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Nuclear Engineering and Design 242 (2012) 433– 444
Contents lists available at SciVerse ScienceDirect
Nuclear Engineering and Design
j ourna l ho me page: www.elsev ier .com/ locate /nucengdes
fficiency analyses of the CANDU spent fuel repository using modified disposalanisters for a deep geological disposal system design
.Y. Leea,∗, D.K. Choa, M.S. Leea, D.H. Kooka, H.J. Choia, J.W. Choia, L.M. Wangb
Korea Atomic Energy Research Institute, 150 Deokjin-dong, Yuseong-gu, Daejeon-city 305-353, Republic of KoreaDepartment of Nuclear Engineering & Radiological Sciences, University of Michigan, 2355 Bonisteel Blvd., Ann Arbor, MI 48105, USA
r t i c l e i n f o
rticle history:eceived 21 December 2010eceived in revised form 16 October 2011ccepted 17 October 2011
a b s t r a c t
Deep geological disposal concept is considered to be the most preferable for isolating high-level radioac-tive waste (HLW), including nuclear spent fuels, from the biosphere in a safe manner. The purpose of deepgeological disposal of HLW is to isolate radioactive waste and to inhibit its release of for a long time, sothat its toxicity does not affect the human beings and the biosphere. One of the most important require-ments of HLW repository design for a deep geological disposal system is to keep the buffer temperaturebelow 100 ◦C in order to maintain the integrity of the engineered barrier system. In this study, a referencedisposal concept for spent nuclear fuels in Korea has been reviewed, and based on this concept, efficientalternative concepts that consider modified CANDU spent fuels disposal canister, were developed. To
meet the thermal requirement of the disposal system, the spacing of the disposal tunnels and that of thedisposal pits for each alternative concept, were drawn following heat transfer analyses. From the resultof the thermal analyses, the disposal efficiency of the alternative concepts was reviewed and the mosteffective concept suggested. The results of these analyses can be used for a deep geological repositorydesign and detailed analyses, based on exact site characteristics data, will reduce the uncertainty of theresults.. Introduction
With the continuing growth of industry, demand for energy islso rising and concern about global warming increases exponen-ially. The role of nuclear energy has gained importance because ofhe high price of crude oil and global warming. In Korea, since therst commercial nuclear power plant at Kori site was initiated in978, there are now 21 operating nuclear power plants: 4 CANDUeactors at Wolsong and 17 PWR reactors at the Kori, Younggwangnd Uljin sites. The electricity from these nuclear power plants pro-ides around 40% of the country’s total electricity and, according toorea’s National Basic Energy Development Plan, this proportionill rise to about 60% in the near future. To implement this plan,
ne of the important factors is the backend fuel cycle, i.e. the safeanagement of high level radioactive waste (including spent fuels)
rom the nuclear power plants.The spent fuels discharged from nuclear reactors are now being
ccumulated at nuclear power plant sites, as a policy decision on
ecycling or direct disposal of spent fuels has not yet been madend is still being discussed in Korea. The long-term R&D programor HLW, including spent fuels disposal technology development,∗ Corresponding author. Tel.: +82 42 868 2071; fax: +82 42 868 8198.E-mail addresses: [email protected], [email protected] (J.Y. Lee).
029-5493/$ – see front matter © 2011 Elsevier B.V. All rights reserved.oi:10.1016/j.nucengdes.2011.10.039
© 2011 Elsevier B.V. All rights reserved.
