IRID Annual Symposium 2014

Research and Development on Processingand Disposal of Radioactive Waste
July 18, 2014
International Research Institute for Nuclear Decommissioning
The contents of this presentation include the results of “Establishment of basic technology for decommissioning and safety of nuclear reactors for power
generation in 2015 (technological study and research concerning forming an idea for processing and disposing of radioactive waste resulting from the accident)“,
a project commissioned by the Ministry of Economy, Trade and Industry, and the 2014 subsidiary for decommissioning and contaminated water measures(development of technologies for processing and disposing of waste resulting from the accident).
Unauthorized duplication and transcription prohibited Technology Research Union International Research Institute for Nuclear Decommissioning©International Research Institute for Nuclear Decommissioning

Table of Contents
1. Overview of radioactive waste
management
2. Waste resulting from accident at
Fukushima Daiichi nuclear power station
3. Outline of research into processing anddisposal of radioactive waste
3-1 Characterization
3-2 Long-term storage
4. Issues and measures©International Research Institute for Nuclear Decommissioning21. Overview of radioactive waste management(1) Mid- and long-term roadmap for radioactive waste management
Generation of waste → Storage & management → Processing (Conditioning) → Disposal
Depends on decommissioning
scenario

Segregation, volume
reduction, long-term storage

Needs research & development based on properties of waste

2017
Storage

2021

Reduction of waste generated

Continuing safe storage

Optimization of storage

Processi
ng/disposal

Establishment of basic policy

Research &development

After
recovering
fuel debris
ＨＰ

Planning
decommissioning
scenario

Installation ofwaste
packagingequipment

Confirming prospect on safety
HP SW-2

HP SW-1

Investigating legislation

HP ND-1

DecommissioningAfter 20-25 After 40 years
years

HPSW-3

Waste packaging,transportation& disposal

HP SW-4

Legislation
HP ND-3

Gaining
Consensus
on decommissioning
scenario

HP ND-2

Design/production ofequipment & facility

Decommissioningwork

R&D on decommissioning technology

Point of decision as to whether to proceed to the next process. Additional research and development and review of schedule andwork will also be discussed and decided.

HP SW-1: Making basic policy for processing & disposing ofsolid waste (2017)
HP SW-2: Confirmation of prospect on safety for processing& disposing of solid waste (2021)
HP SW-3: Determination of specifications and production
method of waste package (third term)
HP SW-4: Prospect of installing waste package production
equipment and disposal of waste (third term)

HP ND-1: Planning decommissioning scenario (2015)
HP ND-2: Determination of techniques for decontaminationand dismantling equipment (third term)
HP ND-3: Prospect of waste disposal and end of necessary
research and development (third term)©International Research Institute for Nuclear Decommissioning31. Overview of radioactive waste management(2) System for implementing research and development
Agency of Natural Energy and Resources, Ministry of Trade,
Economy and Industry
Fund management
corporation2013 subsidy for decommissioning &countermeasures against contaminated water(for development of technologies for processing
and disposal of waste)

International Research Institute for Nuclear Decommissioning (IRID)
Japan Atomic Energy Agency (JAEA)

Establishment ofcooperative relation
 Universities, manufacturers, relatedorganizations, etc., at home andabroad‐ Information sharing‐ Conducting joint research, etc.

Tokyo Electric Power Co. (Tepco)

Collecting various information andopinions from home and abroad Academia, international organizations,etc., at home and abroad‐ Setting up task force in Atomic Energy
Society of Japan‐ Setting up expert council at OECD/NEA©International Research Institute for Nuclear Decommissioning42. Waste resulting from accident at Fukushima Daiichi nuclear power station(1) Features of waste

Generated by accident in an uncontrollable stateContamination stemming from reactor fuel of units 1 to 3*
Estimating the quantity is difficult because decommissioning work changes depending on the situation.
Data (especially on composition of long-lived nuclides) is very limited because contamination stretches over awide area and there are locations of high dose

Scatter/dispersion

Water treatment equipment[Debris/felled trees, etc.]

DebrisFallentrees

Contaminated
water

Soil

○ Large quantity and spread overwide area○ Insufficient past employment ofprocessing and disposing of trees
and soil○ Mainly surface contamination by
scattering and dispersion, partly by
seeping contaminated water[Secondary waste stemming fromtreatment of contaminated water]

Secondary waste[Fuel debris/waste fromdismantling]○ Large quantity and high dose
○ Access is difficult at present andcollecting raw waste is difficult.

Pipes forreplacement, reservoir,etc.

○ Insufficient past employment ofprocessing and disposal○ Collecting raw waste is difficult.
○ Quantity of some waste and nuclides
can be estimated depending on thefeature of equipment.

*: Contamination due to activation products and waste generated fromoperation may be included.
©International Research Institute for Nuclear Decommissioning5

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(2) Comparison between wastes from accident and operation
Degree of countermeasure
Item of uncertainty

Waste fromoperation

Waste from
accident

Generation of waste [quantity, type, period]◎△

Handling (collecting/classifying) [difficulty]◎△

Characterization [sufficiency of information, difficulty ofsampling, representativeness of sample]○△

Technologies for processing and packaging waste○?-△

Burial and disposal methods and safety assessment△-○？

Regulations, technical standards, guidelines, siting△-○？

◎: Fully understood or good prospect, ○: Fair prospect, △: Limited?: Cannot be discussed

 Waste generated from operation has its own problem but is fairly under control.
•••

Information on basic properties of waste, including quantity at present, future change, activity and chemical substances contained in individual
waste is identified.
Both unprocessed and processed wastes are appropriately stored and managed in accordance with the current regulations.
Regulations and standards, as well as disposal method and safety assessment method, have been in place.

