfy the waste stabilization requirement, and is currently either being solidified with cement or placed in HICs. , All waste packages must satisfy U.S. Department of Transportation (DOT) regulations for radioactiv'e material transportation. Typically, most of the process waste is placed , in large disposable liners which are put within shielded casks for transportation. Steel boxes are the standard containers for plant trash. Shielding is not normally required on these boxes, and they meet DOT requirements for transport. Statistics for the last six years (1980-85) show 57% of the waste shipments required shielding as shown in Table 3-6. Note that an unshielded shipment contains about twice the waste volume of a shielded shipment. l The waste form and standard container for each Maine Yankee l waste stream is as follows: o Spent Resins - Spent ion exchange resins are dewatered and placed in large disposal liners (170 ft 3). Recent 3-7
4* . g ' -'
. disposal data, shown in Table 3-7, indicate one or two +
liners of spent resins are disposed each year. o Evaporator Bottoms - Evaporator bottoms are mixed with
- f. cement to form concrete within large disposal liners (170 1
r ft 3). As shown in Table 3-8, the shift from steel drums e : (7.5 ft 3) to large liners (170 ft3) has resulted in a J ! marked decrease in container handling. i o Filter Cartridges -- Filter cartridges also are disposed 1 now in liners to reduce handling. Historical data for this 7. waste stream, shown in Table 3-9,-indicate a very small l j disposal volume. i o Contaminated Trash - Contaminated trash is compacted i within large steel boxes (100 ft 3). The historical data for compressible trash (Table 3-10) and non-compressible trash (Table 3-11) show that most trash by volume at Maine Yankee is the dry compressible type, and hardly any is the non-compressible type. 1 Comparison of trash volumes (Table '3-10) to the total LLW disposal volumes (Table 3-4) indicates about three-quarters ot I the total disposed volume is-trash and one-quarter is process waste. h 3.2 OTHER WASTE GENERATORS i ' Other than Maine Yankee, the principal LLW generators in Maine
' the Portsmouth Naval Shipyard, Mount Desert Biological are Laboratory, Jackson Laboratory, Atlantic Antibodies, University of Maine, and a few other institutions. Table 3-12 presents a summary of the historical LLW generation rate of these 5
institutions and their near term projections. - 3-8 i {
The second largest LLW generator in Maine is the Portsmouth Naval Shipyard. Its LLW is generated from the overhaul of nuclear submarines, and consists mainly of trash and ion j exchange resins. It is assumed that Portsmouth is a U.S. Department of Defense (DOD) facility. The LLRW Policy Amendments Act of 1985 clearly defines the State and federal disposal responsibility for LLW. This Act specifies the State , 1 - is responsible for all Navy LLW not exceeding Class C limits and not related to the decommissioning of nuclear submarines. l Therefore, this study assumes the Portsmouth waste is the responsibility of the State. (A review of site disposal records indicates that Portsmouth is currently shipping low-level radioactive waste to the Maxey Flats, Kentucky I disposal site.) ! I j The balance of the State's LLW ic generated by hospitals, l research institutions, and the University of Maine. This category of LLW is usually called " institutional" waste. i Typical LLW generated by medical facilities consists of ! disposable items such as syringes, vials, test tubes, absorbent materials, and gloves used during clinical and diagnostic procedures. In such procedures, specific pharmaceutical drugs are tagged with short-lived radionuclide tracers. Educational facilities generate two basic types of LLW: medical waste and research waste. Medical waste is.similar to that generated by medical facilities, while research wastes include wastes generated as the result of various research activities in such fields as biology, chemistry, and physics. An example of research waste is animal carcasses, which had radioisotope tracers introduced, involved in the study of behavior or i kinetics of biochemical and biological systers. l In general, the LLW generated by institutions includes disposable items, laboratory ware and equipment, spent liquid 3-9 4 4
.--m-,-w , . - - - . ,
. ,,q , ._ -.m . .e --
, - . , ,- ,-,.-.-%. - - , , , , , - . - - -- - - ,~ , _ , , , , - -
scintillation vials and fluids. A characterization of LLW generated by these institutions in 1982 is presented in Table 3-13. Typically, institutions do not have sufficient waste volume to ship directly to the waste disposal facility. Usually, a waste broker will pick up the institutional waste and store it until a full truck load is ready for transport. Steel drums are the typical wasta container for institutional waste. I Recent trends show an increase in LLW generation by institutions. Commercial site disposal records for 1985 3 indicate institutional waste from Maine were 383 ft , which is an increase from previous practice and previous estimates. Typical waste generation models usually assume a growth rate in the institutional waste by a few percent per year. However, the rapidly rising cost of LLW disposal is increasing economic incentives to reduce the LLW generation. In consideration of these factors, this study assumes the projected future LLW generation rate of all other Maine generation 3 to be a constant 3 1,400 ft /yr, based on 1,000 ft /yr for Portsmouth and 400 3 ft /yr for all institutions. A review of the characterization of the institutional waste in Table 3-13 indicates that the average waste concentrations are I within the the limits for Class A waste. Based upon typical institutional waste practices, it is expected that all the institutional waste generated in Maine will be Class A waste. It is assumed this waste will be stored / disposed in 55-gallon steel drums (DOT Spec. 17H). With special relevance to institutional waste, it should be noted that Part 61 requires I that the waste form must not be capable of generating toxic gases; must not be pyrophoric; and, if containing hazardous, biological, pathogenic, or infectious material must be treated to reduce the potential hazard. 3-10
3.3 TOTAL NON-DECOMMISSIONING WASTE
SUMMARY
The total stored LLW inventory is presented in this section. . Under its current operating license, Maine Yankee is expected to operate until October 21, 2008. The interim storage and disposal scenario assumes storage begins in 1993 as required by the LLRW Policy Amendments Act of 1985. Therefore, 16 years (between 1993 and 2008) of operational Maine Yankee waste must be stored. Under the assumptions discussed in Section 3.1, this Maine Yankee waste has the following characteristics: Waste Volume (ft ) 80,000 Class A 98.6% Class B 1.0% Class C O.4% Greater than Class C None Waste Containers Steel boxes (100 ft3) Steel drums (7.5 ft3) Disposable liners (170 ft3) Assuming the containment will be ready to receive LLW for disposal in 2012, the volume of stored LLW from other waste generators will represent a 20-year in'.entory (between 1993 and 2012). Under the assumptions disenased in Section 3.2, this institutional waste will have the '.ollowing characteristics: Waste Volume 28,000 1 Class A 100% j waste Container Steel drums (7.5 ft ) The total stored LLW inventory will have the following characteristics: Total Waste Volume (ft3) 108,000 Class A Volume (ft3) 106,880 l 3-11 ) I
l I-800 Class B Volume (ft3)
-Class C Volume (ft ) 320 Greater than Class C 0 Waste Containers Steel boxes (100 ft3)
Steel drums (7.5 ft3) Disposable liners (170 ft ) I
; 3.4 MAINE YANKEE DECOMMISSIONING WASTE i
At the end of its useful life, the Mai~ne Yankee Atomic Power Plant will be decommissioned, all nuclear-fuel removed, and its _ operating license terminated. Although there are many possible scerarios for decommissioning, this study assumes that all buildings and equipment outside the Reactor Containment Building boundary will be immediately dismantled. It is further assumed that this active phase of decommissioning takes three years from the time of reactor shutdown. Even though all the nuclear fuel has been removed from the site, many of the remaining structures and equipment will be contaminated with recidual radioactivity. This contaminated i i equipment must be either decontaminated or disposed of as radioactive waste. Radioactive equipment inside the containment l that cannot be decontaminated easily would be left in place. I Structures outside the containment would be decontaminated prior to dismantling, and the radioactive waste resulting from ! these activities would be placed inside the containment. I This section will define both the types and quantities of
.adioactive waste that would be produced during the decommissioning of the Maine Yankee Atomic Power Plant.
i i l ! 3-12 - e .
. - _ . . --y., ..-- r.- . -- . , - - . . . , , . . _ - - , , , .-- _ . . . - , - , . - _ , . . , - , , . - . . , , . . . . _ ,_.-,..-...#- . , . .-- -
i I I I 3.4.1 Types of Decommissioning Waste During decommissioning, the residual radioactivity inside the 1 plant must be processed and handled with the resulting radioactive waste properly stored within the in-situ (encapsulation) boundary. The following types of residual radioactivity would be present: (1) Activated Stainless Steel - The core- support structures inside the reactor vessel are mostly made of type 304 stainless steel and are referred to as the " vessel internals". These vessel internals, along with the stainless steel clad on the vessel inner surface, are made radioactive as a result of the neutron bombardment during l plant operation. These items represent by far the greatest 6 quantity of residual radioactivity in the plant (over 10 curies at shutdown). Since the radioactivity is distributed throughout the stainless steel, the internals cannot be decontaminated, but, on the other hand, the activity is integrally bound within a highly corrosive-resistant medium. 4 (2) Activated Carbon Steel - Except for a thin stainless steel i cladding, the 8-5/8" thick reactor pressure vessel is made of SA 533 carbon steel. The elements of the carbon steel are activated during operation by neutrons that have not ) been stopped by the surrounding water or internals. The specific activity of the carbon steel reactor vessel is considerably lower than the stainless steel internals, but like the internals, it cannot be decontaminated since the radioactivity is bound within the steel. t l (3) Activated Concrete - The reactor vessel is surrounded by a concrete shield wall over six feet thick. Neutrons ] , activate the elements of the concrete and the steel rebar, producing radioactive elements distributed vertically I 3-13 u
-m ,m ,&y.--. .,w,-. +me-- y --.%--- y ---a.-
y e - - . w , . . y w .p ,
4 along approximately 15 feet of the reactor cavity, and to + a depth of about 16 inches. This source has a low specific r activity and-is immobilized within the concrete and rebar matrix. It cannot be decontaminated, but the activated portion of the concrete could be physically removed from i the uncontaminated concrete. (4) Contaminated Equipment - During operation, the reactor coolant water contains activated corrosion products and ) i fission products which have leaked from the fuel. These radionuclides travel throughout the radioactive liquid systems of the plant, and a portion adheres to the inside surfaces of pipe and equipment. Typical contaminated
' I .ystems include the reactor coolant system, chemical and volume control system, radwaste 'traatsent systems, and fuel pool cooling and cleaning system.