was launched in 1997. There has been active research on the devel-opment of the Korean HLW Disposal system and a basic referencedisposal concept in deep geological formation for spent fuels hasbeen developed through the R&D program. Recently, research hasbeen performed on the demonstration and the efficiency analy-ses of the disposal system to make it safer and more economical.The purpose of disposal is not only to isolate the high-level waste(HLW) from humans, but also to inhibit leakage of any radioac-tive materials into the accessible environment. The extremely highlevel and long-time scale radioactivity of HLW has led to the ideaof deep repositories in stable geological formations. A deep geo-logical repository concept for the disposal of heat-generating HLWfrom nuclear reactors involves placing the disposal canisters withHLW at a depth ranging between 500 and 1000 m in a plutonicrock mass. Research concerned with the emplacement of HLWin a geological repository has been performed in many countriesaccording to their own nuclear energy policies. Such geological dis-posal remains the preferred option for safe management of HLWin several countries, including Korea (IAEA, 1994, 2009). In thisstudy, a Korea reference disposal concept and the layout of therepository for spent fuels were reviewed. Based on this concept,
alternative disposal concepts, considering modification of the dis-posal canister for the CANDU spent fuels, were drawn to enhancethe disposal efficiency. Spacings of the disposal tunnels and of thedisposal pits for each alternative concept were also drawn up to
4 ring and Design 242 (2012) 433– 444
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34 J.Y. Lee et al. / Nuclear Enginee
eet the thermal requirement of the system, following heat trans-er analyses. From the results of the analyses, the disposal efficiencyf the alternative disposal concepts was reviewed and the most effi-ient concept suggested. The results of study can be used for a deepeological repository design and detailed analyses based on exactite characteristics data will reduce uncertainty of the results.
. Disposal concept for spent fuels
Disposal concepts for HLW and spent fuels are being devel-ped in many countries. Sweden and Finland have developed theirwn disposal concept for the spent fuels in deep geological gran-te formation and have already selected a disposal site (Hanssont al., 2009; Posiva Oy, 2010). France and Japan have completedevelopment of a general concept for HLW and/or spent fuels inoth hard rock (such as granite) and soft rock (such as clay) andre now proceeding towards specifying a location (ANDRA, 2005a,005b; NUMO, 2004). In a number of countries, geological disposal
s the reference management option and developments are movingorward to the site selection stage. In Korea, a reference geolog-cal disposal concept for PWR and CANDU spent fuels has beeneveloped (Lee et al., 2007; Choi et al., 2005) and further researchowards demonstrating the developed technologies and enhancingisposal efficiency is being carried out.
.1. Design requirements
The disposal system for spent nuclear fuels is based on geologicalisposal, where the spent fuels are encapsulated into disposal can-
sters and disposed of in a repository excavated deep into bedrockdeep geological repository). The objective is to isolate the spentuel from the biosphere for as long as it could be harmful to people’sealth and to the environment. The general principle of radiologicalafety is that the repository shall be designed such that radiationoses to humans and the environment shall be kept as low as iseasonably achievable. The estimated total amount of spent fuels,ased on the national basic energy development plan in Korea, isescribed in Table 1.
In order to safely dispose of the spent fuel from the nuclearower plants in a deep geological host rock, it is necessary to encap-ulate it in disposal canisters that can endure the high pressure andorrosive environment of the disposal depth. A disposal canisterherefore consists of an inner vessel and an outer shell. The mate-ial of the inner vessel is carbon steel, to endure both the hydraulicressure and the buffer swelling pressure. The material of the outerhell is copper, to withstand corrosive environment undergroundor a long time (Choi et al., 2005; Lee et al., 2005). The dimensionsor the PWR and CANDU canisters are: outer diameter 1.02 m, totalength 4.83 m and weight approximately 25 t, including the weightf the fuels, as shown in Fig. 1. The capacity of a canister is limitedo 4 assemblies, in the case of PWR spent fuels, or to 297 bundles, inhe case of CANDU spent fuels. The outer dimensions of the CANDUanister and of the PWR canister are same to facilitate handlingith same devises for packaging and placement.
The safety of the repository system must be ensured for bothhort and long time frames, so that, in all conceivable circum-tances, the radionuclides of the disposed waste within a repositoryystem consisting of engineered barriers and natural barriers can
able 1otal amount of spent nuclear fuels to be disposed of in Korea.
