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Symposium on Nuclear DataTokyoTech, Nov. 29-30, 2018
Geological Disposal of High-Level Radioactive Waste:
Long-term Safety and Reduction of Environmental Impact
Hidekazu ASANO
Radioactive Waste Management Funding and Research Center(RWMC)
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Key words
Geological Disposal
Long-term Safety
Radionuclide Migration
Radiation Dose vs. Radiological Toxicity
Environmental Impact
Viewpoints
Data needs from new fields
Topics/GD1. Concept “Isolation & Containment”2. Post closure & long-term safety 3. Load reduction4. Integrated approach5. Data required 6. Summary & Needs
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High-level
Liquide waste
Second progress report on research and development for TRU waste Disposal in Japan – Repository design, safety assessment and means of
implementation in the generic phase-, JAEA-Review 2007-010, FEPC TRU-TR2-2007-01, March 2007
Processes and generation of radioactive waste at reprocessing plant
Se-79Zr-93Tc-99Pd-107Sn-126Cs-135
I-129/AgI
C-14
TRU waste
TRU waste HLW: vitrified waste
*
*
1. Concept “Isolation & Containment”(1/9)
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H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second Progress Report on Research and Development for the
Geological Disposal of HLW in Japan,
Japan Nuclear Cycle Development Institute (JNC), 2000.
- TN1410 2000-004, Supporting Report 3, Safety Assessment of the Geological Disposal System
Process conditions for vitrified waste(Glass)
*
*
1. Concept “Isolation & Containment”(2/9)
Radioactivity of vitrified waste vs. Time after disposal*
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* H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second
Progress Report on Research and Development for the Geological Disposal of HLW in Japan, Japan
Nuclear Cycle Development Institute (JNC), 2000.
TN1410 2000-003, Supporting Report 2, Repository Design and Engineering Technology
Systems for nuclide migration
1. Concept “Isolation & Containment”(3/9)
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* H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second Progress Report on Research
and Development for the Geological Disposal of HLW in Japan, Japan Nuclear Cycle Development Institute (JNC), 2000.
TN1410 2000-003, Supporting Report 2, Repository Design and Engineering Technology
1. Concept “Isolation & Containment”(4/9)
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HLW(Vitrified waste)Result : safety assessment/long-term & post-closure safetyRadiation dose caused by groundwater usage by a person living on the surface,Estimation modeled as 40,000 - vitrified wastes disposed at 1,000m below the surface, and contact with groundwater after 1,000 years.
Figure from NUMO-TR-04-01, 2004
Original Data from H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second Progress Report on Research
and Development for the Geological Disposal of HLW in Japan, Japan Nuclear Cycle Development Institute (JNC), 2000.
- TN1410 2000-001, Project Overview Report
- TN1410 2000-004, Supporting Report 3, Safety Assessment of the Geological Disposal System
1. Concept “Isolation & Containment”(5/9)
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Barrier performance- Containment capability of each barrier component by multi-barrier system
Np-237 Cs-135
Nuclide inventory in the EBS components and Geosphere
Np*, low solubility & large distribution coefficient- 107 years, approx. 1 % of Np-237 as of the initial
inventory moves to geosphere from EBS
- approx. 1/105 of the original Np-237 reaches to the biosphere
(*T1/2 =2.14×106 years)
Cs**, highly soluble & less sorptive- approx. 95% of the initial inventory of Cs-135 is
isolated at the area of EBS & geosphere
(**T1/2 =2.30×106 years)
1. Concept “Isolation & Containment”(7/9)
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1. Isolation
隔離
2. Containment
閉じ込め
c. Buffer material/緩衝材b. Overpack/オーバーパックa. Vitrified waste/ガラス固化体(Bq/kg)
d. Geological formation/host rock 地質/岩盤
e. Biosphere 生物圏f. Human being 人間活動
Engineered Barrier人工バリア
Multi barrier system多重バリアシステム(a,b,c,d)
Natural barrier天然バリア
10.Radiation exposure被ばく線量評価 : Sv/y
9. Dose conversion factor:/DCF線量換算係数 : Sv/Bq
8. Water utilization : Ingestion河川水利用 : 経口摂取
7. Dilution 希釈
6. Advection, Dispersion移流、分散:透水(量)係数
5. Sorption:distribution coefficient収着:分配係数
4. Diffusion : diffusion coefficient拡散:拡散係数