4
- Even secondary side systems, such as steam generator blowdown, may be slightly contaminated due to steam l generator leaks. The contamination may be reduced or removed by various chemical and mechanical means at the
- time of decommissioning. The effectiveness of decontamination will depend on many ' factors such as the l
amount of contamination, system temperature, type of material, surface characteristics and hardness of the I
- oxide scale. Decontamination may be attempted either to
! reduce radiation fields for ease of access in dismantling, or to reduce radiation sufficiently to permit reclaiming 7 the material as scrap. At the time of decommissioning,- 1 decisions will be made on individual items, but suffice to i say that only a portion of the contaminated equipment will be decontaminated to a degree sufficient for release, and
- the remainder will be treated as radioactive waste.
! (5) Contaminated Structures - Leakage of radioactive liquids and airborne contamination during plant operation will i 3-14 _ , _ . . - . , , . _ ~ . _ _ _ _ _ _ . . , _ . . _ _ _ ,. - _ _ . _ . , , . . _ , , , _ _ - _ _ . . , . . . _ . . . , , . _ , , , _ - _
- o. . .
l i result-in the surface contamination of many concrete and metal surfaces. Although this activity is very low, it is difficult to decontaminate due-to the porous nature of concrete and other surfaces. The activity can usually be removed from concrete only by scraping away the top several inches. This results in low . activity, but. high volumes of bulk waste. Decontamination of various building surfaces will be required during decommissioning to prevent- even larger amounts of structural material from being handled as radioactive waste. i ! (6) Solidified Liquid Waste - Radioactive liquids are encountered in decommissioning as a result of the over 400,000 gallons of reactor coolant and borated water l storage, in addition to any chemical decontamination or rinse solution used. These liquids are cleaned-up prior to
! discharge, but the treatment process results in radioactive evaporator bottoms, ion exchange resins,
! and filter cartridges. These wastes would be solidified in containers with cement or other appropriate material, and stored in the in-situ (encapsulated) containment boundary. i I (7) Dry Waste - Decontamination and dismantling activities ! will produce contaminated clothing, trash and other dry waste. This waste will be compacted in drums or boxes and 4 stored in the in-situ containment boundary. 1 3.4.2 In-Situ Decommissioning Waste i j l There are three recognized methods of decommissioning a nuclear power plants immediate dismantlement, delayed dismantlement, + and entombment. Encapsulation is another concept in which all
; buildings outside the encapsulation boundary (i.e., the
{ Containment) would be dismantled and all radioactive materials stored inside the encapsulation boundary. d 3-15 1 i
- - , , . .y,. . .. , . . . - _ , - , , -_-,,._-..m,. ,. -. , . , . , ,.-,.-r
1 I ,During decommissioning, all nuclear fuel ils removed from the site and.neither nuclear fuel nor high level waste would be stored at the site. Equipment. inside the Containment that is non-radioactive or that can be easily decontaminated would be removed and sold for salvage value. Other equipment inside the containment would be either left in-place or dismantled and stored in the containment. Buildings outside the Containment.would be decontaminated and i dismantled. All equipment and structural material that could not be decontaminated would be cut up, packaged and placed inside the Containment. All radioactive wastes produced during the decommissioning period would be also packaged and placed in the Containment. This period'is assumed to take three years from the time of reactor shutdown. 4 The quantities of radioactive waste produced during decommissioning is estimated using the genericsourcesand
~
nethodologies found in NUREG/CR-0130. This report, j commissioned by the NRC, estimated that complete dismantlement of a large size (3,500 MW(t)) plant would produce over 567,000 3 i ft of radioactive waste containing over 4 million curies. The data in this report was adjusted to take into account specific features of the Maine Yankee Plant such as'its smaller size i (2,630 MW(t)). In addition, a decontamination program to segregate and reclaim non-radioactive material was assumed. Table 3-14 indicates that the immediate dismantlement of the Maine Yankee Atomic Power Plant would produce approximately 3 330,000 ft of radioactive waste. However, one third of this waste is already inside the containment and would remain there j
, in in-situ decommissioning. Table 3-15 shows the isotopic I distribution of these wastes. The waste characterization, in accordance with 10 CFR 61, includes about 96.4% Class A, 2.5%
i Class B, 0.1% Class C, and 14 greater than Class C. l l
\
l i 3-16 j 4
. o 10 CFR 61 states that wastes which exceeds the Class C limit are not generally acceptable for near surface disposal.
The waste in question is a portion of the highly activated stainless steel internals including the lower core barrel, thermal shield and core shroud. These components exceed Class C limits because of the high concentrations of Ni-59, Ni-63 and Nb-94 radionuclides, and would have to be removed and shipped to a federal site suitable for such disposal. They represent a very small volume, but over 99% of the residual radioactivity. A portion of the activated core grid plate and support columns l is classified as Class C because of the concentration of Ni-63. This waste must meet additional requirements to ensure stability and preclude exposure from inadvertent intrusion. Parts of the core barrel and other miscellaneous internals
- would be classified as Class B due to the concentration of Ni-63. Twenty percent of the solidified liquid waste containing spent resins and filter cartridges, as well as 30% of the dry waste is classified as Class B due to Cs-137 concentrations.
Class B waste must meet more rigorous requirements in waste form to ensure stability. Table 3-14 indicates that, after removal of wastes greater than Class C, the residual radioactivity remaining in the containment would be 17,870 curies. Added to this is the 3 224,469 ft of radioactive waste from outside the containment, comprising 11,724 curies of radioactivity. 3-17
3.5 REFERENCES
- 1. Conference- of Radiation Control Program Directors, Inc.,
"The 1984 State by State Assessment of Low-Level Radioactive Waste Shipped to Commercial Disposal Sites", DOE /LLW-50T.
Prepared for U.S. D.O.E., Frankfort Kentucky, page A-43, December 1985.
- 2. O. I. Oztunali, G. W. Roles, " Update of Part 61 Analysis Methodology", NUREG/CR-4370, Vols. 1&2, Prepared for the U.S. Nuclear Regulatory Commission by the Envirosphere Corp., January 1986.
- 3. Oztunali, Roles, page A-45.
- 4. Oztunali, Roles, Table A-3.
- 5. Maine Department of Environmental Protection, "The Siting Design and Cost of Shallow Land Burial Facilities in Northern New England", Volume 1, Augusta, Maine, page 83, May 1985.
- 6. LLW Siting Commission, " Low-Level Radioactivate Waste Disposal Options for Maine", Augusta, Maine, page 16, February 1984.
- 7. R. I. Smith, G. J. Konzek, W. E. Kennedy, Jr., " Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station", NUREG/CR-0130, prepared by Battelle Pacific Northwest Laboratory for the U.S. Nuclear Regulatory Commission, June 1978.
i i i 3-18 I l l
TABLE 3-1("I' TYPICAL PWR TRASH Compactible, Noncompact., Material Combustible Noncombust. Anti-contaminant clothing X Cloth (rags, mops, gloves) X X Conduit X Contaminated dirt X Contaminated tools and equipment Hand tools X X Eddy current equipment X Vessel inspection equipment X Ladders X Lighting fixtures X Spent fuel racks X Seaffolding X Laboratory equipment X X Filters Filter Cartridges X X HEPA filters X X Respirator cartridges X Glass X Irradiated metals X Flux wires X Flow channels X Fuel channels X In-core instrumentation X Shim rods X High density concrete block X-Miscellaneous metal X Aerosol cans X Buckets X Crushed 55-gal drums X
] X l Fitting Pipes and valves X Miscellaneous wood X X Paper X Plastic Bags, gloves, shoe covers X Sample bottles X Rubber X Sweeping compounds X (a) Source: Reference 4 (modified). f 3-19
o Q \ l TABLE 3-2I ") SPECIFICATION FOR WATER PURIFICATION SYSTEMS f I. Primary Coolant System Demineralizers - Number 3 Purification i 1 Deborating j Nominal Rating (gpm) 80 200 Maximum Flow (gpg)) Resin Volume (ft 32 Filter - Number 2
- Type Cartridge Design Rating (gpm) 200 Filter Size (microns) 2 II. Spent Fuel Cooling System Demineralizer - Number 1 Resin Type xed bed Resin Volume (ft3) 32 Maximum Flow (gpm) 200 Filter - Number 1 Type Cartridge Design Rating (gpm) 200 Filter Size (microns) 2 i
(a) Source: Maine Yankee FSAR 6 3-20 e - -w - --er ,.1-.. - - - -.-%-
.. o 1
- TABLE 3-3I ")
MAINE YANKEE WASTE ISOTOPIC ACTIVITY Activity (Ci/m ) Major Half-Life Process Radionuclides (Years) Waste Trash Co-58 0.19 0.142 0.0063 Co-60 5.2 0.284 0.00945 Cs-137 33.0 0.213 0.00945 Fe-55 2.6 0.213 0.0189 Ni-63 120.0 0.568 0.0189 Total 1.42 0.063 (a) Based on waste disposed January-June 1986, estimated accuracy 10 percent. i i b 3-21 i
. # 3
TABLE 3-4 MAINE YANKEE LOW-LEVEL RADIOACTIVE WASTE GENERATION Volume Activity Year Cu, Pt. Curies 1973 2,364 3 1974 5,625 531 1975 8,164 1,254 1976 6,344 504 1977 6,373 25,730 1978 19,874 .4,136 1979 12,814 2,772 1980 16,111 4,797 1981 14,642 1,666 1982 7,785 30 1983 11,921 103 1984 12,335 359 1985 12,670 111 1986 5,630 62 8,200I ") ID) 1987 300(b) 1988 7,600I ") 1989 4,700(a) 300(b) 300 1990 - 2008 95,000(b),(c) 5,700(b),(c) Total 258,152 48,608 a Notes: (a) Volume projected from State Radioactive Waste Annual Report (b) Estimated 3 (c) Average 5,000 ft /yr, 300 Ci/yr r e 3-22 I
TABLE 3-5 MAINE YANKEE WASTE CLASSIFICATION i i. Class A Class B Class C Voluge 4 of Voluge 4 of Voluge 4 of Year ft Total ft Total ft Total 1982 7,785 100.0 0 0 0 0 1983 11,922 100.0 0 0 0 0 1984 11,996 97.2 340 2.8 0 0 1985 12,380 97.7 120 1.0 170 1.3 Total 44,082 98.6 460 1.0 170 0.4 f f r 3-23 i i
. _ _ . , _ . . . . , . . , . - . _- -.m .