 Many uncertainties poses important technical problems to disposal of waste from the accident at
Fukushima Daiichi. Solving these uncertainties and bringing the waste under control are the major goalsof countermeasures and technology development.
©International Research Institute for Nuclear Decommissioning

6

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(3) Situation of contaminated water treatment at Fukushima Daiichi nuclear power
stationDecontaminationequipment:
Oil separation
Decontamination
apparatus (AREVA)
Cesium adsorption(KURION,SARRY)

Unit No. 4

Unit No. 3

Unit No. 2

Unit No. 1

Unit No. 5

Unit No. 6

Process main building
Miscellaneous solid waste
volume reduction
processing building

About 400 m3/day(quantity flowing into buildings)

About 800 m3/day
Desalination
systemAbout 400 m3/day
About 400 m3/day(to tank)

Increase in storedcontaminated water

Multi-nuclide Removal
Equipment (ALPS)

About 800 m3/day (groundwater from mountain)

Tepco HP modified©International Research Institute for Nuclear Decommissioning

7

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(4) Outline of contaminated water treatment equipment and waste generated
Cesium removal
Oilseparator

Waste zeolite
M

Cesium absorption
vessel (KURION)

Waste sludge
M

M

M

Decontamination
apparatus(AREABA)
SPT(B)
tank

Secondary cesium
absorption vessel(SARRY)

Desalination

Waste zeolite

Reactor building
Fresh
water
tank

Desalination
plant (RO)Concentrat
ed liquid
waste tank

Desalination plant(evaporator andconcentrator)

Removal of radioactive nuclides
other than tritium
Advanced Liquid
Processing
System (ALPS)

Waste slurry
Concentrat
ed salt
water tank

- Iron (III) hydroxide- Carbonate

Waste adsorbent- Activated carbon- Titanate- Ferrocyanide
compound- Activated carbon withAg loaded- Titanium oxide- Chelate resin- Resin-based
adsorbent

Treatedwater tank©International Research Institute for Nuclear Decommissioning

8

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(5) Outline of cesium adsorption apparatus (KURION)
Outline of cesium adsorption apparatus (KURION)

Appearance of cesium adsorption apparatus

Pressure vessel(adsorption vessel)

Pump

Pump
Pressure vessel(adsorption vessel)

Pump

Pressure vessel(adsorption vessel)

Outlet

Inlet

Skid for removing oil andtechnetium

Skid for removing cesium

Skid for removing iodine

(H)

(AGH)

 Dimensions: About 1.4 m (dia.) 
2.4 m (H) Weight: About 15 t
 Zeolite-filled stainless steel vessel
within carbon steel container as
shield Adsorption vessel housed in adsorption skid
 To reduce exposure of workers, vessel isreplaced when the dose equivalent rate on skid
surface reaches about 4 mSv/h.
 Processing capacity: 1,200 m3/day/4 systems= 12.5 m3/h/system Adsorbent KURION-H
KURION-AGH©International Research Institute for Nuclear Decommissioning

92. Waste resulting from accident at Fukushima Daiichi nuclear power station(6) Outline of secondary cesium adsorption apparatus (SARRY)
Outline of secondary cesium adsorption apparatus(SARRY) Started operation in August 2011
 Dimensions: 1.4 m (dia.)  3.6 m (H) Weight: About 24 t

Outlet

Inlet

Pump

FilterAdsorption vessel

FilterAppearance of secondary cesium adsorption vessel Zeolite-filled stainless steel vessel withindouble-layered carbon steel container forshielding, with lead in its annulus
 To reduce exposure of workers, vessel isreplaced when its dose equivalent rate on
the surface of the adsorption vessel
reaches about 4 mSv/h.
 Processing capacity: 20 to 25 m3/h/system 2 systems = 40 to 50 m3/h
 Adsorption apparatus makeup• Preprocessing filter• Preceding stage: Synthesized zeolite
for light decontamination• Succeeding stage: Titanium silicate* forthorough decontamination• Trailing dust filter*Name: Crystalline titanium silicate, CST, titanosilicate©International Research Institute for Nuclear Decommissioning10

2. Waste resulting from accident at Fukushima Daiichi nuclear power station（7）Outline of Multi‐nuclide Removal Equipment （ALPS※１）
 purpose
Existing water treatment facilities can mainly remove cesium. Besides, other nuclides are
removed to reduce their concentration in the contaminated water aiming at the level of
notification below ※２※１：Advanced Liquid Processing System※２： Notification for Dose Equivalent Limits on the Basis of the Ministerial Ordinance forCommercial Power Reactors

 Multi‐nuclide removal equipment structureCharacteristics and supposed generation of sludge and adsorptionPre
treatment