3. Precipitation : solubility沈殿生成:溶解度
2. Dissolution : solubility溶解:溶解度
1. Contact with groundwater地下水との接触
HLW高レベル放射性廃棄物
Partitioning/核種分離 &
Transmutation/核変換高速炉、ADS
Reduction of waste volume
and harmfulness減容、有害度低減
on the ground
Radionuclide migration核種移行
Temporal & spatial
uncertainty
時間的、空間的な不確実性
1. Concept “Isolation & Containment”(9/9)
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SSR-5 Specific Safety Requirement/Disposal of Radioactive waste, IAEA 20111.10.The specific aims of disposal are:(a) To contain the waste;(b) To isolate the waste from the accessible biosphere and to reduce substantially the
likelihood of, and all possible consequences of, inadvertent human intrusion into the waste;(c) To inhibit, reduce and delay the migration of radionuclides at any time from the waste to
the accessible biosphere;(d) To ensure that the amounts of radionuclides reaching the accessible biosphere due to any
migration from the disposal facility are such that possible radiological consequences are acceptably low at all times.
1.12. Disposal facilities are not expected to provide complete containment and isolation of waste over all time; this is neither practicable nor necessitated by the hazard associated with waste, which declines with time.
Concepts relating to disposal of radioactive waste, Important !!
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SSR-5 Specific Safety Requirement/Disposal of Radioactive waste, IAEA 2011RADIATION PROTECTION IN THE POST-CLOSURE PERIOD2.15Safety objectiveThe safety objective is to site, design, construct, operate and close a disposal facility so that protection after its closure is optimized, social and economic factors being taken into account. A reasonable assurance also has to be provided that doses and risks to members of the public in the long term will not exceed the dose constraints or risk constraints that were used as design criteria.
Criteria(a) The dose limit for members of the public for doses from all planned exposure situations
is an effective dose of 1 mSv in a year [3].(b) To comply with this dose limit, a disposal facility (considered as a single source) is so
designed that the calculated dose or risk to the representative person who might be exposed in the future as a result of possible natural processes affecting the disposal facility does not exceed a dose constraint of 0.3 mSv in a year or a risk constraint of the order of 10–5 per year.
(c) …..inadvertent human intrusion after closure,….. less than 1 mSv….. are not warranted.(d)….. human intrusion….. annual dose of more than 20 mSv….. options for waste disposal
are to be considered.(e)…..annual doses in the range 1–20 mSv….. reasonable efforts are warranted.(f) Similar considerations apply….. for deterministic effects in organs may be exceeded.
2. Post-closure & long-term safety
NuclideT1/2(year)
DCF(μSv/kBq)
Contents(Kg/tonSNF)
U-235 0.7 Billion 47 10kg
U-238 4.5 Billion 45 930kg
Pu-238 87.7 230 0.3kg
Pu-239 24,000 250 6
Pu-240 6,564 250 3
Pu-241 14.3 4.8 1
Np-237 2.14×106 110 0.6
Am-241 432 200 0.4
Am-243 7,370 200 0.2
Cm-244 18.1 120 0.06
Se-79 2.95×105 2.9 0.006
Sr-90 28.8 28 0.6
Zr-93 1.53×106 1.1 1
Tc-99 2.11×105 0.64 1
Pd-107 6.50×106 0.037 0.3
Sn-126 1×105 4.7 0.03
I-129 1.57×107 110 0.2
Cs-135 2.30×106 2.0 0.5
Cs-137 30.1 13 1.5
Rad
ioto
xici
ty/I
nge
stio
n
Time(year)
SNF, Nuclides & RadiotoxicitySNFGlass/HLWwith P&TNatural-U(9tons, incld. Daughter nuclide)
Time vs. Radiological Toxicity/Ingestion/SNF-1tHM(UO2/PWR, 5years, 45,000MWD/t, U&Pu/Separation/99.5%, MA/Separation/99.5%)
From, Text ”Nuclear Fuel Cycle, 8-1 Roles of Partitioning & Transmutation”, K. Tsujimoto, Reprocessing and Recycle Technology Div.,
Atomic Energy Society of Japan, 12
3. Load reduction-1
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Disposal area⇒ Waste occupied area ⇒ Mechanical stability of disposal tunnel⇒ Temperature/buffer material<100℃
Disposal tunnel to
tunnel distance xD [m]
Pitch between
waste y [m]
Waste occupied
area [m2]
Vertical
emplacement10 4.44 44.4
Remove the heat generating nuclides
Heat generation rate of the vitrified waste
Radionuclide content of the waste
Reduction of foot-print of the repository
* H12: Project to Establish the Scientific and Technical Basis for HLW Disposal in Japan, Second Progress Report on Research and Development for the
Geological Disposal of HLW in Japan, Japan Nuclear Cycle Development Institute (JNC), 2000.