. . I l
TABLE 3-6 MAINE YANKEE SHIELDED VS UNSHIELDED SHIPMENTS Shielded Unshielded Year Total Shipments Number % Number % 1980 51 46 90.2 5 9.8 1981 32 22 68.8 10 31.2 1982 17 5 29.4 12 70.6 1983 21 5 23.8 16 76.2 1984 21 6 28.6 15 71.4 1985 22 10 45.5 12 54.5 i I Total 164 94 57.3 70 42.7
)
i 1 i N l J l 3-24
I v TABLE 3-7 MAINE YANKEE SPENT RESINS Voluge Activity Year Type Number ft Ci 1973 None O C 0 1974 Liner, Drum 1, 5 143 514 1975 Liner, Drum 2, 18 241 1,117 i 1976 Liner, Drum 3, 5 224 462 1977 Liner 1 70 7 1978 Liner 3 221 39 1979 Liner 1 85 520 1980 Liner 7 1,183 4,511
. 1981 Liner 2 252 1,222 1982 None 0 0 0
' 1983 Liner 1 121 86 1984 Liner 2 340 336 1985 Liner 1 170' 73 Total Liner, Drum 24, 28 3,050 8,886 I
t j l 3-25 i l
l TABLE 3-8 MAINE YANKEE EVAPORATOR BOTTOMS e Voluge Activity Year Type Number ft Ci 1973 Drum 265 1,946 2.7 1974 Drum 458 3,374 10.8 1975 Drum 670 4,924 125.7 1976 Drum 393 2,905 38.0 1977 Liner, Drum 10, 105 2,397 18.7 1978 Liner 36 6,001 54.9 1979 Liner 19 3,230 11.9 1980 Liner 31 8,310 134.7 1981 Liner, Drum 14, 61 4,828 17.6 1982 Liner, Drum 4, 244 2,522 19.0 1983 Liner, Drum 21, 3 4,777 5.6 1984 Liner 20 3,400 8.4 1985 Liner 16 2,720 11.0 Total Liner, Drum 171, 2,199 51,333 459.0 l 3-26 I
. O e
TABLE 3-9 MAINE YANKEE FILTER CARTRIDGES, SLUDGES Voluge Activity Year Type Number ft Ci 1973 1973 to 1978 filter 1974 cartridges were solidified 1975 with bottoms and shipped 1976 in 55 gallon drums 1977 ........ 1978 Drum 220 1,687 40.3 1979 Drum 22 253 9.2 1980 Drum 98 739 136.0 1981 Drum, 19.5ft 3 STC 6, 3 104 410.2 i 1982 None 0 0 0 1983 Liner 1 136 0.1 1984 None 0 0 0 1985 None 0 0 0 Total 2,919 595.8 i l I 4 3-27 l i
l 0 e l l l TABLE 3-10 , MAINE YANKEE i DRY COMPRESSIBLE WASTE (CONTAMINATED TRASH) Voluge Activity Year Type Number ft Ci 1973 Drum 57 418 0.6 1974 Drum 295 2,108 5.9 1975 Drum 407 2,999 10.9 l
! 1976 Drum, Box 295, 8 3,215 4.3 1977 Drum, Plywood Box 278, 19 3,874 4.7 1978 Plywood Box 93 11,950 2.2 4
1979 Plywood Box 72 9,216 1.3 1980 Compacted Ply. Box 50 5,878 0.9 1981 Comp. Steel Box, Drum 78, 106 9,459 16.0 i 1982 Comp. Steel Box, Drum 46, 41 5,264 11.0 1983 Comp. Steel Box, Drum 62, 24 6,888 12.1 1984 Com Stl Bx, Drum, Liner 67, 29, 3 8,596 14.8 1985 Com Stl Bx, Drum, Liner 84, 56, 1 9,780 26.6 Total Drum, Box, Liner 1,588, 606, 4 79,644 111.2 l 3-28 6
-- -r - -m - - --.m- -,
TABLE 3-11 MAINE YANKEE IRRADIATED COMPONENTS Voluge Activity Year Type ft Ci 1973 None 0 0 1974 None 0 0 1975 None 0 0 1976 None 0 0 1977 Poison Pins 32 25,700 1978 In Core Detectors 15 4,000 1979 Shim Rods in PB-1 30 2,230 1980 Surveillance Capsules 1 14 1981 None 0 0 j 1982 None 0 0 i 1983 None 0 0 l i 1984 None 0 0 i 1985 None 0 0 j Total 78 31,944 l 3-29 1 _,j
.. .. s 4
TMEF. 3-12I *I* IDI MhDE'S IDi IAVEL imSTE i 1978 - 1986 Source 1978 1979 1980 1981 1982 1983 1984 1985 1986 4 Maine Yankee Cubic Feet 19874 12814 16215 14642.5 7785 11921.5 12065.5 12,670 i Curies 4136 2772 4805 1665.8 30.9 103.5 359.5 111 5630((c) 62 c) Portsmouth Naval Shipyard I~ Cubic Feet 6000 3000 2000 3000 1000 1000 2000 1000 1000 1 Curies 2 3 3 9 <1 <1 ' 4 1 1 1
- Mount Desert Biological Laboratory Cubic Feet 135 180 169.5 144 75 100 100 l Curies .0086 .0221 .1262 .0096. .011 Jackson Imboratory f
Cubic Feet 30 60 128 45 150 1 105 100 100 Curies . 008 . 024 .057 .028 .022 .500 i ! Atlantic Antibodies i Ctaic Feet 0 0 0 0 22.5 0 0 45 45 Curies 0 0 0 0 .0839 0 1 Foundation for Blocd Paaaarch Cubic Feet 22.5 22.5 0 0 0 0 7 7 , ! Curies .0025 . 0031 0 0 0 0 l University of Maine, orono i Cubic Feet 0 0 0 0 75 0 l l Ventrex Laboratories Cubic Feet 0 0 0 0 0 0 0 l 4 Maine 'Ibtal i Cubic Feet 25904 16099 18500 17867 ~ 9127 13065 14245 10327 6882 i Ibtes: (a) Source: Reference 5 (inconsistent with Table 3-4 since from different sources).
! (b) 1978-1984 data indicates waste shipped for burial. 1985-1986 data are estimates of the annual waste vol m e required to be disposed at a Iow Invel Waste Facility. A blank- indicates information that was not
! readily available. (c) Actual, from Table 3-4. 1, ! 3-30 q .
--w,-w.4, w t-#,-- - - -y -
yi- -.r-r-------er--,--m-,,v.-,-w .,...-c m w, . , , , --
-..----r_.. - . , ,,c -.s.----r-.--,.r- ,,-~ee~-,+r-- .,-w. . - - .
I l l TABLE 3-13 CHARACTERIZATION OF LOW-LEVEL RADIOACTIVE WASTES (") Waste Total Volume Activity Generator Disposed (ft3) Waste Form /Vol (ft3) Isotope (Curies) 1,000 Trash and Ion Exch Co-60 1.00 Portsmougg) Naval Shipyard Resins /lOOO Mount Desert 170 Animal Carcasses /35 C-14 0.0045 Laboratory Absorbed Liquids /40 H-3 0.029 Lab or Biological Waste /140 I-125 0.0001 Scintillation Liq /50 Na-22 S-35 Cl-36 0.0045 d I-131 Jackson 141 Lab or Biological Laboratory Waste /140 C-14 0.0045 Sealed Source /l H-3 0.105 i I-125 0.010 Atlantic 22 Compact Trash /15 I-125 0.074 Antibodies Absorbed Liquids /7 Notes: (a) Based on results of the 1983 Low-Level Radioactive Waste ! Management Survey by the Maine Department of Human Services; 1982 Data. Source: Reference 6. (b) Five-year annual average provided by Portsmouth Naval Shipyard. l 1 1 3-31 i
1 TABLE 3-14 MAINE YANKEE DECOMMISSIONING WASTE Location Volume ft (a) Activity, Ci(b) A. In Containment Activated Stainless Steel
> Class C 3,804 1,560,000 Class C 494 7,876 Class B 4,131 1,782 Class A 1,116 68 Activated Carbon Steel IC} 6,744 6,843 Activated Concrete 20,233 388 Contaminated Equipment 36,014 833 Contam. Concrete / Metal 36,910 80 Total in Containment 109,446 1,577,870 Excl. > Class C 105,642 17,870 B. Outside Containment Contaminated Equipment 58,513 784 Contam. Concrete / Metal 148,366 35 Solidified Liquid Waste Class B 1,802 8,001 Class A 7,694 2,349 Dry Waste Class B 2,509 385 Class A 5,585 170 Total outside Contain. 224,469 11,724 j i
)
Notes: (a) Packed volume in burial containers (b) Activity at 3 years after reactor shutdown (c) All wastes Class A unless indicated otherwise l 4 3-32
1 TABLE 3-15 4 ISOTOPIC DISTRIBUTION OF DECOMMISSIONING WASTE 1 1 Fractional Activity Activated Activated Activated Contam Eq & Contam Conc / Isotope St. Steel Car. St. Concrete Solid Waste Metal & Dry Wat
- C-14 1.12-4 4.86-5 - - -
Ar-39 - - 2.72-3 - Ca-41 - - 4.81-4 - Mn-54 3.99-3 1.05-2 1.02-3 1.44-2 1.41-4 Fe-55 4.44-1 8.50-1 9.56-1 - 1.17-2 , Co-60 4.66-1 1.29-1 3.10-2 9.81-1 6.07-2 < Ni-59 5.92-4 8.18-5 8.18-5 - - Ni-63 8.83-2 9.50-3 9.44-3 - - l Sr-90/ - - - - 1.49-3 Y-90 i Nb-94 3.57-6 - - - - Mo-93 2.78-7 3.32-6 - - - Cs-134 - - - - 1.11-1 Cs-137 - - - 4.96-3 8.15-1 I Notes:
. (a) Normalized distribution 3 years after reactor shutdown.