Fuel type Amount (tU) Number of assemblies Number of bun
PWR 20,000 45,500
CANDU 16,000 842,000
Total 36,000
Fig. 1. Disposal canister of Korea reference disposal system.
be sufficiently confined by the adequate implementation of multi-ple barriers to the system, individually and/or in combination. Toassure the engineered barrier’s long-term integrity, i.e. to preservethe physical and chemical properties of bentonite, the peak tem-perature of buffer material (bentonite blocks) should be lower than100 ◦C. This is because a transformation of the montmorillonite inthe bentonite into illite or similar minerals would lead to the lossof buffer functions (Svensk Kärnbränslehantering, 2006). From thisconstraint, the thermal analyses to determine the space of disposaltunnels and that of disposal pits should be done on the basis of thedecay heat of the spent fuels in the canisters.
2.2. Multi barrier system
The host rock and the canister, together with a bentonite bufferplaced around the canister and in backfilled disposal tunnels, areconsidered as barriers that will isolate the spent nuclear fuel fromthe biosphere. To guarantee the long-term safety of geologicaldisposal, HLW multi-barrier concepts generally utilize both anengineered barrier system (EBS) and a natural barrier system (NBS).
The natural and engineered barrier components of the disposal sys-tem should be optimized to allow for efficient and safe repositoryoperation and to protect the repository’s structural integrity.dles Capacity of disposal canister Number of disposal canisters
4 assemblies 11,375297 bundles 2835
14,210
J.Y. Lee et al. / Nuclear Engineering and Design 242 (2012) 433– 444 435
system
2
nwtp(af
2
tmtrtcp
Fig. 2. Concept of multi-barrier
.2.1. Natural barrier system (NBS)The rock formations themselves will act as a natural barrier in a
umber of ways that could include limiting the flow of ground-ater or of any gas released from the waste packages, limiting
he movement of radio nuclides towards the ground surface, androtecting the wastes deep underground from extreme changeshuman or natural) that may occur on the Earth’s surface. Based on
nationwide survey (KAERI, 2002), we assume that the host rockor a repository is granite as a geological formation.
.2.2. Engineered barrier system (EBS)The host rock, as a natural barrier system, is expected to isolate
he radioactive materials permanently from the ecosystem and toaintain the safety of the ecosystem from radioactive exposure by
he disposed spent fuels. However, the properties of the natural bar-
iers are inherently variable and uncertain, making them difficulto predict with high confidence. Therefore, an engineered barrieroncept is implemented to establish the reliability from a safetyerspective. The engineered barrier system includes the spent fuelsFig. 3. Concept of disposal pit emplacem
for geological disposal system.
or HLW waste form to be disposed of, the disposal canister, thebuffer material, and the backfill material in the disposal tunnels, asshown in Fig. 2.
Because the buffer blocks surrounding the canister are expectedto maintain the integrity of the canister and to delay any nuclidemigration, its hydraulic conductivity should be low, it shouldadsorb radionuclides, and it should exhibit a high thermal con-ductivity to effectively remove the decay heat, thus keeping thetemperature below 100 ◦C. The thermal requirement of such abuffer is a major criterion for designing the underground repositorylayout (Kim et al., 2003).
2.3. Disposal concept for spent fuels
2.3.1. Disposal concept
In general, there are two methods, known as “in-room” (hori-zontal) emplacement and “disposal pit” (vertical) emplacement, forplacing high level radioactive waste disposal canister in a disposaltunnel. In this study, a vertical emplacement method – as would
ent in deep geological repository.
436 J.Y. Lee et al. / Nuclear Engineering an
Table 2Dimension and number of disposal tunnels.
Item Description
Dimension ofdisposal tunnels
Width 5.00 m, height 6.15 mLength 251 mPWR: 37 pits = >36 × 6 m + *35 m = 251 mCANDU: 55 pits = >54 × 4 m + *35 m = 251 m*35 m is for operation and concrete plug, etc.
baopdat
2
tcbtaiittibcAFoprpPf
2
f
Number ofdisposal tunnels
PWR: 1.05 × 11,375 = 11,944 pits:323 disposal tunnelsCANDU: 1.05 × 2835 = 2977 pits:54 disposal tunnelsTotal: 377 disposal tunnels
e expected in Korea – was considered. The spent fuels packed in disposal canister would be deposited in an array of disposal pitsn the floor of a disposal tunnel. After the placement of the dis-osal canister, the gap between the canister and the wall of theisposal pit would be filled with a compacted bentonite block as
buffer material and the disposal tunnel would be backfilled withhe mixture of crushed rock and bentonite, as shown in Fig. 3.