TN1410 2000-003, Supporting Report 2, Repository Design and Engineering Technology
3. Load reduction-2
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Reactor FuelSpent
fuelReprocessing Vitrification
Vitrified
waste
Geological
disposal
-UO2/
MOX
Burn-
up
Cooling
period
Separation
process
Separation
ratioNuclides
Glass
Matrix
Melter
operation
Waste
loading
Storage
period
Waste
occupied
area
LWR UO2
45
GWD
/THM
4 Purex 99.5 U, Pu - -Approx.
20wt%
50
years44m2/glass
LWR,
FR,
etc.,
UO2
MOX/
Pu
thermal
MOX/
full
Low
~High
>4
years
Nuclides
and their
separation
ratio
Request
from
geological
disposal
MA:
Np,
Am,
Cm
Measure -
Higher
waste
loading
Heat
generation
Repository
area
Waste
occupied
area
Waste
emplace-
ment
method
(V,H)
Cs/Sr - -Heat
generation
Mo Measure Yellow
phase-
PGM:
Ru, Rh,
Pd
MeasureSedimen-
tation-
Case Item Unit Parameters(example) Notation
1 Fuel type - UO2, MOX separate UO2 &MOX
2 Burn-up GWd/THM 28, 33, 45, 55, 70 influence by BU difference is negligible
3 Cooling period/spent fuel Year 4, 10, 20, 30 classification by color coding
4Nuclide separation ratio
Cs,Sr % 0, 70, 90 classification by symbol
5 MA % 0, 70, 90 same as above
6 Mo,PGM % 0, 70 same as above
7 Waste loading ratio wt%/glass 20, 30, 35Value of CAERA index*
8 Waste occupied are m2/glass 44 ~ 300
4. Integrated approach(1/3) : Reference & Variation
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Case
SNFCooling period[year]
Cs/Srseparation
[wt%]
MAseparation
[wt%]
Mo/PGMseparation
[wt%]
Vitrified waste, waste loading
[wt%]
CAERA*[kg/m2]
Reduction of waste occupied
area [%]
1 4 90 0 70 35 2.25 43
2 15 70 0 70 25 1.35 72
3 20 70 0 70 25 1.15 84
4 30 0 0 0 21 0.97 100
5 40 0 0 0 21 0.97 100
6 50 0 90 70 35 2.25 43
7 100 0 70 70 35 2.25 43
Result : reduction of waste occupied area/hard rock, -1,000m, vertical emplacement
UO2
CAERA(kg/m2) = Waste loading ratio(wt%/glass) ー Na2O content ratio(wt%/glass)
Waste occupied area(m2/glass)× Glass weight(kg ) ×
1
100
Maximum temperature of buffer material
*CAERA : Comprehensive Analysis of Effects on Reduction of disposal Area(包括的な検討による廃棄体専有面積削減効果)
Results from RWMC’s own research program entitled “ Study on the effects of advanced nuclear fuel cycle technology to the geological disposal concept, Fy.2017”
4. Integrated approach(2/3)
5.1 General Properties of Spent Nuclear Fuel
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(1) Calculation code & nuclear dataTo estimate isotope composition of spent nuclear fuel(SNF) for evaluation of;
- amount of nuclear fuel materials- management of SNF- fuel cycle properties incl. various chemical/physical processes - safety assessment for radioactive waste generated from reprocessing of SNF
(2) Code & Data/example- ORIGEN-2.0, -2.1, -2.2, -ALP : ORNL Isotope Generation and Depletion Code - ORLIBJ40 : cross section data library- ENSDF : nuclide decay library