Adsorbent

Iron
coprecipitation
carbonate
precipitation

carbonate

Sr, Mg, Ca

Adsorbent １

activated charcoal

colloid

Adsorbent ２

Sr

Adsorbent ５
Adsorbent ６

titanate
Ferrocyanide
compoundAg impregnatedcarbontitanium oxidechelate resin

Adsorbent ７

resin adsorbent

Adsorbent ３
Adsorbent ４

Ａ，Ｂ，Ｃ line system
1 line （50% running flow）treatment amount ：250m3 per day
3 lines are in operation （750m3 per day)

ColumnMajor component Component to be removed
organic substance, αferric hydroxide Ⅲ
nuclide, Co, Mn

Cs
I
Sb
Co
Ru etc.
Negatively charged colloid©International Research Institute for Nuclear Decommissioning

11

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(8) Cesium adsorption vessel and HIC (High Integrity Container) temporary storage

KURION (cesium adsorption apparatus) adsorption
vessel storage (stored in concrete box culvert)

SARRY (secondary cesium adsorption apparatus)
adsorption vessel storage

Top coverthrough-hole
About 150 mm

Cesium adsorption vessel storage

1828.8mm

About 100 mm dia.

About 400 mm

Serial No.
Serial No.

About 200 mm
Cover closing/stopping guide

Top cover throughhole (Sunlight will notgo through.)Corridor

1524mmHIC storage

11.4mmt
Appearance of
HIC type 1 HIC

Box culvert

Temporary storage of HIC©International Research Institute for Nuclear Decommissioning

12

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(9) Features of debris and felled trees

Features of debris

Features of felled trees• Reinforced concrete scattering
around reactor building• Reinforced concrete removed form
remaining buildings• Metallic objects scattering aroundreactor building and removed fromremaining buildings• Organic (cellulose) waste• Low dose on trunk• Surface contamination and partiallycontaminated inside• Mixed with salt and scattering
inhibitor• Volume of branches and leaves canbe reduced by chipping.
• Humification has progressed,
making it difficult to be separated
from soil.

–
–
–
–
–

Contamination of low to high concentration
mainly by radioactive material in gas form
In addition to the surface, the inside is also
partially contaminated.
Salt and scattering inhibitor adhered.
SUS, carbon steel, aluminum alloy, led(shield)
PVC sheath (organic matter) such as ofcables©International Research Institute for Nuclear Decommissioning13

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(10) Temporary storage conditions of debris and felled trees (1/3)
Ｇ

Ａ・Ｂ
Ｇ
ＢＡ
Ｌ

Ｉ

Ａ

Ｃ

Ｈ

Ｄ

Ｉ

P
Ｆ

Ｌ

Ｅ

Ｃ

Ｍ

固体廃棄物貯蔵庫

固体廃棄物貯蔵庫
Ｄ

W
Ｑ
Ｒ

Ｖ
Ｏ

Ｑ
Ｓ
瓦礫保管エリア
伐採木保管エリア
瓦礫保管エリア（予定地）
伐採木保管エリア（予定地）
セシウム吸着塔保管エリア
スラッジ保管エリア
セシウム吸着塔保管エリア（運用前）
スラッジ保管エリア（運用前）

使用済セシウム吸着塔
保管施設

Ｎ
Ｊ

Ｕ
Ｔ

©International Research Institute for Nuclear Decommissioning

14

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(11) Temporary storage conditions of debris and felled trees (2/3)

 Conditions of soil-covered temporary storage facility
Soil for covering andshielding: 1 m min.

Observation hole

Water shut-off sheets

Cushioning material
Protective sheet

Debris, etc.

About
5m

GroundProtective
soil

Bentonite sheet

Structure of soil-covered temporary storage facility
Protective
soil
CushionDebrisInternal status©International Research Institute for Nuclear Decommissioning15

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(12) Temporary storage conditions of debris and felled trees (3/3)
Temporary storage conditions of felled treesFelled trees are separated into trunks, roots,
and branches and leaves, and temporarily
stored at fixed locations on Fukushima Daiichi
site.
Branches and leaves are stored in a soilcovered temporary storage tanks (whose
temperature is regularly measured), to lower
the dosage and risk of a fire.

Temporary storage of felled trees (trunks)
A-A’ crosssectional view

Water shut-off
sheetsGas release tube

Protective sheet
Isolation
distance:
2 m min.

Thermometers
Covering soil

About
1m
Felled trees
reduced in volume

About
3m

GroundTemporary storage of felledtrees (branches and leaves)

Structure of felled tree temporary storage tank©International Research Institute for Nuclear Decommissioning

16

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(13) Change in quantity of stored debris
120,000

Stored quantity (m3)100,000

80,00060,000Outdoor accumulation (less than 0.1 mSv/h)

40,000Covering with sheet (0.1 to 1 mSv/h)

20,000Soil-covered temporary storage facility, temporary storage facility, container (1 to 30 mSv/h)0
2012年02月
Feb 2012

Solid waste storage (30 mSv/h min.)2012年08月
Aug 2012

2013年02月
Feb 2013

2013年08月
Aug 2013

2014年02月
Feb 2014©International Research Institute for Nuclear Decommissioning17

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(14) Changes in quantity of felled trees stored
120,000