.3.2. Repository layoutThere are several determining factors that affect the reposi-
ory layout. One important factor is to divide the repository intoontrolled and uncontrolled areas. Canister handling will alwayse performed in the controlled area. Technical rooms in this con-rolled area include four shafts: a canister shaft, a personnel shaftnd two ventilation shafts. Excavation and backfilling will be donen the uncontrolled area. Technical rooms in the uncontrolled areanclude the corresponding access tunnel, the personnel shaft andwo ventilation shafts. The boundary between the controlled andhe uncontrolled areas will move during the lifespan of the repos-tory. Disposal tunnels and panel tunnels will be excavated andackfilled in phases during the operation of the repository, thushanging configurations of the controlled and uncontrolled areas.bove factors have led to the layout that is presented in Table 2 andig. 4. The controlled area is on the left and the uncontrolled arean the right side. Disposal of the canisters starts from the left androceeds to the right. This concept will receive feedback from theesults of the safety assessment. As shown in Fig. 4, there are 7 dis-osal panels and one panel for future use. Of these, 6 panels are forWR spent nuclear fuels and one panel is for CANDU spent nuclearuels, because their heat generation properties are different.
.4. Alternative disposal concepts
Currently, research on the recycling of the PWR spent nuclearuels is being carried out actively to reuse the valuable material in
Fig. 4. Layout concept of a dee
d Design 242 (2012) 433– 444
the spent fuels in Korea. Accordingly, research on the disposal sys-tem for HLW from the recycling process is also being performed.CANDU spent fuels are however considered to be disposed ofdirectly, because they have little possibility of reuse. Furthermore,based on the reference disposal concept of the spent fuels, there areextensive efforts to enhance the disposal efficiency in Korea. As oneof these efforts, alternative concepts of CANDU disposal system arebeing developed by modifying the disposal canister. Such a modi-fied canister uses baskets of 60 bundles for CANDU spent fuels, asillustrated in Fig. 5. According to the canister’s capacity for baskets,4 kinds of modified disposal canisters (CANDU-1B, 2B, 4B, 8B) weredesigned and 4 disposal concepts (DC-1CAN, 2CAN, 4CAN, 8CAN)developed, as shown in Figs. 6 and 7, and Table 3. They all haveto meet the thermal requirement of the disposal system. For thispaper, thermal analyses were performed to design the appropriatedisposal concepts for each kind of disposal canister and the effi-ciency of each disposal concept was reviewed in terms of disposalarea, uranium density, excavation volume and volume of copperand cast iron, etc.
3. Thermal analyses
In developing a disposal concept, deep geological repository lay-outs are determined on the basis of thermal analyses, so as to meetthe design requirement of buffer temperature limit. Japan, whichplans to dispose of its HLW in vitrified waste forms, has classifiedrocks into “hard rock” and “soft rock” and has further suggestedan optimized layout by defining required spacing between the dis-posal tunnels (or pits) to ensure the thermal stability and structuralintegrity, as determined by the maximum bentonite block’s tem-perature (JNC, 1999). Sweden and Finland have performed similarresearch to determine such a spacing to meet the thermal require-ment in the granite rock formation (Hokmark and Claesson, 2005;Kari Ikonen, 2003). Canada, which has a horizontal disposal con-cept in deep geological tunnels, has performed research to establishoptimized conditions, not only by setting up disposal tunnel andcanister spacing requirements according to the cooling time, butalso by analyzing the uranium density, thermal stress per unit vol-ume, etc. (Baumgartner, 2005). In this study, aimed at improvingthe efficiency of the CANDU spent fuels disposal system, severalconcepts were drawn up in accordance with a modified disposalcanister (Figs. 6 and 7, Table 3). Thermal analyses of these werethen carried out to determine the spacings of disposal tunnels and
disposal pits to meet the design requirement that the peak temper-ature of the bentonite block should not exceed 100 ◦C. To comparethe disposal efficiency between the disposal concepts more read-ily, the spacing of the disposal tunnels was fixed at 40 m, and thep geological repository.
