Post-Accident Management: Lessons Learned and … Management: Lessons Learned and Preparedness...

International Conference on the Safety of Radioactive waste Management

SESSION 4

Post-Accident Management:

Lessons Learned and

Preparedness

Session 4 – PAWM IAEA-CN-242

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ORAL PRESENTATIONS

No. ID Presenter Title of Paper Page

04 – 00 INV 04 S. Nomura

Japan

Five Years Progress on Waste Management of

Fukushima-Daiichi Nuclear Accident

4

04 – 01 136 T. Kilochytska

Ukraine

Post-Accident Waste Management in Ukraine :

Challenges and Steps Needed to Resolve the

Accident Waste Problem

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04 – 02 201 M. Tichauer

France

Path Forward to the Revision of the French

Doctrine of Waste Management in Post-

Accident Situation : New Challenges in the

Light of Experience Feedback and Unveiled

areas of Development

11

04 – 03 153 Y. Koma

Japan

Inventory Estimation for Accident Waste

Generated at the Fukushima Daiichi NPS

15

04 – 04 142 N. Rybalka

Ukraine

Safety Assessment Approach for Decision

Making Related to remedial Measures and

Radioactive Waste Management

19

Session 4 -PAWM IAEA-CN-242

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POSTER PRESENTATIONS

No. ID Presenter Title of Paper Page

04 – 05 46 S.R. Mallampati

Korea, Republic of

Potential Nano-Fe/Ca/CaO Composite Enabled

Environmental Remediation Technologies for

Radioactive Waste

24

04 – 06 112 V. Havlová

Czech republic

The Treatment of Highly Radioactive Waste

Originating from a Severe Accident at a VVER

440 NPP

29

04 – 07 139 Y. Meguro

Japan

Approaches of Selection of Adequate

Conditioning Methods for Various Radioactive

Waste in Fukushima Daiichi NPS

33


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04 – 00 / INV 04. Post-Accident Waste Management – Lessons Learned and Preparedness

FIVE YEARS PROGRESS ON WASTE MANAGEMENT OF

FUKUSHIMA-DAIICHI NUCLEAR ACCIDENT S. Nomura, K. Katoh, K. Okano

Nuclear Damage Compensation and Decommissioning Facilitation Corporation (NDF), Japan

E-mail contact of main author: [email protected]

1. Introduction

More than five years have passed since catastrophic Fukushima-Daiichi (1F) nuclear accident

occurred. Much effort has been made to implement both off-site remediation and on-site 1F

decommissioning. Five years progress is overviewed in terms of management of radioactive

waste generated by the accident.

2. Off-Site Activities of Decontamination and Storage of Specified Waste

Severe accident of nuclear power plants at Tokyo Electric Power Company Holdings, Inc.

(TEPCO)’s 1F site was triggered by a huge tsunami of about 15m high. Due to the station

blackout, a core meltdown, leakage of gaseous fission products (FPs) and hydrogen explosion

occurred sequentially. Radioactive plume was released from the site and formed high dose rate

(around 200 mSv/y) area in the north-west direction within 30 km from 1F site.

The decontamination work of the land except for the difficult-to-return zone is nearly completed,

and the related evacuation orders are planned to be lifted in March 2017. For the difficult-to-

return zone, reconstruction hubs will be set to promote necessary activities and the evacuation

order may be lifted partially in FY2021 with activating decontamination work.

A huge amount of off-site waste has been generated by decontamination activities. The wastes

with radioactivity above 8,000 Bq/kg are defined as “specified waste”. The Ministry of the

Environment (MoE) evaluated the radiological decay effect of the wastes, and the amount of the

waste would be reduced to 60% of its original volume after 5 years and to 23% after 15 years.

The interim storage facilities are planned to be constructed and operated to accommodate a huge

amount (up to 22 Mm3) of soil and waste to be generated in Fukushima. The construction is

scheduled to start in November 2016.

3. On-Site 1F Decommissioning Project

Significant progress has been made in the on-site decommissioning work by TEPCO. Each

damaged core is under a stable condition and removal of Spent Fuels (SFs) was already

completed for Unit-4 in December 2014. For Unit-1 to 3, preparations of SF retrieval are in

progress. A lot of rubble went into SF Storage pools (SFPs) by the hydrogen explosions. Due to

high dose level, full remote handling is needed to remove rubble before SF retrieval. 2

TEPCO and KAJIMA Corporation have successfully developed and operated the rubble handling

and transfer system with the complete remote operation using GPS.

mailto:[email protected]


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Contaminated water management is the most important issue from the early stage of

decommissioning for damaged LWR, such as the cases of Three Mile Island Unit 2 (TMI-2) and

1F Unit 1-3. Nuclides from contaminated water remaining in the reactor and turbine buildings

have been removed by introducing facilities such as multi-nuclide removal equipment (ALPS).