Total activity given in Table 3-15. l 4 (b) Read as 1.12x10 3-33 l l
1
'4.0 ENCAPSULATION: IN-SITU DECOMMISSIONING AND LOW LEVEL WASTE MANAGEMENT This section describes encapsulation; the concept of decommissioning a nuclear power plant by leaving the containment building intact, moving radioactive waste stored on
[ , the plant site from both plant operations and off-site sources, ! as well as decommissioning waste, inside the containment, then sealing the structure and treating it thereafter as an above f ground low-level waste facility. i 1 4.1 Description of Concept The concept presented here combines provisions for disposing of l all Maine (Maine Yankee plus other State of Maine sources) low-level radioactive waste with eventual decommissioning of the Maine Yankee nuclear power plant. It .is assumed that beginning in 1993, Maine radioactive waste generators will no longer be able to ship low-level waste to the three currently licensed repositories, and that all Maine Yankee waste, as well as waste generated elsewhere in the State, will be stored in the Low Level Waste and Equipment Temporary Storage Building ("Onsite Storage Facility" - OSF) located on the Maine Yankee site. This storage will continue until the plant reaches the I end of its licensed operating lifetime in the year 2008. I For the purpose of this study, the following assumptions have been made: Decommissioning of the plant will begin in 2009 and l a., t for three years. During that time, following removal of the fuel assemblies and highly radioactive reactor internals, and draining of all radioactive liquids, all plant structures outside the Containment Building will be dismantled, and all radioactive waste (" decommissioning waste" plus waste already l stored in the OSF) will be moved into the Containment. 4-1 L
I I i At the conclusion of this decommissioning process, all unnecessary penetrations into the Containment will be j permanently sealed. The equipment hatch, personnel hatch and necessary power and ventilation penetrations would be sealed under administrative control. Following plant decommissioning and storage of existing low-level waste, additional low-level waste generated in the State would be shipped to the site and placed in the i Containment. This would be done on a periodic basis. When the facility capacity is fully utilized it will stop accepting waste, and the remaining penetrations will be sealed. The
- facility, consisting of a sealed containment containing low-level waste, will be left intact as a low-level waste repository.
The sequence of events, starting with the present disposition of low-level waste and extending to final creation of a low-level waste repository is illustrated in Figure 4-1. The
" time-line" is divided into four periods: operating, decommissioning, active repository, and post-closure. These are discussed in the following sections. Note that all dates given for events are assumed as the most likely as known today.
However, the sequence of events and conclusions are not affected by changes in these dates. 4.1.1 Operating Period This is represented as Periods 1A and 1B in Figure 4-1. The entire period extends from the present to the end-of-life for the Maine Yankee plant, which is assumed to be in the year 2008. The first subperiod, lA, ends in December 1992, and represents the current method of handling low-level waste generated in the State of Maine. Main'a Yankee low-level waste 1 4-2
is processed, packaged and shipped to an off-site repository. Low-level waste generated elsewhere in the State is similarly disposed of. Beginning January 1, 1993, hcwever, the Low Level Waste Policy Act requires Maine waste to be disposed of elsewhere. The alternative investigated here is to store it at the Maine Yankee site. During subperiod IB, starting January 1, 1993, all i low level waste generated either at the nuclear plant or at other facilities in Maine would be moved into the Onsite ] Storage Facility on the Maine Yankee site. The types and amounts of waste involved are discussed in Section 3. A further discussion of the Onsite Storage Facility is presented later in Section 4.6. Period 1B would end at the end-of-life of Maine Yankee, assumed to be October 21, 2008.
- I 4.1.2 Decommissioning Period Decommissioning operations would begin following final plant shutdown. Period 2 is assumed to start January 1, 2009 and end on December 31, 2011, a three-year duration. It is not necessary here to discuss the details of decommissioning a nuclear power station; these have been covered for various different options, including immediate dismantlement, entombment, mothballing, and,1,2,3recently, intact decommissioning, in a number of studies. The proposed method of decommissioning is probably closest to the intact method 3
presented in an AIF/NESP study, with the addition here of using the Containment to store low level waste. All systems containing radioactive fluid would be drained, and those systems, along with other contaminated structures and equipment outside the Containment, would be decontaminated. All 1 structures outside the Containment would be dismantled, and all i 4-3 i o i
. o items containing residual radioactivity (discussed in Section
- 3) would be stored on-site in a temporary storage area until the Containment is prepared to receive low-level waste.
The nuclear fuel and highly radioactive reactor internals (greater than Class C activity) would be removed from the reactor vessel and shipped off-site to an appropriate federal facility. The Containment would be prepared by removing any j salvagable items, and constructing any racks or other
' to store the low level waste structures that may be necessary (This will be discussed further in Section 5). It is intended that the minimum of work be performed inside the Containment in order to minimize radiation exposure to the workers.
As soon as the Containment is ready, waste currently stored in the Onsite Storage Facility, waste shipped from other Maine j facilities, as well as waste generated at the plant from decommissioning activities would be appropriately treated, packaged and moved into the containment. Once the site (with the exception of the Containment) is free of residual radioactivity, all unnecessary penetrations in the containment would be sealed (This is discussed in Section 4.3). t 4.1.3 Active Repository Period Period 3 would begin following closure of the containment, which is assumed to take place by January 1, 2012. During this f period, the facility, licensed as a low-level waste repository, would receive waste from Maine generators on a convenient, periodic, basis (say, once-a-year). At those times, the sealed penetration (s) under administrative control would be opened, and the waste placed inside the Containment. This period could be extended as far into the future as desired, but it is assumed here that it would last for 30 years, ending December 4-4
J 31, 2041. Title of the low-level waste repository would pass to. the State ~ of. Maine after decommissioning is complete, and at the beginning of the active repository period. 4.1.4 Post-Closure Period Starting January 1, 2042, the repository would no longer receive any low-level waste, und penetrations under j administrative control (e.g., the equipment hatch) would be l permanently sealed. In accordance with the provisions of 10 CFR 61, monitoring would be maintained for five additional years, designated as Period 4A. Then, on January 1, 2047, a 100-year i period of institutional control (Period 4B) would begin. This latter period would extend to December 31, 2146. 1 i 4.2 Decommissioning Boundary I An encapsulation boundary will be established such that all radioactivity outside the boundary would be removed to inside the boundary, which would then function as a low-level waste repository. The selected boundary must contain any included radioactivity for a period of time sufficiently long to reduce j within safe limits any risk to the public. The regulatory requirements are set forth in 10 CFR 61. The residual radioactivity in the plant, including from decommissioning and } l from items stored in the Onsite Storage Facility, is discussed in Section 3. 1 ! The physical features of the Containment Building, discussed in Section 2 (e.g., 4'-6" thick, reinforced concrete outer walls), i make it particularly attractive to form the encapsulation ! boundary. The structure is designed to the highest standards to i 4-5 4 e e
be able to safely withstand, and contain any radioactive material released, following an extremely unlikely postulated design basis accident involving greatly elevated temperatures i and pressures. The building is also designed to safely withstand design basis ! environmental events, such as postulated, low-probability of occurrence, earthquakes and hurricanes. In addition, all i penetrations through the Containment wall (see Section 4.3) are i also designed to meet severe accident and environmental conditions, and, thus, may be sealed to form a structurally acceptable boundary. , I Figure 4-2 illustrates schematically what the encapsulation f boundary would look like. The boundary separates radioactive items stored inside, from the radiation-free environment outside, and as such, forms the boundary of the low-level waste repository. The technical feasibility of this concept rests on I demonstrating the suitability of the Containment Building only. Therefore, whether or not other buildings could be included is immaterial. 4.3 Sealing Preparations i The Maine Yankee Nuclear Plant has many piping, electrical, . HVAC and access penetration 7 through the Containment. There are l almost 100 piping penetrations (including spares) ranging in
! pipe size from 3/8" to 36", and corresponding sleeve size, from 8" to 42". Particularly large ones includes the purge duct
' exhaustr the fuel transfer tube (going to the Fuel Building);
the main steam lines running from the steam generators to the Turbine Building; the feedwater lines, returning water from the
! steam turbine equipment to the steam generators; and, safety injection lines, providing emergency cooling water to the !
reactor vessel. 4-6 i l 1 . l 1
.- ~ . . .. .., -. . . _ _ . _ . _ _ _ _
7
. It is envisioned that the penetrations would be cut on both the inside and outside of the Containment, filled with concrete, 7
and finally sealed by welding 1/4" steel plate to the 3 containment Building steel liner. l i t 4.4 Structures and Equipment Inside Containment i i i j It is desirable to prepare the Containment Building to receive 1 and safely store low-level radioactive waste for a long period l
- of time with the minimum amount of structural work as possible. !
I ) i This is consistent with the principle of ALARA dose reduction l for the plant and construction personnel. Thus, contaminated
- l piping will not be cut, nor equipment removed (with the i j exception of greater than Class C waste), in the Containment j unless necessary to provide access for transporting low-level
! waste, or locations for storing the waste. Section 5 will l explore this further. I i-i ! 4 i 4.5 Reactor Internals ] With the exception of the nuclear fuel itself, by far the most ) l radioactive components in the plant, containing over 99% of the- l 6 1 , plant's residual radioactivity (Reference 3 estimates over 10 j curies remaining at plant shutdown), are the neutron-activated, stain 1'ess steel reactor internals. These include the thermal
! shield, core support structure, in-core instrumentation, and control rod guide tube assemblies. Since the radioactivity is distributed throughout these items (from neutron activation) they cannot be decontaminated. The internals are classified as greater than Class C waste, and will be shipped to a suitable !- federal repository during the decommissioning period. Keeping l
- 4-7 i i.
g,3-i.- -g-parre y ymmme. g--.i,-*--mm-w &4mmm*5--9e m ' gea - my3e- s ,pm' -ma u" -+ ?--y q q- va d e- w wtw Wmm.*w7 m y i P+- *9.y----- = w y e- ww- W
the reactor internals in place in a sealed containment was considered, and found feasible, in the Intact Decommissioning study of Reference 3. i 4.6 Waste Handling The various waste handling operations and facilities during the periods previously defined (and illustrated in Figure 4-1), will be described. 4.6.1 Operational Waste During the operational period (period IB) of Maine Yankee (1993-2008), low-level waste will be stored in the temporary Onsite Storage Facility (OSF) until permanent disposal is available. The OSF will accept waste from Mcine Yankee and other waste generators in Maine. While Maine Yankee is being decommissioned (2009-2011), the OSF will continue to accept wastes from other Maine generators. In order to minimize any impact on Maine Yankee operations, the OSF will be available to other generators only on a scheduled basis, perhaps quarterly. This schedule for other generators is consistent with the small volume of waste involved and current practice of prearranged (scheduled) disposal. The "public" availability of the OSF will be scheduled to coincide with periods of light maintenance activity at Maine Yankee and to avoid heavy periods, such as during refueling and major equipment repair. It is assumed all waste received and stored at the OSF will be packaged according to U.S. Department of Transportation (DOT) and Nuclear Regulatory Commission (NRC) transportation 4HB
specifications; i.e., all waste form and packaging will be suitable for transportation on public roads and disposal without further waste processing or treatment. Applicable procedures for waste preparation for shipment are given in Maine Yankee procedures. All Maine Yankee low-level waste will i be processed for disposal prior to storage. All waste generators would certify that their waste form and packaging are ready for disposal and long-term storage. The OSF storage capacity will be increased as necessary to handle any expanded storage requirements. The OSF will have controlled access and radiological monitoring capability. However, the OSF will not have capability for repackaging or
, waste treatment. If necessary, repackaging and waste treatment will be performed in other Maine Yankee facilities.