(3) Reactor core, fuel type and fuel burn-up/UO2*
Burn-up GWd/THM 28 45 70
Power ratio MW/THM 38
Operation Period
Day 736.84 1184.21 1842.11
Enrichment/U-235
Wt% 2.6 4.5 6.5
Cross section library
-PWR34J40(3.4wt%)
PWR41J40(4.1wr%)
PWR47J40(4.7wt%)
burn-up and cross section data/UO2 fuel
U-232 0.0001 μg/gU
U-234 10×103 μg/gU-235
U-236 250 μg/gU
Tc-99 0.01 μg/gU
Impurities of U-isotopes in UO2 fuel
5. Data required(1/6)
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(4) Reactor core, fuel type and fuel burn-up/MOX*
Burn-up GWd/THM 28 45 70
Power ratio MW/THM 38
Operation Period
Day 736.84 1184.21 1842.11
Enrichment/Pu
Wt% 7 12 17
Cross section library
-PWRM0210J40(10wt%)
PWRM0213J40(13wr%)
PWRM0213J40(13wt%)
burn-up and cross section data/MOX fuel
Burn-upPu-enrichment
Nuclide composition (wt%)
GWd/THM wt% P u - 2 3 8 Pu- 2 3 9 P u - 2 4 0 P u - 2 4 1 Pu- 2 4 2 A m - 2 4 1 U - 2 3 5 U - 2 3 8
28 7 0 . 1 4 7 3 . 8 1 5 1 . 7 5 0 . 6 5 1 0 . 4 4 8 0 . 1 8 9 0 . 1 6 6 9 2 . 8 3 4
45 12 0 . 2 5 2 6 . 5 4 3 . 0 0 1 . 1 1 6 0 . 7 6 8 0 . 3 2 4 0 . 1 6 6 8 7 . 8 3 4
70 17 0 . 3 5 7 9 . 2 6 5 4 . 2 5 1 . 5 8 1 1 . 0 8 8 0 . 4 5 9 0 . 1 6 6 8 2 . 8 3 4
Chemical composition/MOX fuel
*UO2 & MOX fuel data/settled on TokyoTech/RWMC joint research program for nuclear fuel cycle and geological disposal, Fy.2017
5. Data required(2/6)
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(5) Heat generation /nuclides/SNF/UO2
70GWd/THM
Results from RWMC’s own research program entitled “ Study on the effects of advanced nuclear fuel cycle technology to the geological disposal concept, Fy.2017”
5. Data required(3/6)
SNF cooling period after discharge(year)
Hea
t ge
ne
rati
on
(kW
/TH
M)
SNF cooling period after discharge(year)
45GWd/THMSNF cooling period after discharge(year)
SNF cooling period after discharge(year)
Co
ntr
ibu
tio
n o
f n
ucl
ides
C
on
trib
uti
on
of
nu
clid
es
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Effects of Nuclide Separation on Repository Layout Removal of heat generation elements Increase waste loading of vitrified waste Layout : Pitch, Distance and Area
Repository layout, exampleCrystalline rock, -1,000m
Vertical emplacement
Time vs. Heat generation rate of vitrified waste
0
50
100
150
200
250
300
350
0.001 0.01 0.1 1 10 100 1000
ガラス固化体の発熱量
, W
/本
処分後の時間, 年
4年
20年
50年
100年
燃焼度:45GWd/t
Heat generation rate/Vitrified waste/without MA SeparationVitrified waste
- PWR- UO2- Burn-up : 45GWd/t- SNF cooling period : 4, 20, 50 & 100 years
Time after Disposal, year
0
20
40
60
80
100
120
140
160
0.001 0.01 0.1 1 10 100 1000
温度, 度
時間, 年
P313 P400
P500 P600
燃焼度:45GWd/t
冷却期間:4年
坑道離間距離:3D
0
20
40
60
80
100
120
140
160
0.001 0.01 0.1 1 10 100 1000
温度
, 度
時間, 年
P313 P400
P500 P600
P640
燃焼度:45GWd/t
冷却期間:50年
坑道離間距離:3D
Time vs. Temperature, at Overpack/Buffer material interface