Stored quantity (m3)100,000

80,00060,00040,000Outdoor accumulation

20,000Felled tree temporary storage tank0
2012年02月
Feb
2012

2012年08月
Aug 2012

2013年02月
Feb 2013

2013年08月
Aug 2013

2014年02月
Feb 2014©International Research Institute for Nuclear Decommissioning18

2. Waste resulting from accident at Fukushima Daiichi nuclear power station

1,200

1,200

Processing column

1,000

1,000

Mobile processing system
800

800

HIC
600

Sludge600

Secondary cesium absorption vessel
400

400

200

Cesium absorption vessel0
2011年05月
May 2011 2011年11月
Nov 2011 2012年05月
May 2012 2012年11月
Nov 2012

Quantity of sludge stored (m3)

Quantity stored in absorption tower (main, base, tower)(15) Changes in quantity of water-treated secondary waste stored2000
2013年05月
May 2013 2013年11月
Nov 2013 2014年05月
May2014©International Research Institute for Nuclear Decommissioning

19

2. Waste resulting from accident at Fukushima Daiichi nuclear power station(16) Draft of policy for reducing volume and recycling waste
Type of waste

Temporary processing

Future plan

• Reducing volume by pulverizing• Encouraged recycling as sub-basematerials

Recycling as reproduced concrete to beconsidered

Metal debris• Reducing volume by cutting

Volume reduction by melting andcreating ingots and recycling into castingproducts to be considered

Felled trees• Reducing volume of trunks with a low
surface dose rate by burning, while
continued storage in the original form
outdoors• Reducing volume of branches andleaves with a high surface dose rate,
while continuing soil-covered storage
for the purpose of shielding

Recycling to be considered if there areneeds

Concrete debris* The level of contamination that will serve as a criterion for recycling should be decided by assessing
in detail the radiation workers and users have been exposed to, and radiation released to theenvironment.
* A large amount of metal waste will be generated in the future because it is planned to replace 377
steel cylindrical tanks (flange-joined) with welded tanks.
©International Research Institute for Nuclear Decommissioning20

３．１ Outline of research into processing and disposal of radioactive waste(1) flow chart and research items on disposal of radioactive waste
Analysis on radioactive nuclide in waste

Characterization••

Analysis on concentrations ofradioactive nuclide
Physical properties and chemicalcomposition

Hydrogen generationfrom the waste[H2] ≤1.51%(less than
explosive limit)

Long term storage•
JAEA staff collecting
rubble near 4th reactor

Assessment of safety on long term
storage- Safety of hydrogen gas- Corrosion of container- Stability of waste

Analytical work on 3H, 14C

Measurement of thermal conductivity of waste

Conditioning （processing for disposal
package）• Technical research on conditioning• Basic test for technical assessment- Vitrification- Geopolymer solidification

Calculated result onhydrogen concentration

Basic test for technicalassessment on conditioning

Disposal••Measurement of effective thermal
conductivity of zeolite layer

Investigating disposal concept
Disposal safety assessment⁃ Domestic and international research on
disposal concept and safetyassessment techniques, and theirapplicability

Vitrification

Geopolymer
solidification©International Research Institute for Nuclear Decommissioning21

3. Outline of research into processing and disposal of radioactive waste(2) Schedule
First term
第１期

Item/year
事項/年度
HP of mid- and long-term road map

中長期ロードマップのHP
1. Characterization(1) Analysis of radioactive nuclides

2011 2012 2013

Second
term
第２期

Third
term
第３期

2014 2015 2016 2017 2018 2019 2020 2021
(First)(前）(Middle)（中）(Second)10年後以内▽
Within after 10
（後）
years

▲HP SW‐1

After
2020～25
25年後
years

After 3030～40
40
years
年後

▲HP SW‐2▲HP SW‐4▲HP SW‐3

１．性状把握• Analysis of radioactive nuclides, understanding of(1)放射性核種分析
chemical composition and physical property• Accumulation of analysis data, improvement of
・放射性核種分析、化学組成・物理性状の把握
accuracy in assessing characteristics
・分析データ蓄積、性状の評価精度向上(2) Development of nuclide analysis technologies(2)核種分析技術開発• Development of technologies for analyzing nuclides
difficult to measure
・難測定核種等分析技術開発(3) Inventory evaluation(3)インベトリ評価• Development of assessment techniques
・評価手法の開発• Improvement of accuracy in estimating inventory
・インベントリ推定確度の向上
2. Investigation on long-term storage
２．長期保管方策の検討(1) Study of long-term storage method(1)長期保管方策検討
3.
Investigation concerning processing of waste
３．廃棄物の処理に関する検討(1)
Technical research(1)技術調査• Research into waste package technology
・廃棄体化技術の調査• Basic test for assessing applicability
・適用性評価のための基礎試験• Study of applicability(2)・適用性の検討
Selection of candidates for processing technology• Narrowing down on technology based on basic test(2)処理技術候補の選定
results
・基礎試験結果を踏まえた技術絞り込み
4.
Investigation
concerning disposal of waste
４．廃棄物の処分に関する検討(1) Evaluation of applicability of existing processing(1)既存の処分概念・安全評価手法の適用性評価
concept and safety assessment technique• Research into existing disposal concept and safety
・既存処分概念・安全評価手法の調査
assessment technique• Stud of applicability
・適用性検討• Study of disposal concept with promising candidate
・有望候補による処分概念検討• Safety assessment with candidate technique(2)・候補手法による安全性評価
Establishment of disposal concept and safetyassessment techniques• Rationalization and advancement of disposal(2)処分概念・安全評価手法の確立
concept and safety assessment techniques
・処分概念・安全性評価手法の合理化・高度化• Improvement of reliability of disposal concept andsafety assessment techniques
・処分概念・安全性評価手法の信頼性向上
5.
Databasedevelopment
５．データベース開発•
Trial
production
・試作・試運用 and trial operation• Operation and incremental advancement
・運用・段階的高度化• Operation and improvement of database・データベースの運用・改良©International Research Institute for Nuclear Decommissioning