I The waste packages will be stored under conditions that will
} assure package integrity during 10 years of storage and subsequent package handling. The problems associated with long-term storage and technical feasibility of long-term storage are explicitly addressed in Section 5.1.
As required by NRC 10 CFR 20.311 and 61.80, the OSF will maintain a waste manifest and record keeping system to keep track of the location and contents of all waste containers. The waste packages will be segregated by source and contents. The Maine Yankee waste will be kept separate from other waste. Special designated areas in the OSF will be established for each category of waste, e.g., Class A, B, and C.
) ; 4.6.2 Decommissioning Waste i l As described in Section 3.4.2, the vast majority (greater than 901) of the decommissioning waste is bulk materialt e.g.,
4-9
. o 1
contaminated and activated metal, contaminated and activated
; concrete, and contaminated equipment. (The amount of process waste from decontamination activities is a relatively small portion.) The bulk material may first be cut-up (" sized"), if necessary, to fit into large bulk waste containers, e.g., steel 3
boxes (100 ft 3), disposable liners (170 ft ) or any other i l suitable containers. l The bulk waste containers should have sufficient structural l strength to contain the debris material since they are designed 1 > for solidified (cecent) waste. Use of large construction j dumpsters are also possible if sealed to prevent releases. Due to limited storage space in the OSF, the bulk decommissioning waste of low specific activity (LSA) will be stored on a flat surface next to the OSF. The containers will be treated to minimize the effects of surface corrosion. All other decommissioning waste, such as higher activity bulk waste and ! processed waste will be stored in the OSF. While Maine Yankee is being decommissioned, the OSF will continue to accept low-level waste from other Maine generators on a scheduled
- basis, as described previously in Section 4.6.1.
i i 4.6.3 Disposal When the Maine Yankee decommissioning preparations are completed, the containment and decommissioning boundary will be ready to accept low-level waste for disposal. The placement of the low-level waste in the Containment will be organized in such a manner to minimize any releases and maximise long-tera j stability and isolation. Accordingly, class a and c waste will be segregated from Class A waste. All Class C waste will be placed in the 472 ton reactor pressure vessel. The pressure j l vessel (8-5/8"-thick carbon steel) provides an additional (and 1 ! sufficient) intruder barrier to isolate all class C waste. l Details.of the placement plan appear in section 5. j i 4-10 I . .
In addition, all the Class B solidified liquid waste and all the dry waste can also be placed inside the pressure vessel. The remaining class B waste (operating waste) will be placed in the refueling pool. The Class A waste containers will be stacked in various locations in the Containment to a height of 20 feet. Shelves may be required to provide structural support to prevent rupture of the waste containers during placement. 4.7 Surveillance, Maintenance and Security The surveillance, maintenance and security requirements of the low-level waste during the operating period, Msine Yankee decommissioning, and after final closure of the encapsulation ! boundary are discussed below. I
; 4.7.1 Operating Period During plant operations, the long-term storage environment of low-level waste in the OSF will be controlled by normal Maine Yanxee plant procedures. Since the OSF is within the plant boundary, OSF security and environmental surveillance will be part of Maine Yankee's plant system. The major maintenance
] ] , requirement during long-term storage is the concern for package i integrity and the generation of gases. Periodic t non-destructive testing (NDT) and inspection of package integrity will be necessary to assure safe storage and f i subsequent safe handling and placement during final disposal. I } 4-11
o e 4.7.2 Decommissioning Period During decommissioning of Maine Yankee, the operation of the OSF will remain the same as during plant operation. The Maine ;
' Yankee radioactive waste treatment system will be one of the l ,
last plant systems to be decommissioned.- The major maintenance i item is the outdoor storage of the bulk waste containers 3 L resulting from decommissioning, approximately 200,000 ft . Use , of a thick high-density polyethylene membrane (or similar device) to protect the waste containers from the elements should be sufficient for the short-term period involved until. l l the Containment is ready to receive waste. i l i 4.7.3 Final Closure Period After placement of the waste in the encapsulation boundary and facility closure, the disposal site begins a " surveillance" , period, of 5 years as mandated in 10 CFR 61. During this surveillance period, the site operator maintains a detailed monitoring program to demonstrate the disposal system is significant discrepancies are : performing as designed. If detected during this period, then adjustments to the disposal system are made. The only contemplated maintenance activities are common grounds keeping. : Following the surveillance period, the disposal facility license may be transferred to the institutional custodial agency and site owner (10 CFR 61.30). The NRC has the authority f to approve this license amendment. Associated with the transfer of site responsibility is the accessibility by the custodial agency to funds for the permanent maintenance of the site. (This trust fund for institutional care was endowed by the waste generators.) The custodial agency is required to maintain
, the site for whatever length of time necessary (100 years 4-12
a . 1 minimum) to ensure compliarce with the Performance Objectives
! (Part 61 - Subpart C). The four Performance Objectives involves (1) radiation protection limits for the general populace (61.41).
(2) protection of the individual from inadvertent' intrusion (61.42). i (3) radiation protection for workers (61.43). (4) long-term stability of the disposed waste and disposal I site (61.44). t i These performance objectives control the level of effort } necessary for site maintenance, survoillance and security. The j environmental surveillance monitoring program must provide "early warning of releases of radionuclides from the disposal site before they leave the site boundary." (61.53(d)). Site j security must be sufficient to prevent inadvertent intrusion, e.g., daily patrols and physical intruder barriers. Site maintenance must be commensurate with the long-term stability of the site. a i I a y J 4 I I I 1 i 4-13 i e l
4.8 References
- 1. W. J. Manion, T. S. LaGuardia, "An Engineering Evaluation of Nuclear Power Reactor Decommissioning Alternatives",
AIF/NESP-009, November 1976.
- 2. R. I. Smith, G. J. Konzek, W. E. Kennedy, Jr., " Technology, Safety and Costs of Decommissioning a Reference Pressurized i Water Reactor Power Station", NUREG/CR-0130, prepared by
, Battelle Pacific Northwest Laboratory for the U.S. Nuclear
! Regulatory Commission, June 1978. " I j 3. S. Ostrow, et al., " Intact Decommissioning of Nuclear Power Plants: A Dose Assessment", AIF/NESP-034, March 1986. f i i I , I i 4-14
__ __ . . _ _ . _ . . _ . _ _ _ _ _ _ _ _ _ _ _ _ . . . . _ . . . . _ - _ _ ____ _ _ . _ .__ _.m.. _ _ _ . _ . _ _ _ . _ .__ _ _ . _ _ _ _ _. i l ACTIVE REPOSITORY OPERATING PERIOD DECOM. PERIOD, PERIOD POST-CLOSURE PERIOD IB 2 3 4A 4B IA Continue to ship Low-Level Radio- In-situ decommis- Facility, licensed Facility Institutional sioning of Maine as LLW repository, permanently control peried all Maine LLW to active Waste out-of-state Policy Act makes Yankee (3 yr). and run by State closed. (100 yr). repository. LLW a State of Maine LLW responsibility Authority, receives starting 1/1/93. Encapsulate all LLW from Maine Environmental ! LLW (stored in generators. monitoring OSF + decom. period (5 yr).
,8, Store all Maine waste) in
- v. LLW in Maine Containment Yankee's Onsite Storage Facility.
Title of repositor:
- passes to State at Maine Yankee end of decom.
- period.
operating license
- expires 10/21/2008.
' 41/1/2009 1/1/87 1/1/93 0/21/2008 1/1/2012 1/1/2042 1/1/2047 1/1/2147' FIGURE 4-1: IN-SITU DECOtMISSIONING AND LOW-LEVEL WASTE MANAGEMENT TIME LINE
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j I I l _ 4-16
N 5.0 TECHNICAL FEAS,IBILITY OF CONCEPT The concept of combining low-level waste storage and disposal with decommissioning of the Maine Yankee nuclear power plant i was discussed in Section 4. This section examines the technical feasibility of the concept. No attempt has been made to do detailed engineering or to optimize any figure of merit (e.g., occupational dose exposure, lowest cost, lowest shortest i decommissioning time, etc.). Rather, the intent is to 1 investigate whether the concept is feasible and faces no
- [
insurmountable impediment which would preclude its reasonable execution. 5.1 Low-Level Waste Storage During Normal Operation The concept calls for storing all Maine Yankee operational low-level waste, along with waste generated elsewhere in Maine, in the Onsite Storage Facility (OSF) from January 1, 1993 (as set in the Low Level Waste Policy Act) until the plant is shut down in 2008. This period of fifteen years is triple the design basis of only five years storage which the Onsite Storage Facility was licensed to accommodate. In addition, the original design did not allocate storage space to non-Maine Yankee waste. Feasible alternatives to provide additional storage space include either expanding the existing facility or building an additional facility on the site. As discussed in Section 1.4.2.1, exceeding the five year design limit in the existing building would probably require a part 30 (10 CFR 30) license. A further consideration for storing waste more than five years in the OSF would include the necessity to ensure the structural 5-1 i'
- integrity of the containers, since it would be undesirable to be faced with handling degraded containers during the decommissioning period when they would be moved into the Containment.