SNF cooling period: 4 years
Distance : 3D
SNF cooling period: 50 years
Distance : 3D
Time after Disposal, year
Tem
per
ature
, ℃
Tem
per
ature
, ℃
Tem
per
ature
, ℃
Time after Disposal, year
Hea
t gen
erat
ion
rat
e o
f v
itri
fied
was
te,
W/g
lass
un
it
Results from RWMC’s own research program entitled “ Study on the effects of advanced nuclear fuel cycle technology to the geological disposal concept, Fy.2016”
5. Data required(4/6)
20
0
50
100
150
200
250
300
350
0.001 0.01 0.1 1 10 100 1000
ガラス固化体の発熱量
, W
/本
処分後の時間, 年
0%
50%
70%
90%
燃焼度:45GWd/t
冷却期間:50年
No MA Separation : Pitch = 6.4 m50% MA Separation : Pitch = 3.2 m
50% reducing the repository area
0
20
40
60
80
100
120
140
160
0.001 0.01 0.1 1 10 100 1000
温度
, 度
時間, 年
0%
50%
70%
90%
燃焼度:45GWd/t
冷却期間:50年
坑道離間距離:3D
廃棄体ピッチ:3.2m
Vitrified waste- PWR- UO2- Burn-up : 45GWd/t- SNF cooling period : 50 years- Nuclide separation : Np, Am & Cm - Separation ratio : 0, 50, 70 & 90%
Hea
t g
ener
atio
n r
ate
of
vit
rifi
ed w
aste
,
W/g
lass
un
it
Time after Disposal, year
Repository- Crystalline rock- 1,000m below the surface- Horizontal emplacement
Time vs. Temperature at Overpack/Buffer material interface
Burn-up : 45 GWd/t
Cooling period: 50 yearsBurn-up : 45 GWd/t
Cooling period: 50 years
Distance : 3D
Pitch : 3.2m
Time after Disposal, year
Tem
per
atu
re, ℃
Heat generation rate/Vitrified waste/with MA Separation
Results from RWMC’s own research program entitled “ Study on the effects of advanced nuclear fuel cycle technology to the geological disposal concept, Fy.2016”
5. Data required(5/6)
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MA recycle(100 & 99.9%) by large-scale FR/MOX,/equilibrium phase/discharged fuel composition- Case 1 : Recycle/Np, Am, Cm - Case 2 : Recycle/Np, Am- Case 3 : Recycle/Np- Case 4 : no recycle
*Results from Hokkaido Univ./RWMC joint research program for MA transmutation by fast reactor, Fy.2017Calculation by G. Chiba, Hokkaido Univ.
5. Data required(6/6)
100% :Separation/recycle
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6. Summary & Needs
General & Current status/Geological Disposal - Implementers published their own safety case report, incl. Japan- Long-term safety/concept of GD/Isolation & containment is proven- Further improvement
- Site specific- Uncertainty, time & space
View points/Nuclear energy use - GD : Reference vs. Variation- Radionuclide : Migration vs. Existing - Long-term safety : Dose vs. Toxicity- Environmental Impact : Definition (Radiation, amount/RW, area/Repository)
Needs1. General property of SNF : BU & Fuel/chemical composition 2. Separation/Partitioning : MA, Am, FPs 3. Recycle : MA, Am, etc.,4. Vitrification : Quality of glass & Glass-melter operation5. FR : Fuel cycle and discharged wastes6. EI : Environmental load index
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Thank you for your attention.