22

３．Outline of research into processing disposal of radioactive waste(3) R&Ds on waste management strategy
Objective of HP SW‐1
Lining up the candidate operations from generation, storage and to disposal(waste management stream) with their argument
R&D Items
1) Making plans to investigate waste management stream
Drawing up plans to create and improve the waste management stream
including subsequent processing and disposal, considering its origin and its
pathway concerning contamination with respect to each waste2) Listing relevant items in investigation on characterization, long term
storage, processing and disposal
3) Investigating candidates of waste management streams that can beapplicable for processing and disposal for each waste
 Investigating a proper candidate among waste management streams
that can be applied to processing and disposal by studying andunderstanding the effect of relevant items©International Research Institute for Nuclear Decommissioning23

３． Outline of research into processing disposal of radioactive waste(4) R&Ds on characterization
Objective of HP SW‐1
Presenting collected information on characteristics of major waste
R&D Items
1) Analysis of waste sample
 Collection and analysis of rubbles including sample inside a reactor building, soils, secondary
waste after water treatment, aiming to 50 samples per year .
2) Inventory evaluation
 Amount and inventory evaluation based on information of analytical value and origin of waste Establishment of analytical inventory estimation method through creation of nuclide transfer
model based on nuclide discharge behavior from a reactor, dissolution behavior to contaminatedwater, nuclide sorption behavior into construction material, and adsorbing materials.
 Making inventory data set of each separate waste based on evaluation with analytical results andmodeling
3) Data collection on secondary waste Data collection for investigating long term storage methods and processing technologies
 Data collection on physical and chemical characteristics of secondary waste generated fromsecondary cesium adsorption equipment or multi‐nuclear removal equipment
4) Investigation of nuclide analysis which has difficulty in measurement
 Establishing chemical procedure for unanalyzed nuclides
 Analytical method of high dose waste©International Research Institute for Nuclear Decommissioning24

３．Outline of research about processing and disposal ofradioactive waste(5) R&Ds on long term storage
Objective of HP SW‐1
Presenting the way of long term storage for secondary waste from water
treatment for optimization and rationalization of processing and disposal.
R&D Items
1) Stabilization of multi‐nuclide removal equipment slurry As for slurry from pretreatment of multi‐nuclide removal equipment, presenting theway to remove water and to stabilize.
 Other countermeasures are taken according to the situation on the site.
2) Evaluation of vessel from Cs adsorption
 As for the safety of storage for Cs adsorption vessel (KURION), presenting an outlook
for hydrogen generation and corrosion of materials. For the corrosion,countermeasures are to be investigated.
3) Evaluation of vessel from second Cs adsorption
 As for the safety of storage for Cs adsorption vessel (SARRY), presenting an outlook
for hydrogen generation and corrosion of materials.
©International Research Institute for Nuclear Decommissioning25

３．Outline of research about processing and disposal ofradioactive waste(6) R&Ds on processing
Objective of HP SW‐1
Assembling information on treatment and conditioning technologies for existing waste, presenting candidate
conditioning technique which could be applicable for wastes (secondary waste from water treatment, rubbles, trees orother) which characteristics are understood to some extent, and conducting initial selection.

R&D Items
1) Basic test of treatment and conditioning To get data needed for evaluation of treatment and conditioning technique for each waste basedon the waste research result, characteristics of waste and the study for disposal method.
2) Research for technique of treatment and conditioning After the research of existing technique of treatment and conditioning, making a technique
catalog including their outline, status of technical development, past employment, throughput,
performance of waste package, etc.
 Evaluating techniques with consideration on waste properties and latest result on waste disposal
method, and extracting data required for future selection.
3) Initial selection of candidate technique Selecting technique of treatment and conditioning for R&Ds aiming at HP SW‐2 considering theresult of 1) and 2) and the R&D result from characterization and disposal method.
©International Research Institute for Nuclear Decommissioning26

３．Outline of research about processing and disposal ofradioactive waste(7) R&Ds on waste disposal
Objective of HP SW‐1
Presenting candidate of disposal concept which considered to be applicable for eachaccident waste and their assessment method.
R&D Items
1) Research and sorting out of properties for disposal concept2) Extracting important isotopes (investigating disposal concept suited for accident waste
based on existing disposal concept)
 To understand disposal classification of waste occurred by the accident
 To understand influencing characteristic of waste occurred by the accident
 To understand the response characteristic of disposal system
 Selection of candidate for proper disposal concept and safety assessment method
３）Investigation on new disposal concept and other As for dismantling wastes which are to be produced in bulk and practically difficult
to apply existing disposal concept as it is, new disposal concept and its applicability
are to be examined.
©International Research Institute for Nuclear Decommissioning27