5.2 Encapsulation Boundary All low-level waste stored in the OSF, as well as that generated during decommissioning activities, would be moved
. into the Containment during the decommissioning period.
Finally, all penetrations into the Containment would be cut and sealed with the exception of those necessary for operation as a repository. The sealed Containment outer wall, dome and base mat would then constitute the boundary of an engineered, above ground low-level waste repository. 1 The AIF/NESP Intact Decommissioning study presented a similar concept for sealing and leaving the Containment in place, but without storing additional waste inside it. That study concluded that it would be feasible (in fact, relativelyy straightforward) to cut and seal the penetrations as postulated. In fact, the sealed penetrations could be made to have the same strength as the structure through which they penetrate, the Containment Building. 5.3 Preparation of Containment for Waste Storage It is envisioned that one of the early tasks in the decommissioning of the plant would be to prepare the Containment to receive low-level waste. Section 5.2 discussed the penetration sealing operations and concluded that they could reasonably be carried out. 5-2
Tne distribution logic of the containerized waste is discussed in Section 5.4, and the containerization scheme and placement are shown in Tables 5-1 and 5-2. As seen in Table 5-2, the floor loadings n several areas are guitehigh, ranging from about 800 lb/ft to almost 2,500 lb/ft . These can be compared to typical design loadings for floors in the Containment (according to structural safety codes used for nuclear plants 2 which allow substantial margin) of the order of 100-300 lb/ft . However, it should be possible to reinforce the floors where the waste would be stored if detailed analysis requires it. l Two alternative floor support plans appear feasible. The first would be to backfill with concrete the lower floors where the I waste would be stored after the waste containers are placed in those areas. The concrete would transmit the loads directly to the foundation base mat. The second alternative would be to provide additional beam supports under the floor supporting the waste. The beams could either be tied to the thick concrete, annular side walls of the annulus area on the lower levels, or,
. if additional strength is required, to new, steel supporting columns. Once again, loads would be transmitted to the base mat. At this stage, it appears that the beam solution would the preferred alternative due to its simplicity of implementation and lower added load to the structure.
1 I The beam support structure would ensure that the static loading resulting from the addition of stored waste would be within t acceptable limits. However, dynamic loading considerations (e.g., from seismic events) must also be addressed. Extra internal bracing around the stored waste could also be added as required. Other modifications that would be made inside the Containment would be relatively minor. They basically consist of clearing equipment out of the annular ring areas so that waste may be stored there. Additionally, to improve access from the 5-3
equipment hatch' where the low-level waste containers would enter to the lower building level (El. - 2'-0"), it would be desirable to enlarge the opening in the wall that currently surrounds the reactor head laydown area from 4 feet to a height suitable for a fork lift truck carrying waste containers to pass through. This fork lift should have the capacity to handle 6 ton loads and lift them to a height of approximately 11.5 feet (the height of the fourth drum in a stack). Specifically, the structural modifications that would be made in the different areas of the Containment are: Elevation - 2'-0" (1) Remove cable trays and instrumentation racks from the annulus area. (2) Remove the ramp and raise the elevator so that waste could be stored in that area. (3) Place additional bracing (if needed) to accommodate the large load on the floor above (El. 20'-0"). (4) As the area is filled with the waste canisters, some i amount of backfilling or bracing may be necessary to minimize the degree of container movement in the event of I an earthquake. (5) A heavy, waterproof coating may be desirable on the outside facing walls to minimize the possibility of water infiltration, j Elevation 20'-0" l (1) Remove HVAC equipment, cable trays, etc. from the area. 1 5-4
(2) Install bracing (as necessary) for the expected high load on the operating deck located above. 1 (3) Install restraints (as necessary) to prevent excessive container movement in the event of an earthquake. Elevation 46*-0" (Operating Deck)
. (1) Remove equipment, etc. from annulus areas.
(2) Remove fans FN 43-1 and FN 43-3. and cover the open stairwell. (3) Brace waste containers to the floor by providing some sort of attachment method. This is to minimize missile generation in the event of an earthquake. (4) The overhead polar crane must be maintained in working order during the decommissioning period as its auxiliary hook will be used to move waste containers, as well as remove existing equipment, and bring in new items (e.g., steel bracing, fork lift). After decommissioning, the crane will no longer be needed. I 5.4 Waste Placement and Storage This section will first consider whether the expected volume of waste will be able to physically fit in the usable free spaces in the Containment (subject to removal -of equipment as discussed in Section 5.3), and then how the waste would be moved and placed in the storage locations. t 1 5-5 l
J i 5.4.1 Waste Volumes and Storage Areas j Table 3-14 lists the amounts and types of radioactive waste generated during decommissioning operations, while Section 3.3 does the same for non-decommissioning waste. Taken together I (but not including the portion of the radioactive inventory located in the Containment at the time of decommissioning since it will not have to be moved), they represent the total
- inventory of low-level waste which must be placed in the 1
4 Containment during the decommissioning period. Table 5-1 l i summarizes this information, showing the waste description,
! volume, and required number of containers of a particular size l needed to store it. Typically, the container size chosen was 1
J less than the maximum possible to facilitate handling, and 1 .I maximize the amount of waste that could be fit into a given 1 area (i.e., smaller containers pack better than larger ones). , As such, it differs from past practices, and projections made under the assumption of continuing current low-level waste disposal practices, given in Section 3 Table 5-1 also includes the non-Maine Yankee waste assumed to be placsd in the Containment yearly for 30 years following decommissioning. The heaviest containers were placed at the lowest elevations to better transmit loads to the base mat, and to minimize dynamic effects during a seismic event. Assumed, average waste 4 densities were taken from Reference 2. Stacking heights were limited to 14 feet in the two levels below the operating deck to allow 4-foot clearance for possible placement of. , reinforcement beams for supporting the higher levels, as well as to allow sufficient clearance for ease of waste container placement. Table 5-2 summarizes where the various types and amounts of 1 waste listed in Table 5-1 would be stored in the Containment.
- The arrangement is by no means unique or optimal, but is-certainly feasible. It can be seen that most of the 5-6 4
..-m
1 I l , i l t contaminated equipment, and the concrete and metal debris from decommissioning are placed in the lowest level (Elevation i -2'-0"). This represents the heaviest waste. The second level (Elevation 20'-0") is used primarily for storing non-Maine 4 Yankee waste. Barrels are stacked only. two high here to minimize the floor load. i " All the Class C waste (most radioactive), and most of the Class
~
B waste would be stored in the reactor vessel (from which the t greater than Class C internals may have been removed), which, with its head placed back on, represents a very substantial additional containment for the waste inside it. (Note, i locations can be found by referring to Figures 2-3 to 2-7.)3 The remainder of the Class B waste, along with almost 50,000 ft of Class A waste would be stored in the refueling cavity. The j total weights of waste placed in either the reactor vessel or the refueling cavity appear similar to the normal equipment plus water load for the former, and water load (during
- refueling) for the latter. Therefore, the waste should be adequately supported in these areas. Finally, the balance of
- the contaminated material and the balance of the other Class A waste would be stored in locations on the operating deck and in i
l the annular regions below this deck. 5.4.2 Waste Handling i ! Tne handling schemes postulated were chosen for their-y simplicity and appear feasible. Two periods are considered; i
- during and after the three year decommissioning period. l I
; Waste would enter the Containment through the equipment hatch
$ which opens onto the Elevation 20'-0" level. The auxiliary hook
] of the overhead crane will be used during the decommissioning period (but will not be required thereafter) to lower pallets holding waste containers to the -2'-0" level, and to lift waste 5-7
. o l
1 up to the operating floor. Fork lift trucks would then be used to transport the waste to the selected storage locations. After completion of the decommissioning phase, non-Maine Yankee waste will periodically be brought into the Containment and placed in locations on the Elevation 20'-O" level. This will minimize the amount of handling required, and, consequently, operator time in the Containment. The smaller personnel hatch (7' diameter vs. 22' diameter for the equipment hatch) could also be used for entry. This would minimize any airborne contamination that might be released to the environment each time the Containment were opened. It also should be feasible to
! use a robot to carry the waste into and around the Containment if this is desired.
5.5 Containment Integrity 4 The encapsulation boundary for the low-level waste repository consists of the Containment outer wall, dome and base mat. It should maintain its integrity for a period of time sufficiently l
; long to safeguard the public from the contained waste (which is radioactively decaying over time). The AIF/NESP Intact 1
Decommissioning Study made an extensive evaluation of the ability of the sealed Containment to withstand the long-term degradation processes in concrete and steel (carbon and stainless) subjected to environmental effects such as groundwater, weather, atmospheric pollution, high wind, and i seismic disturbances. A range of environmental conditions were chosen in the evaluation process for the " generic" nuclear plants (PWR and BWR), including " worst case" salt water, groundwater and atmospheric chemistries. A number of Containment failure mechanisms were considered, with the following initiating events (Reference 1, Section 5.1.11.1): I 5-8
(1) Weathering of concrete. (2) Rebar corrosion. (3) Water permeation of concrete. (4) Rise of groundwater table. (5) Site flooding. - (6) Severe environmental loads at design basis. t (7) Extreme environmental loads at design basis. (8) Catastrophic load beyond design basis. The Intact Decommissioning study concluded that a nuclear power plant containment is sufficiently robust, designed to the highest civil engineering standards, to withstand the effects of very severe design basis events (including high temperature and pressure during a LOCA) for a long period of time following decommissioning, without any inspection or maintenance being done. In the Maine Yankee case, routine inspection of the Containment exterior (e.g., for concrete cracking), and performance of any required repairs, will continue through the decommissioning period, through the nominal 30 year period after that when non-Maine Yankee waste will be accepted, and through the following period of institutional control. Thus, the Maine Yankee Containment should stand as a barrier to release of radionuclides for at least the 100 years given by
~
the Intact Decommissioning study. Furthermore, The Intact study very conservatively chose 100 years as a "round number" by which time all of the short and most of the medium-lived radionuclides would have decayed away; there is good evidence that the Containment would actually' continue to perform its function for several hundred years more. 5s9
l 1 It is not the intent here to duplicate the extensive analysis of the Intact Decommissioning study since that study's results for a reference PWR should apply directly to the Maine Yankee plant. There is strong evidence that the Maine Yankee Containment should successfully function as a barrier to radionuclide migration for at lest several hundred years. In addition, all of the most radioactive waste, Class C, and most of the Class B waste will be stored inside the closed reactor vessel. The Intact Decommissioning study showed that the extremely thick, steel vessel would provide a barrier to water intrusion (and subsequent, possible radionuclide migration out into the environment) for several thousand years even if the vessel were constantly immersed in sea water with the " worst" water chemistry. 5.6 Environmental Surveillance The technical feasibility to maintain an environmental surveillance program, as required by NRC's 10 CFR Part 61.53, is not encumbered by this permanent disposal concept. In fact, ; the environmental surveillance to detect the release of radionuclides is facilitated by an above-ground disposal structure as opposed to a below-ground burial facility. Any leakage from the Containment would be more promptly detected by the monitoring system as compared to a below-ground facility. Potential releases from an underground disposal system would be attenuated and delayed by the soil cover. Leakage from a surface facility can be more readily detected, isolated and remediated than from an underground system. l 5-10 P
5.7 Waste Integrity The stability of Class B and C waste is not adversely affected by this permanent disposal concept. Part 61.56(b) requires the Class B and C waste to have structural stability (so as not to adversely affect the disposal system) and to limit exposure to an inadvertent intruder. Because the containment structure inherently provides structural stability after disposal, it is redundant and conservative to have the waste form also satisfy the stability requirement. As an additional precaution all the Class C waste is placed and sealed in the reactor pressure vessel. In summary, waste integrity and waste stability requirements can be met by this disposal concept. 8 5-11
5.8 References
- 1. S. Ostrow, et al., " Intact Decommissioning of Nuclear Power Plants: A Dose Assessment", AIF/NESP-034, prepared by Ebasco Services, Inc. for the National Environmental Studies Project of the Atomic Industrial Forum, March 1986.