Through such operation, secondary wastes (absorbent, column, sludge etc.) which are needed to

be safely stored have been generated. Release of treated water containing tritium is a key issue for

waste management along with concerns from sociological aspect.

Dose rate on the ground except for the area surrounding the damaged reactor buildings is

successfully reduced to less than 5μSv/h by shielding of facilities, removal of scattered rubble

and facing activities. Effective dose rate at the site boundary achieved a target of 1 mSv/y.

Reduction of waste volume and safe storage is a priority action for management of waste

generated by the accident. TEPCO projected the volume of solid waste generated and stored by

FY2027. With incineration and volume reduction, the volume stored in outdoor temporary

storage will be reduced from 750,000 m3 to 200,000 m3 by FY2027. For safe storage, incinerator

ash and processed rubble will be stored in the storage facilities to reduce risks.

The risk reduction strategy for a mid- and long-term on-site decommissioning is described in the

“Technical Strategic Plan 2016” for decommissioning of 1F Units developed by NDF. Risk

assessment has enabled to identify, quantify and prioritize all the major risk sources. Among

them, SFs in SFPs of damaged reactor buildings, and contaminated water in the buildings should

have the highest risks and risk reduction should be conducted as soon as practical. However, the

risk level of stable storage of solid wastes is sufficiently low, and it can be maintained by

ensuring continuous management in the future. For strategy on 1F waste management,

minimization of waste generation and safe storage should be ensured with enhancing

characterization of solid waste and developing waste stream to carry out the study on processing

and disposal.

One of the most challenging issues on the waste management is related to fuel debris retrieval.

Once the operation has started, a large quantity of high-dose radioactive solid waste may be

generated. To retrieve fuel debris safely, smoothly and efficiently, it is important to know the

locations and characteristics of fuel debris, and research has been conducted based on the lessons

learned from the cases such as TMI-2, Chernobyl and Windscale.

The information on the characteristics of solid waste has been accumulated by sampling

specimens and conducting radiological analysis, but analytical capability still needs to be

improved. JAEA is planning to build Radioactive-Material Analysis and Research Facility within

1F site in this regard.

4. Conclusion

A huge amount of off-site specific waste is planned to be managed by constructing and operating

interim storage facilities. On-site management of solid waste generated by the accident should be

sustained as long-term key activities, such as safe storage, characterization, processing and

disposal of various wastes. Effective collaborations among JAEA, TEPCO, NDF, International

Research Institute for Nuclear Decommissioning (IRID), and other domestic and international


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organizations and companies are strongly required to tackle challenging projects on 1F

decommissioning.


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04 – 01 / ID 136. Post-Accident Waste Management – Lessons Learned and Preparedness

POST-ACCIDENT WASTE MANAGEMENT IN UKRAINE: CHALLENGES AND

STEPS NEEDED TO RESOLVE THE ACCIDENT WASTE PROBLEM

T. Kilochytska1, L. Zinkevich

2, I. Shybetskyi

3, J. Krone

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1State Nuclear Regulatory Inspectorate of Ukraine, Kyiv, Ukraine

2State Agency of Ukraine on Exclusion Zone Management, Kyiv, Ukraine

3Radioenvironmental Centre of National Academy of Sciences of Ukraine (REC), Kyiv,

Ukraine 4DBE TEHNOLOGY GmbH, Peine, Germany


Abstract. The paper deals with the radioactive waste that arose as a result of the Chernobyl accident (i.e.,

accident waste’), the primary sites where the waste has been located within the Chernobyl Exclusion Zone, the

challenges related to the further management of the waste, and the corresponding need for improving the

radioactive waste (RW) classification in Ukraine. Finally alternative approaches for solving the problems

associated with accident waste disposal and the related economic obstacles are discussed.

Key Words: Chernobyl NPP, accident waste, waste management.

1. Accident waste: inventory and status

A huge amount of radioactively contaminated materials and radioactive waste (RW) resulted

from the Chernobyl accident. These materials and waste are concentrated within the 30 km

Chernobyl Exclusion Zone (ChEZ). The main sites of RW accumulations within the ChEZ

are the following:

Shelter Object (SO) covering the destroyed 4th

unit of the Chornobyl NPP (ChNPP),

Radioactive waste disposal sites (RWDS): “Buriakivka”, “Pidlisny”, “3rd

Line of

ChNPP”

Numerous temporary localization sites of radioactive waste (TLSRW),

Storage and disposal facilities of the “Vector Complex”.