J.Y. Lee et al. / Nuclear Engineering and Design 242 (2012) 433– 444 437
Fig. 5. Packaging concept of modified disposal canister for CANDU spent fuels.
fied C
mmcu
3
gwwmata
Fig. 6. Concepts of modi
aximum temperature at 95 ◦C. Based on this fixed spacing andaximum temperature, the spacing of the disposal pit for each
oncept was derived through several thermal analyses based sim-lations, as described below.
.1. Analyses model
A schematic model with calculation domain for a quarter of theeological repository system was illustrated in Fig. 8. This modelas chosen because a number of disposal tunnels and disposal pitsere located parallel to each other, with the same spacing and geo-
etrical shape. As shown in Fig. 8, the analyses’ domain of the uppernd the lower parts in the model was set at a distance of 500 m fromhe repository depth, which was considered far enough not to beffected by the decay heat. Also, the radioactive decay heat source
ANDU disposal canister.
in the disposal pit mirrored the decay heat history by the spentfuels in the disposal canister.
The calculation tool used for these analyses was the Heat Moduleof ABAQUS. This tool is a commercially available computer codeusing a finite element method (Dassault Systems, 2010), and hasalso been reviewed and verified in terms of its applicability for thedesign of the spent fuel disposal system (Andersson, 2005). Oneexample of the finite element models for these thermal analyses isshown in Fig. 9.
3.2. Initial and boundary condition
In the analyses, the initial conditions were set up with theassumptions that the temperature of the groundwater at the sur-face is 15 ◦C, and that the thermal gradient is 3 ◦C for every 100 m
438 J.Y. Lee et al. / Nuclear Engineering and Design 242 (2012) 433– 444
Fig. 7. Alternative disposal concepts (unit: mm).
Table 3Summary of alternative disposal concepts for CANDU spent fuels.
Disposal concept DC-1CAN DC-2CAN DC-4CAN DC-8CAN
Canister CANDU-8B CANDU-4B CANDU-2B CANDU-1BDimension of canister (D × H, unit: m) 1.28 × 4.9 1.28 × 2.66 1.28 × 1.54 1.28 × 0.98Numbers of basket in a canister 8 4 2 1Numbers of canister in a disposal pit 1 2 4 8
ddbgu
Numbers of spent fuel bundles in a disposal pit 480
Dimension of disposal tunnel (W × H, unit: m) 5 × 7.4
Dimension of disposal pit (D × H, unit: m) 2.08 × 7.39
epth (Kim and Bae, 2003). The top and bottom of the analyzed
omain are therefore considered to be constant temperaturesoundaries and set at 15 ◦C and 45 ◦C, respectively, according to theeothermal conditions of the repository. The vertical sides were setp with adiabatic boundary conditions, because of the symmetricFig. 8. Schematic model with calculation
480 480 4805 × 5.16 5 × 4.04 5 × 3.782.08 × 7.89 2.08 × 8.9 2.08 × 10.92
model for the thermal analysis of the repository. The density, ther-
mal conductivity and specific heat of the rock, buffer, backfill andcanister are shown in Table 4.Once the spent fuels have been disposed of in the reposi-tory, a transient heat diffusion phenomenon arises. Heat transfer
domain for the thermal analyses.
J.Y. Lee et al. / Nuclear Engineering and Design 242 (2012) 433– 444 439
Fig. 9. Thermal analyses
Table 4Material properties for the analyses.