- 2. R. I. Smith, G. J. Konzek, W. E. Kennedy, Jr., " Technology, Safety and Costs of Decommissioning a Reference Pressurized Water Reactor Power Station", NUREG/CR-0130, prepared by Battelle Pacific Northwest Laboratory for the USNRC, Vols 1
& 2, June 1978, Addendum 1, August 1978, Addendum 2, July 1983, Addendum 3, September 1984.
i, f l i I t f i i l l 5-12 l i i \ l
. e TABLE 5-1 CINIAINERIZED IIM 12 VEL WMITE STREAMS Packed Voly Contain y N aber Mat D p Stack Waste Description ft size ft Ckmtain p/m Height nwr==issioning Waste Contaminated Equipnent 58,513 7.5 7,802 0.4384(a) 5 centaminated ocncrete and metal debris 148,366 7.5 19,783 1.25(b) 5 Solidified Liquid Waste Class B 1,802 90 21 2.3(c) 2 Class A 7,694 90 86 2.3(c) 2 Dry Waste Class B 2,509 90 28 2.3(c) 2 Class A 5,585 90 63 2.3(c) 2 Main Yankee Operating Waste Class A 78,880 120 658 1.0(d) 2 Class B 800 90 9 2.3(c) 2 Class C 320 90 4 2.3(c) 2 Instituticnal Waste (1993-2008) Class A 28,000 7.5 3,734 1.0(d) 2 Institutional Waste (2009-2041) Class A 46,200 7.5 6,160 1.0(d) 2 Notes, i (a) Density obtained from calculation of average density of ! decomissioning waste as shown 3 e G.& 5 of h fer m 2, excluding concrete (27.3746 lbs/ft ). (b) Density obtained frce calculation of average debris density frost values 3 gh 6 WeGh5 of M M m 2 h m ete D8.102 lbs/ft ). (c) Assumed worst case for waste stabilized in concrete. (d) Assumed average density equals that of water. 5-13 i
4 TABLE 5-2 DISTRIBUTION OF WASTE IN CONTAINMENT Packed 4 Volgme Average L ad 2 Location Waste Type & Source it lbs/ft Reactor Vessel Class C Waste - Norm op 320 -l 665,127 lb Reactor Vessel Class B Solidified Liq 1,802 l-total for Reactor Vessel Class B Dry Waste 2,509 -l vessel Refueling Pool Class B operating 800 -l 4,207,690 lb Refueling pool Class A Sol Liq 7,694 l-total for Refueling Pool Class A Dry Waste 5,585 l pool Refueling pool Class A Operating 35,000 -l Elevation -2 ft Ring Area Contaminated Concrete 148,366 2,494 Ring Area Contaminated Equipment 7,986 874 Ramp Area Contaminated Equipment 5,431 874 Head Laydown Contaminated Equipment 23,959 874 Elevation 20 ft 4 Ring Area Institutional Waste 16,949 797 (1993-2008) Ring Area Institutional Waste 46,200 797 (2009-2041) Jib Crane Area Institutional Waste 5,926 797 (1993-2008) Elevation 46 ft (Operating Deck) Ring Area Institutional Waste 2,454 797 (1993-2008) Ring Area Class A Operating 39,179 954 FN43-1 Area Contaminated Equipment 10,569 874 FN43-3 Area Contaminated Equipment 10,569 874 Pressurizer Class A Operating 4,168 954 Area FN43-1 Area Class A Operating 534 954 FN43-2 Area Institutional Waste 2,673 797 (1993-2008) 5-14
e 6.0 RADIOLOGICAL ASSESSHENT Radiological assessment during each of the phases which will be i followed in the waste disposal and decommissioning scenarios l are considered of prime importance to the successful implementation of the central concept of the study. As such, the assessments discussed below are based on the most recent studies conducted. Primary among these are: " Update of Part 61 1 Impact Analysis Methodology", for waste disposalt and, " Intact 2 Decommissioning of Nuclear Power Plants: A Dose Assessment", for decommissioning. The former study was conducted for the NRC and the latter for the AIF/NESP. Details of the various waste and decommissioning scenarios may be found in Section 4. Exposures to workers and the public are limited by the 10 CFR 20 regulations. Traditionally, and as will be shown below, these limits are seldom reached in actuality. A common practice ! is to establish action levels which- are fractions of the limits. Actions to reduce or mitigate doses are implemented as these action levels are reached. This effectively minimizes exposure and is in keeping with the concept of ALARA. Public and occupational doses will be examined for transportation, disposal, decommissioning, and operation of the Containment as a low-level waste repository. Container number and sizes for a fixed amount of waste are taken from Section 3 rather than from Section 5 which chose many smallcontainers to j demonstrate that the waste could be placed in the Containment. However, dose estimates are relatively insensitive to the details of the container number-size tradeoff, and the conclusions are unaffected. 6-1 i
s e 6.1 Transportation Doses Transportation impacts are expected to be minimal. Waste generated on-site will also be processed and stored on-site and, therefore, is eliminated as a source of radiological transport impact. This effectively reduces occupational and public exposure which would otherwise be incurred during shipment to a distant repository. Off-site generated waste will be assumed to be transported by truck to the Maine Yankee plant site. The population and occupational exposures incurred during the transport of off-site waste is calculated based on the number of loaded miles, which eliminates consideration of the return trips when the truck is empty. The number of shipments and the number of rest stops complete the basis for the transport scenarios. 6.1.1 Population Transport Dose Two components of population exposures are considered; the dose to populations along the transit route and the dose to on-lookers at the rest stops. The population dose during transit is estimated using the following equation: P=TxKxD. x (L /V ) x TDOZ , ( 6.1.? 1 i i i i where, P = Total population exposure, person-mrem T = Total number of shipments, number / year 6-2
K = Source strength, mR-ft / hour D, = Population density in region i, people /mi 1 L, = Distance in region i, miles / trip 1 V, = Speed of the vehicle, miles / hour 1 2 TDOZ = Dose factor, (miles /ft) i The total number of shipments per year is estimated assuming a 3 generation rate of 1,400 ft / year (Section 3.2) of class A 3 waste packaged in 55 gallon drums (7.5 ft ). Each shipment consists of 70 drums which requires three shipments per year. Note, that although the exact dose numbers depend on the choice of number and size of containers to carry the total amount of shipped waste (the more containers, the lower the activity of each), the dependence is not strong, and the conclusions remain the same whatever the choice. The source strength, K, corresponds to that of an assumed point source at the center of the truck as extrapolated from the radiation level at the truck's surface. For the case at hand, 2 the source strength term is calculated to 1be 33 mrem-ft /hr. following the methodology in NUREG/CR-4370, Equation 3-9. The assumptions for the calculation are: soil density equal to packaged waste density, and a radionuclide concentration in Table 3-14 of Section 3 for Co-60, H-3 and C-14. Only one region is considered with a population density of j l 2,280 people per square mile and the distance of travel in the region is taken as 150 miles. This is approximately the j distance from Wiseasset to Portsmouth. The truck travel speed ! is assumed to be 50 miles per hour. 6-3
O e The last factor TDOZ g is a complex coefficient relating the perpendicular distance from the vehicle center, the linear attenuation coefficient in air, and the Berger buildup factor for gamma radiation. For regions in the Northeast NUREG/CR-4370 (Table 3-15)1 estimates this coefficient as 7.06X10 -5 (mi/ft)2 . Substituting in Equation 6.1 leads to an , estimated population dose of 48 person-arem. The dose to on-lookers at rest stops is calculated using 'a similar expression. The equation is: P=TxFxD xS x SD x TDZ (6.2) i i i i where,
- P = Total population exposure, person-mrem i
T = Total number of shipments, number / year i 2 K = Source strength, mR-ft / hour (see above) 2 D. = Population density in region i, people /mi 1 S. = Number of stops in region i 1 SD = Duration of stops in region i, hr TDZ, = Dose factor, (miles /ft) 1 1 As with TDOZ,, the value for TDZ is taken from NUREG/CR-4370 , 1 i Table 3-16 which relates gamma ray buildup to and distance from the vehicle. The duration of the stop is assumed as one hour. -7 Evaluating the expression (taking TDZ = 9.57 x 10 2 i -2 (miles /ft) ) leads to a collective dose of 22 x 10 person-mrem. In summary, the total dose to the population, 48 person-mrem, is insignificant. 6-4
o a 6.1.2 Occupational Transportation Dose Occupational transport doses include dose to truck drivers during transit, exposures during rest stops (inspection of the truck) and exposures to waste handlers during loading and off-loading. The dose to the driver is calculated from: H=2xTx (L,/V,) x (K/100) (6.3)
< 1 1 where H is the exposure T, L and V are as previously described, and K is the radiation level at the surface of the vehicle modified by the cab correction factor .