3.1 Characterization(1) Result of estimating distribution of Cs-137
Released into atmosphere: 1.0E + 16 Bq (1.4%)

Released into ocean: 3.8E + 15 Bq (0.5%)

Adsorbed by water treatment facility:
2.5E + 17 Bq (35.4%)

Initial inventory(1F1-3) 7.0E +17 Bq

Remains in reactor (estimated):
4.3E+17 Bq (61.6%)

Contaminated water:
7.2E + 15 Bq (1.0%)©International Research Institute for Nuclear Decommissioning28

3.1 Characterization(2) Model assessment of mass transfer of major nuclides to contaminated water
Changes in concentration of
137Cs in contaminated water

Changes in concentration of

90Sr in contaminated waterCalculated value

Calculated value
90Cs concentration (Bq/mL)

137Cs concentration (Bq/mL)

Centralized RW
HTICentralized RW
HTIYear/month

Year/month
137Cs

134Cs

90Sr

125Sb

106Ru

60Co

54Mn

Initial reactor core inventory (Bq)

7.0E+17

7.2E+17

5.2E+17

4.2E+16

2.2E+18

9.4E+122.8E+14

Initial concentration (Bq/ml)

2.4E+06

2.2E+06

1.6E+05

1.7E+02

2.3E+01

1.4E+02

5.2E+02

Continuous transfer rate (Bq/d)

2.5E+13

2.6E+13

3.6E+13

1.3E+108.8E+09

1.4E+09

1.3E+0930.04y2.065y28.74y2.758Y

373.6Y

5.271y312.1yHalf-life••••

More than 99% of Cs-137 and Cs-134, which moved to contaminated water in the initial stage has been decontaminated.
Estimated that the main fraction at present are those continuously transferred from fuel and others (equivalent to about 2.1% of those remaining
in the reactor/year).
Other than Cs, more than 99% of FPs which transferred to contaminated water in the initial stage, has already been diluted and transported. At
present, the components that are continuously transferring from the fuel and others are estimated to be the main fraction.
The analysis value of Co-60 varies much more than that of other nuclides and thus Co-60 may have flown from more than one sources (wastethat had been stored before the accident). (At present, the main fraction is estimated to be those which are continuously transferring.)©International Research Institute for Nuclear Decommissioning29

3.1 Characterization(3) Method of assessing radioactive concentration
Assessment of radioactive concentration
on debris and felled treesConcrete and debris from areas around units 1, 3, and 4,
where debris was scattering were sampled and theirradioactivity was analyzed.
Felled trees in two storage areas and by collecting pinetree leaves around unit 3 were sampled and theirradioactivity was analyzed.
Debris storage area
Felled tree storage area
Debris storage area (planned)
Felled tree storage area (planned)

Assessment of radioactive concentration of
secondary waste for contaminated water
treatment

Dose of waste zeolite and sludge stemming from contaminatedwater treatment is high and direct analysis of radioactivity isdifficult.
 Indirect assessment using result of analyzing radioactivity ofcontaminated water and treated water is under way.



Basic concept of inventory assessment

Felled tree
Nuclideanalysis

Felled tree
Contaminatedwater

Live tree

Nuclideanalysis

Nuclideanalysis

Nuclideelimination
system 1

Sampling
treatedwater

Nuclideelimination
system 2

Sampling
treatedwater

Nuclides to be analyzed

Solid waste storage
Important antiseismic building

Sampling
treatedwater

Inventory
assessment

Inventory
assessment

Debris (concrete, metal)
Debris (concrete, metal, pebbles)

Front gate
First debris gathering
Second debris
gathering
Third debris gathering Referring to nuclides for assessment of existing
disposal systems, the following nuclides have been
selected to be analyzed:
γ-ray nuclides: 60Co, 94Nb, 137Cs, 152Eu, 154Eu
β-ray nuclides: 3H, 14C, 36Cl, 41Ca, 59Ni, 63Ni,
79Se, 90Sr, 99Tc, 129I, 241Pu
α-ray nuclides: 233, 234, 235, 236, 238U, 237Np,
238,239,240,242Pu, 241,242m,243Am,
244,245,246Cm©International Research Institute for Nuclear Decommissioning30

3.1 Characterization(4) Result of nuclide analysis
Result of analyzing nuclides on debris and felled trees

Result of estimating inventory ofcesium adsorption vessel(debris and felled trees: 57 samples)Radioactive concentration (Bq/g)
NuclideDebris & felled trees

γnuclideβnuclideαnuclideKURION

SARRY

Total adsorbed (Bq)

1.7E+178.0E+16

Average adsorbed(Bq/unit)

4.0E+141.0E+15

Recommended upper-limit
value of concentration*1
TrenchdisposalPit disposal

60Co

ND（<7E-02） - 5.6E+00

1E+04

1E+09

137Cs

2.0E+00 - 1.9E+05

1E+02

1E+08

14C

ND（<5E-02） - 2.7E+00

-

1E+05

63Ni

ND（<5E-02）

-

1E+07

79Se

ND（<5E-02） - 2.1E-01

-

-

90Sr

ND（<5E-02） - 1.0E+02

1E+01

1E+07

99Tc

ND（<5E-02） - 8.9E-02

-

1E+03

238Pu

ND（<5E-03）-

239Pu＋240Pu

ND（<5E-03）

-

1E+04

241Am

ND（<5E-03）-@Am-241

244Cm

ND（<5E-03）-

The concentration exceeds the upper-limit value
for pit disposal and it should be taken into
consideration when processing and disposal
methods are studied.