Items Density (kg/m3) Thermal conductivity(W/m ◦C = J/s/m ◦C)
Specific heat (J/kg ◦C)
Canister 7200 52 504
tahb
Buffer 1970 1.0 1380Backfill 2270 2 1190Rock 2650 3.2 815
hrough the waste form, its container, the surrounding materi-ls and the rock will be mainly by conduction. That is becauseeat transfer by radiation and convection will be negligible afterackfilling.
Fig. 10. Decay heat based on the Korean refe
model: DC-2CAN.
The heat transfer is represented with the following 3-dimensional heat conduction equation (Sizgek, 2005).
∂
∂t(�CpT) = ∂
∂x
[k
∂T
∂x
]+ ∂
∂y
[k
∂T
∂y
]+ ∂
∂z
[k
∂T
∂z
]+ q(t) (1)
where T is the temperature [◦C], t is the time [s], � is the density[kg/m3], Cp is the specific heat [J/kg ◦C], k is the thermal conduc-tivity [W/m ◦C], q(t) is the time-dependent volumetric heat source
[W/m3].The decay heat is calculated using ORIGIN-ARP program basedon the Korean reference CANDU spent fuels for disposal as shown inFig. 10. The regression equation for the heat source from the spent
rence CANDU spent fuels for disposal.
440 J.Y. Lee et al. / Nuclear Engineering and Design 242 (2012) 433– 444
Table 5Thermal regression constants.
Cooling time (year) C1 A1 C2 A2 C3 A3 y0
.4 0.71648 0 0 47.2355 0.00154 218.60 0.02183 0.71
f2
p
wdn
4
4
tTasdatta
4
tmsadaaatp
Fig. 11. Concept of a unit disposal area.
Table 6Results of thermal analyses.
Disposal concepts DC-1CAN DC-2CAN DC-4CAN DC-8CAN
Disposal tunnels spacing (m) 40.0 40.0 40.0 40.0Maximum temp. (◦C) 95 95 95 95
1 ≤ t ≤ 100 251.12 0.02442 5431100 < t ≤ 1E + 6 11.50 5.96E-05 33.2
uels is determined as follows (Choi and Kang, 1999; Cho and Lee,008):
(t) = C1 EXP (−A1t) + C2 EXP (−A2t) + C3 EXP (−A3t) + y0 (2)
here p(t) in W/tU is decay heat generated from spent fuels in aisposal canister and t is the time (in years) after discharging fromuclear reactor. The coefficients are described in Table 5.
. Results and discussion
.1. Unit disposal area
It is necessary to define a unit disposal area to analyze and designhe repository layout based on the results of these thermal analyses.his unit area is defined as the area between the disposal tunnelsnd the disposal pits, as shown in Fig. 11. Therefore, the overallcale of the repository can be estimated by the product of the unitisposal area and the number of disposal canisters emplaced. Fromn economics point of view, it is desirable to design the disposalunnel spacing and the disposal pit spacing such that they minimizehe total repository area (i.e., the disposal density and heat densityre maximized), constrained by the thermal requirement.
.2. Results of the thermal analyses
Thermal analyses were performed to determine the disposalunnel and disposal pit spacings satisfying the thermal require-
ents of the bentonite blocks for each disposal concept. Figs. 12–15how the results of the thermal analyses for the bentonite blocksnd the temperature history of each concept. The spacing of theisposal tunnels and that of the disposal pits, maximum temper-ture and the time taken to reach the maximum according to the
nalyses results are summarized in Table 6. As shown in the tablend figure, from the fixed disposal tunnel spacing of 40 m and fixedemperature of 95 ◦C, the spacings of the disposal pits for each dis-osal concept were 5.3, 4.9, 4.6 and 4.3 m and the times to reach theFig. 12. Result of thermal analyses f
Disposal pit spacing (m) 5.3 4.9 4.6 4.3Year to reach max. temp. (years) 19.1 23.1 27.0 33.5
maximum temperature were 19, 23, 27 and 33 years, respectively,after emplacement of the disposal canister.
4.3. Reviews and discussions
There are several factors in estimating disposal efficiency. Inthis study, factors such as unit disposal area, uranium density,
or disposal concept DC-1CAN.