(K = 33 mR/hr)3 Substituting in the above leads to a dose of 42 X 10 person-rem. The dose to drivers during rest stops assumes the following: Each stop lasts one hour. During this period, it is assumed that the drivers inspect the vehicle for 5 minutes and walk around the vehicle for 10 minutes. During the remaining 45 minutes they are assumed to be too far from the vehicle to receive appreciable exposures. Occupational exposures are calculated using the following formula: H=2xTxS, x SD, x DF x (K/100) (6.4) 1 1 where all the symbols have been explained before except DF. This factor is the distance factor accounting for the proximity of the drivers to t!' vehicle and the shielding afforded by the 1 intervening air, and is taken from NUREG/CR-4370 as DF=0.235.
-4 Substituting leads to a dose of 5 x 10 person-rem.
The last dose considered is to workers loading and off-loading f the truck. This is estimated assuming a package generic j radiation field of 0.1 mR/hr, a loading time of 12 person-minutes per drum, and 3 shipments per year consisting of
-3 70 drums / shipment. The dose is equal to 4.2 x 10 person-rem 6-5
.. o
- (0.1.(mrem /hr) x3 (shipments / year) x 70 (drums / shipment) x 12 i- (person-minutes / drum)).
't 6.2 Disposal Doses The next dose scenario to be considered occurs during disposal. In the following, the scenario and parameters related to the scenario are discussed. Occupational dose is considered following the dose to the public. 6.2.1 Population Disposal Dose The dose to the population is a function of the shielding, location, and waste placement arrangement in the Onsite Storage Facility (OSF). Given a one foot-thick concrete wall, the dose reduction is about a factor of ten. The dose at the surface of the OSF is estimated from the concentration and the DCF for a i , semi-infinite slab source as found in NUREG/CR-4370, Table f D-8. Multiplying gives a dose rate at the surface for process waste of 0.6 rem /hr and for trash of 20 mrea/hr. For a single box of waste, this is an over estimate of the dose, but for an OSF filled with boxes of waste this is a good estimate of the dose rate at its side. Assuming that the process waste is placed in the center of the OSF so that it will be shielded by other waste containers, and taking credit for the shield wall, gives a dose rate outside of 4 the building of 2 mram/hr. The nearest residence is found at a distance of 700 meters. Using a dose rate outside the OSF of 2 mrem /hr gives an annual direct dose of less then 1 mrem. 4 6-6 i
u s 6.2.2 Occupational Disposal Dose Occupational doses during disposal are a function of the waste volume, the disposal package and the time spent handling the package. These are considered below. The Maine Yankee generated 3 waste volume is estimated to be 5,000 ft /yr (Section 3.1.2). 3 An additional volume of 1,400 ft /yr will be shipped from off-site generators. Process waste, assumed to be put in 170 ft liners, is estimated to be 25% of the previous volumes. This leads to 7.4 3 i liners /yr. Trash, the other 754, will be placed in 100 ft boxes. The number of boxes per year is calculated to be 38 i boxes per year. As mentioned before, the assumption of size vs. I number of containers does not affect the conclusions. l 3 The number of drums per year is simply 1,400 ft /yr divided by 3 i 7.5 ft / drum, or 187 drums / year. The time spent handling each I container is taken from NUREG/CR-4370, Table 4-13. These are 710 person-minutes per liner, 120 person-minutes per box and 13 person-minutes per drum. i Multiplying the number of containers by the exposure rate gives
! an annual dose of 2.1 person-rem.
l l 1 l l 1 6.3 Decommissioning Doses Since this is a feasibility study, no attempt will be made to quantify exactly the occupational exposure that would be experienced during the decommissioning period. Rather, doses will be considered in comparison to those of other i decommissioning options. The AIF/NESP Intact Decommissioning 2 study, Table 7-1, lists doses for the 3,500 MW(t) reference PWR plant. However, since the Maine Yankee reactor is rated at 6-7 1
e only 2,630 MW(t), the doses must be scaled appropriately. The formula for the scaling factor is (Reference 2, Section 3.2.1):
~
SF = 0.324 + (2.035 x 10 ) x PPR, (6.5) where "PPR" equals 2,630 for the case of Maine Yankee. Evaluating the formula gives, SF = 0.859. Applying this factor to the table in Reference 2, Section 3.2.1, gives: Decommissioning Option Dose, Person-ren l Intact 574 i Immediate Dismantlement 1,051 Entombment with react internals 773 w/o 859 Mothballing 357 The Intact Decommissioning occupational exposure of 574 person-rem assumes that aside from sealing the Containment l penetrations (as in the scheme considered here), little else is l done in the Containment which is then left standing while all l other plant structures are removed. The Immediate Dismantlement occupational exposure of 1,051 person-rem assumes that the Containment is also removed (resulting in substantial exposure). Since the scheme considered here leaves the Containment standing, and the highest radiation objects (e.g., primary 1 coolant piping and equipment, pressurizer, steam generators) completely alone, but does do some additional work in the lower radiation, outer areas of the Containment to prepare for then
e s place low-level waste, the occupational exposure should be greater than that for Intact Decommissioning, but significantly lower than that for Immediate Dismantlement. I 6.4 Population Dose From Containment Storage j The last exposure to be considered is the dose to the population from the low-level waste that will be stored in the containment following the decommissioning period. The direct dose to people standing outside the Containment should be negligible (significantly less than background) by virtue of the great thickness of the Containment at 4.5 ft of concrete, plus concrete inner walls. 10 CFR 61 sets standards for the storage of the different classes of low-level radioactive waste. Containment times take q into account the fairly rapid (compared to some isotopes contained in greater than Class C activated stainless steel waste) decay of the constituent isotopes. Hence, Class C waste should be confined for at least 500 years. Since, as discussed in Section 5, this waste is not only protected by the Containment, but is also sealed inside the massive, steel reactor vessel (which should remain sealed for at least several thousand years), this confinement criterion should easily be met. Confinement of Class B waste for a period of 300 years is sufficient. Again, most of this would also be placed in the ! reactor vessel. Finally, Class A waste should be confined for at least 100 years, the very conservative period assumed for I - complete containment integrity by the Intact Decommissioning 2 study. Although no additional low-level waste was assumed to be placed in the Con'tainment following plant decommissioning in the 2
- Intact Decommissioning study, the very highly radioactive, 6-9
v e,
- greater than Class C, reactor internals, consisting of i
activated stainless steel, was assumed to remain in place. Some j important nuclides in stainless steel have very long half-lives 1 (in the tens of thousands of years). Nonetheless, the study concluded, after examining many potential exposure pathways i from the source to man, that both the individual and population doses for the public would be acceptable, and approximately no greater than in Immediate Dismantlement. The same conclusion is expected from this study. l t l I l i d i A 1 4 2 6-10 I l *
- l. . --. _ .. -. . . - - . _ . - - - . . . - . - . .-_ . - - . - - . - . - _ -
- r 6.5 References
- 1. o. I. Oztunali, G. W. Roles (USNRC), " Update of Part 61 Impacts Analysis Methodology", NUREG/CR-4370, prepared by Ebasco Services, Inc. for the USNRC, January 1986.
, 2. S. Ostrow, et al., " Intact Decommissioning of Nuclear Power Plants: A Dose Assessment", AIF/NESP-034, prepared by Ebasco Services, Inc. for the National Environmental Studies Project of the Atomic Industrial Forum, March 1986.
1 i ) J J l i l f f i 6-11 1
w e
7.0 CONCLUSION
S AND RECOMMENDATIONS This study examined the technical feasibility of combining low-level waste disposal and management with decommissioning of the Maine Yankee nuclear plant, whereby, beginning in 1993, all low-level waste generated in Maine would be stored in the Maine Yankee Onsite Storage Facility until the plant reaches its scheduled end-of-life in 2008. At that time, the plant systems and structures would be decontaminated and removed with the exception of the Containment Building, which would then be used as a permanent, sealed, State-controlled, low-level waste
; repository. The Containment would continue to receive waste generated in Maine for a further period of 30 years, then it would be permanently closed.
Issues relevant to technical feasibility were considered for three phases: (1) during normal plant operations; (2) during decommissioning; and (3) during use of the Containment as a permanent low-level waste repository. Some of the major conclusions were (1) The Onsite Storage Building can be expanded, or an additional building constructed, to house the waste until the plant's end-of-life. (2) The Containment can be appropriately prepared for waste storage ( spaces can be cleared, floors can be supported ! l if required, and the building can be sealed), adequate l storage space exists for the waste (collected in the Onsite Storage Facility, the decommissioning waste and the waste generated elsewhere in Maine for 30 years after decommissioning), and the waste can be moved into the I Containment and put in place. (3) The Containment can provide adequate, long-term protection for the stored waste, meeting all regulatory conditions for population exposure. l .f 7-1
* 'In all cases, no major technical impediments were found, and it is concluded that the concept is technically feasible.
Moreover, the concept has several a'ttractive features which merit further study: (1) It " solves" the problem of what to do with Maine low-level waste generated after January 1, 1993 when the Low Level Waste Policy Act turns the responsibility over to the states. (2) It avoids having to create another licensed nuclear facility in Maine, which would result if a low-level waste repository were sited away from the Maine Yankee site. (3) Radiation exposures for waste handling, transportation and storage for workers and the population would be lower than if a repository were located at a different site since most of the waste generated in Maine originates at Maine Yankee. (4) The concept of leaving the Containment Building standing ("in-situ" decommissioning) should result in a lower occupational exposure during decommissioning than the currently favored immediate dismantlement option. (5) Although cost was never quantified in the study, it is I reasonable to expect that onsite storage at Maine Yankee would be less expensive than constructing a facility elsewhere and shipping Maine Yankee waste to it. Also, in-situ decommissioning of the plant should be less expensive than immediate dismantlement (which must decontaminate the Containment, remove all radioactive raterial, a then demolish the very massive building). 1 1 The concept of combining low-level waste management and in-situ i decommissioning is technically feasible. If the concept is to i 1 7-2 l
o' *o
'be advanced further, more exacting licensing feasibility, cost and engineering studies should be conducted.
e i I 7-3 e *}}