Result of analyzing contaminated water Contaminated water analyzed: 25 samples
 2.4E-03 Bq/g of 238Pu and 8.3E-04 Bq/g of
239Pu+240Pu were detected from thecontaminated water (centralized RW・HTI/B).
 From the isotopic composition, Pu is
considered to stem from the accident at
Fukushima Daiichi nuclear power station andequivalent to that found in the environment
which is attributed to fallout.

*1: “About upper-limit value of radioactivity concentration for burial disposal of low-level
radioactive solid waste”(Nuclear Safety Commission: May 21, 2007)©International Research Institute for Nuclear Decommissioning31

3.1 Characterization(5) Distribution of radioactivity on standing trees in 1F site (1) Areas from which standing trees were
sampled were decided based on the
distribution of the air dose rate in the site of
the power station. Areas around the reactor
buildings where the air dose rate was highwere divided into small segments (see thefigure on the right).
 Pine trees, representative trees in the site,were sampled. Branches and leaves at aheight of 4 m above ground were collectedfrom three trees in each area.
(Note that the type of the tree and height werechanged depending on the situation of thelocation.) The surface dose rate of the samples wasmeasured and branches and leaves (or grass)with a high dose rate (1 to 3 samples/area) in
all the areas were selected as samples foranalysis.

©International Research Institute for Nuclear Decommissioning3.1 Characterization(5) Distribution of radioactivity on standing trees in 1F site (2)
S

T

Q

R
P

N

L
M
K

Air
dose

D

G

C

F

B

E

A

Q

R

L
M
K

I

: >200 μSv/h: 100‐200 μSv/h: 50‐100 μSv/h: 20‐50 μSv/h: 10‐20 μSv/h: <10 μSv/h

T

H

D

G

C

F

B

E

A

S

T

Q

R

Q

R

P

L
M
K
:
>1 Bq/g: 0.5‐1 Bq/g
: 0.2‐0.5 Bq/g
: 0.1‐0.2 Bq/g
: 0.05‐0.1 Bq/g
: ND (<0.05Bq/g)

G

C

F

B

E

A
I
J

L
M
K
:
>1 Bq/g: 0.5‐1 Bq/g
: 0.2‐0.5 Bq/g
: 0.1‐0.2 Bq/g
: 0.05‐0.1 Bq/g
: ND (<0.05Bq/g)

79 Se

P

N
D

R

K
:
>1 Bq/g: 0.5‐1 Bq/g
: 0.2‐0.5 Bq/g
: 0.1‐0.2 Bq/g
: 0.05‐0.1 Bq/g
: ND (<0.05Bq/g)

14 C

O

Q

M

I

T

H

T

L

J

S

N

S

O
H

D

G

C

F

B

E

90Sr

P

N

O

: >200 Bq/g
μSv/h
: 100‐200 Bq/g
μSv/h
: 50‐100 Bq/g
μSv/h
: 20‐50 Bq/g
μSv/h
: 10‐20 Bq/g
μSv/h
: <10 Bq/g
μSv/h

J

137Cs

P

N

O
H

S

H

D

G

C

F

B

E

A

J

T

Q

R

L
M
K
:
>1 Bq/g: 0.5‐1 Bq/g
: 0.2‐0.5 Bq/g
: 0.1‐0.2 Bq/g
: 0.05‐0.1 Bq/g
: ND (<0.05Bq/g)

3H

P

N

O

I

S

O
H

D

G

C

F

B

E

A
I
J

 Air dose rate, 137Cs, 90Sr, and 3H are similarly
distributed (higher at around reactor building)and the main source is estimated to be theradioactive nuclides that spread at the timeof the accident.
 90Sr concentration is high in area K.

AIJ Distribution of 14C and 79Se will be assessed
based on analysis data to be gathered in the
future and progress of accident analysis.
©International Research Institute for Nuclear Decommissioning3.1 Characterization(6) Relation of radioactivity concentration of 137Cs and 90Sr on debris, etc.

Around Unit 1
Around unit 3

Radioactivity concentration of 90Sr (Bq/g)

Around unit 4

90Sr/137Cs ratio
was about 23% in
the case of theaccident in
Chernobyl.

difference between debris and felledtrees and was in the range of 0.002 to
0.62%.

Felled tree
Standing tree aroundunit 3

Trend of proportional relation can beseen between 137Cs and 90Sr
concentration on debris and felledtrees.

The trend of debris differs dependingon the location where the sample wascollected and the sample. Although at
present the number of data is limited,collecting data will be continued to
improve the precision of the correlation
between the two.