J.Y. Lee et al. / Nuclear Engineering and Design 242 (2012) 433– 444 441
Fig. 13. Result of thermal analyses for disposal concept DC-2CAN.
Fig. 14. Result of thermal analyses for disposal concept DC-4CAN.
Fig. 15. Result of thermal analyses for disposal concept DC-8CAN.
442 J.Y. Lee et al. / Nuclear Engineering and Design 242 (2012) 433– 444
Table 7Efficiency analyses between alternative disposal concepts.
Dispo sal concepts DC-1CAN DC- 2CAN DC- 4CAN DC- 8CAN
Unit Disposal Area 212.0 196.0 184.0 172.0Reduced area 16.0 12.0 12.0
U-Densit y(kU/m2) 43.0 46.5 49.5 53.0Increased density 3.5 3.0 3.5
Heat density ( W/m2) 7.2 7.8 8.3 8.9Increased density 0.6 0.5 0.6Cast Iron/UDA 2.0 2.4 3.3 5.1
Increased volume 0.4 0.9 1.8Copper/UDA 0.2 0.3 0.5 0.9
Increased volume 0.1 0.2 0.4Bento. Vol./UDA 18.7 19.9 22.3 27.0Reduced volume 1.2 2.3 4.7
Backfill/UDA 196.1 126.4 92.9 81.2Reduced volume 69.7 33.5 11.7Excavation Vol. 221.1 153.2 123.1 118.3Reduced volume 67.9 30.1 4.8D. T Vol./ UDA 196.1 126.4 92.9 81.2Reduced volume 69.7 33.5 11.7
6.9
haoastatiotptdeTt
Fc
D. H Vol./UDA 25.0 26.7 30.2 Increased volume 1.7 3.4
eat density, material (bentonite, copper and cast iron) volumend excavation volume, etc. were considered to determine theptimized disposal concept. As shown in Fig. 7 and Table 3 forlternative disposal concepts, they have same disposal capacity forpent fuel bundles in a unit disposal area. But, as the number ofhe disposal canisters in a disposal pit increased, the unit disposalrea, backfill material and excavation volumes were reduced andhe U-density, heat density, disposal canister material (such as castron and copper), and bentonite volume were increased. In termsf the disposal efficiency, as the unit disposal area become smaller,he U-density and the heat density become higher, and the dis-osal concept is then considered to be more efficient. Based on thehermal analyses, the disposal efficiencies of unit disposal area, U-
ensity, heat density, volume of the increasing cast iron and copper,xcavation volume, etc. was reviewed. They were summarized inable 7 and Figs. 16–21. For the alternative concepts, the smallerhe disposal area, the less the volume of canister material increase,ig. 16. Unit disposal area and the reduced area comparison between the disposaloncepts.
37.0
and the more the excavation volume decreases, the disposal sys-tem can be more efficient. The table and the figures show that thedisposal area was decreasing from the concept of DC-1CAN to thatof DC-8CAN. That means that the disposal efficiency was increas-ing. But, the volumes of canister material and the volumes of buffermaterial were increasing and the excavation volume was decreas-ing. So, to determine the best option from the alternative options,it was necessary to review in detail the factors between the alter-native disposal concepts. The reduced value or increased value ofthe factors between the alternative disposal concepts is shown in
, , columns of the Table 7 and triangle chart in each figures.As shown in , , of the Table 7 and in the triangle chart of thefigures, the reductions of unit disposal areas from DC-1CAN to DC-
2CAN, from DC-2CAN to DC-4CAN, and from DC-4AN to DC-8CANwere 16, 12, 12 m2, respectively. Also, the increased U-density andthe heat density were 3.5, 3.0, 3.5 m3 and 0.6, 0.5, 0.6 m3, respec-tively. The increased volume of cast iron and copper for disposalFig. 17. U-density and the increased density comparison between the disposal con-cepts.
J.Y. Lee et al. / Nuclear Engineering and Design 242 (2012) 433– 444 443
Fig. 18. Cast iron volume and the increased volume comparison between the dis-posal concepts.
Fc
citr
Fp
ig. 19. Copper volume and the increased volume comparison between the disposaloncepts.
anister were 0.4, 0.9, 1.8 m3 and 0.1, 0.2, 0.4 m3, respectively. Thencreased volume of bentonite and the reduced volume of excava-
ion for the concepts were 1.2, 2.3, 4.7 m3 and 67.9, 30.0, 4.8 m3,espectively.ig. 20. Bentonite volume and the increased volume comparison between the dis-osal concepts.
Fig. 21. Excavation volume and the reduced volume comparison between the dis-posal concepts.
Therefore, based on these reviews and considering several fac-tors such as disposal area, heat density, volume of material, volumeof excavation, etc. for the effectiveness between disposal concepts,the concept of DC-2CAN had more reduced unit disposal area andexcavation volume and less increased material volume such as castiron, copper and bentonite. So, it was revealed that disposal conceptof DC-2CAN was the most effective disposal concept. Furthermore,the time to reach the maximum temperature and the temperaturechanges are shown as a function of the disposal canister. With theDC-2CAN concept, which had the 2 disposal canister in a disposalpit, the time to reach the maximum temperature was approxi-mately 23 years, and longer than that of the DC-1CAN concept,which was 19 years. A DC-8CAN concept shows the latest maxi-mum temperature, which was about 33 years after emplacement.These results mean that an impact to the structure of the host rockshould be considered with regard to the concepts.
5. Conclusion and future works
In this study, as one of the efforts to enhance the disposal effi-ciency for a deep geological disposal system, a Korea referencedisposal concept and the layout of the repository for spent fuelswere reviewed. Based on this concept, alternative disposal conceptsusing modified disposal canisters for CANDU spent nuclear fuelswere drawn up to enhance the disposal efficiency. A spacing of thedisposal tunnels and of the disposal pits for each alternative dis-posal concept was drawn up to meet the thermal requirements ofthe overall disposal system, following heat transfer analyses. Fromthe results of the analyses, the disposal efficiency of the alternativeconcepts was reviewed and the most effective concept suggested.The following conclusions may be drawn:
• To enhance the disposal efficiency of the deep geological disposalsystem for spent nuclear fuels, a Korea reference disposal sys-tem for PWR and CANDU spent nuclear fuels have been reviewed.Modified disposal canisters for CANDU spent nuclear fuels wereintroduced and alternative disposal concepts have been devel-oped according to these modified disposal canisters. Also, thespacings of the disposal tunnels and those of disposal pits to meetthe design requirement for each concept have been determinedthrough thermal analyses.
• The alternative disposal concepts have been reviewed to beestimated with regard to the disposal efficiency of the deep geo-logical repository system. As the number of disposal canisters
4 ring an
--
-
-
•
•
dut
A
PbK
R
A
A
A
44 J.Y. Lee et al. / Nuclear Enginee
in a disposal pit increased from the DC-1CAN concept to theDC-8CAN concept:
the unit disposal area was reducing; the uranium density and heat density of the unit disposal areawere increasing;
the volumes of the backfill and the disposal tunnel excavationwere reducing; and
the volumes of cast iron, copper, bentonite, and the disposal pitexcavation were increasing.
Based on the efficiency analyses using the estimation factorsbetween the disposal concepts, it has been revealed that DC-2CAN was the most preferable disposal concept.Also, as the number of disposal canisters in a disposal pitincreased, the time of the peak temperature was delayed. Theseshould be considered and reviewed in terms of structural stabilityof the underground facilities.
The results from this research are expected to be used in theesign of a HLW underground repository. More detailed analyses,sing site-specific characterization data, are nevertheless requiredo minimize any uncertainties.
cknowledgement
This work was supported by Nuclear Research and Developmentrogram of the National Research Foundation (NRF) grant fundedy the Ministry of Education, Science and Technology (MEST) inOREA.
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