90Sr/137Cs ratio (about 23%) for waste

Lower-limit for detection

Radioactivity concentration of 137Cs (Bq/g)

90Sr/137Cs ratio showed rather small

generated from the Chernobyl accident
is close to the composition in fuel and it
is considered that the difference indevelopment between the two accidents
reflects on the ratio of 137Cs and 90Sr in
waste.
©International Research Institute for Nuclear Decommissioning34

3.2 Long-term storage(1) Assessment of generation of hydrogen and hydrogen cyanide, and analysis oftemperature in storage tank during temporary storage of waste sludge
Assessment of generation of hydrogenand hydrogen cyanide Assessed by conducting irradiation test under condition
where seawater components and sludge coexist(1) Radiation chemical yield of hydrogen (G value)──

The G value, to which ferrocyanides and seawater contribute,was within 2 times that of pure water at maximum.
Hydrogen concentration in the storage tank is not considered toreach the explosive limit (4%) because the tank is ventilated.

(2) Generation of hydrogen cyanide (HCN)─

Waste sludge temporary storage tank

Hydrogen cyanide in gas phase was less than detection
limit (less than 5 ppm) at irradiation equivalent to storage for10 years (6MGy).

Mitsubishi Heavy Industry News, No. 5131, November 14, 2011

Analysis of temperature in storage tank

Result of calculating temperature
distribution in tank

 Temperature initially rises at a rate of about 0.03C/h and thengradually reaches steady state. It is assessed that it levels out
at outdoor temperature +20C after about 50 days.
 When the outdoor temperature is 40C, the center temperature
is about 60C, which is sufficiently lower than thedecomposition temperature of ferrocyanide (250C to
280C). It is thus considered that hydrogen cyanide is not
generated by thermal decomposition.
©International Research Institute for Nuclear Decommissioning35

3.2 Long-term storage(2) Analysis of temperature and hydrogen distributions in zeolite adsorption vessel Distribution of temperature and hydrogen in
the waste zeolite adsorption vessel wasassessed, assuming that the lower part of
the water outlet tube is occluded byremaining water, cutting off inflow of air,in the initial stage of storage.
 It was considered that cesium
concentration was not uniform but
distributed in the vertical direction in theadsorption vessel.

Analysis result (example)

Vent tube

Adsorption
vesselWater outlet tube

ShieldRemaining water(with the lower part of thewater outlet tube
submerged)

Schematic diagram of adsorption vessel(when water remains)Temperature
distribution

Hydrogen concentration
distribution

Hydrogen (H2) concentration calculated through analysis
Water inlet tube



1.83% max.

 Decay heat: 618 W
 Hydrogen generation (25C, 1 atm)
Zeolite layer: 24.3L/d
Submerged layer surface:
0.76L/d Thermal conductivity of submergedlayer is 1.1 times that of water

Air (discharge of hydrogen)

Plug (opened)

261C max.

Cs adsorption

Water level

Maximum
temperature

H2 concentration

618W
unequal adsorption

24 cm(lower part of water outlet
submerged)

261C

≦1.8%

504W
uniform adsorption

Ditto

211C

≦1.6%

Ditto

0cm

210C

≦1.1%

• Cs amount adsorbed and distribution in vessel (axial direction) were calculated from operation data andadsorption analysis code (ZAC) in the case of 618 W.
• In the 504 W case , Cs adsorption was calculated from analysis data of contaminated water and treated waterand adsorption equilibrium, on the assumption that Cs was uniformly adsorbed.

If water remains in the adsorption vessel, air is discharged through the upper plug. Hydrogen (H2) concentration is kept below
the explosive limit (4%) even when Cs amount adsorbed is maximum. Zeolite layer temperature is assessed to be lower than
the self-ignition temperature of hydrogen (about 560C).
Without water in the adsorption vessel, air flows in from the water outlet tube and is discharged from the upper vent plug.
Consequently, the hydrogen concentration is relatively low.
©International Research Institute for Nuclear Decommissioning36

4. Issues and measures
Possible measures

Issue

Characterization

Processing and conditioning waste

Disposal

Large quantity of waste
generated Selection of option of
decommissioning scenario

 Volume reduction (such as
incinerating organic materials andhigh-temperature melting)technologies

Insufficient inventory
assessment

 Increase of actually measured
data (existing technique) Development of innovative
nuclide analysis technique Simplifying and automating
existing nuclide analysis
technologies

 Homogenization (including blending) technology
 High-temperature melting
technology

Possibility that types andnuclide composition of raw  Ditto
waste differ from those of
conventional waste

Existence of substancesthat have not considered
for conventional waste,such as impurities andmixtureNo regulations forclassifying waste anddisposal

 Development of chemicalcomponent analysis technique－ Securing volume at disposal site Improvement of installation density by
optimizing waste package installation
method, layout, etc.

Confirmation of safety throughassessment conservatively taking
dispersion of inventory for safetyassessment into consideration

 Development of highperformance solidifier Establishment of method andtechnology for processing
specific substances (such as
ferrocyanides)

 Ditto Decontamination technology
 Segregation technology

 Confirmation of safety based on study of
scenario, model, and impact to parameters(including changes) for safety assessmentand on fluctuation of assessment result Development of disposal system thatsuppresses influences of impurities andmixtures Improvement and development of
scenarios, models, and parameters
through phenomenological understandingof influencing processes, etc.

－

 Applicability assessment of disposal
concept for each division of waste andwaste packages targeted (disposal site
feasibility and safety)©International Research Institute for Nuclear Decommissioning

37

Thank you for your attention.

©International Research Institute for Nuclear Decommissioning

38