Between 400,000 and 1,740,000 m3 of RW are located inside the SO or at associated onsite

storages with a total activity of about 4.1E+17 Bq (as of 2005). Of this over 10% of the RW

is classified as high-level waste (HLW), a large amount of which consists of concrete, metal

structures and equipment, and material used for backfilling the damaged unit 4. Over 2,800 t

of HLW are classified as fuel-containing materials (FCM), including lava-like fuel-

containing material, fragments of the reactor core's graphite cladding and fuel dust [1].

The management strategy for the SO was adopted in 2001 [2] and includes three stages:

Stabilization of unstable structures and components of the SO

Construction and commissioning of a new safe confinement (NSC)

Extraction of RW and fuel containing materials and their safe disposal.

Implementation of the third stage is the most difficult challenge for Ukraine due to:

The accelerated degradation of FCM and intense dust generation

The need to construct facilities for the treatment of SO liquid waste that contains

transuranic elements and organic compounds



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The need to develop unique technologies for dismantling unstable structures and for the

subsequent retrieval of RW and FCM.

Currently, the construction of the NSC is ongoing; the designs for dismantling SO's unstable

structures and for retrieving RW and FCM are being developed at a conceptual level. The

above mentioned management activities should be completed before the end of the NSC's

designed lifetime of 100 years.

The total volume of RW inside the ChEZ (excluding the SO) amounts to about 2.8 million

m3. More than 2.0 million m

3 of this waste with a total activity of about 7.25E+15 Bq is

located at the RWDSs and TLSRWs. The waste consists mainly of intermediate level waste

(ILW) and short-lived low level waste (LLW) [1]. Accident waste varies widely with respect

to specific activity and radionuclide and material composition. In contrast to other types of

RW (e.g., operational RW), accident waste is characterized by a wide spectrum of

radionuclides, including those with long half-lives.

To date, only the operation of the RWDS "Buriakivka" (near-surface trench-type repository)

has been fully licensed. Other RWDS's and TLSRW's established shortly after the accident

for the disposition and management of accident waste are being monitored in order to ensure

an acceptable level of safety. Most of the accident waste at TLRSWs is stored under

conditions that do not meet the requirements of Ukrainian radiation safety norms.

The Vector Complex has been constructed inside the ChEZ specifically to manage the huge

quantities of accident waste. The complex includes the following near surface disposal

facilities for short lived L/ILW:

Lot-3 for ChNPP operational waste

SRW-1 and SRW-2 in particular for accident waste

Long-term safety of these facilities is provided by a system of engineered barriers designed

and constructed in a manner similar to those implemented at the RW disposal facilities

located at Centre de l'Aube (France) and El Cabril (Spain).

2. Challenges of accident waste management

2.1. Drawbacks of the current system of radioactive waste classification in Ukraine

The current Ukrainian waste classification system [3] focuses mainly on ensuring radiation

safety of the population and operational personal during pre-disposal waste management

activities (collection, sorting, transportation and storage) at nuclear facilities. It does not

support a cost-effective disposal for all types of existing waste. The existing system only

considers two repository disposal options: near-surface repositories for short-lived waste and

deep geological repositories (DGR) for long-lived waste.

2.2. Ukrainian safety requirements for near-surface repositories

A large portion of the activity contribution for most of the accident waste results from the

large quantities of long-lived radionuclides generated as a direct result of the reactor core

accident. The contribution from long-lived radionuclides constitutes the main problem

associated with near-surface disposal. Additionally, it should be noted that the Ukrainian

standards are overly-conservative when compared to IAEA recommendations:

According to [4], waste with a specific activity of long-lived alpha emitting nuclides of

up to 4,000 Bq/g can be disposed in near-surface facilities; whereas in Ukraine a limit

of 0.1 Bq/g is often applied.

According to [5], the future exposure of a representative of the critical group by a

disposal facility should not exceed a dose constraint of 0.3 mSv a year. In Ukraine this

limit is 0.01 mSv a year [3].


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The Ukrainian normative documents require the application of very "strict" intrusion

scenarios for evaluating the safety of the disposal facility and the suitability of the

waste for near-surface disposal (for example, the direct consumption of water from a

well that has been drilled into the waste pile at the end of the institutional control

period).

Taking into account the current strict national safety requirements for near-surface disposal as

applied in Ukraine, the majority of accident waste would require disposal in a geological

repository. Following this approach it would be practically impossible to dispose this waste in

such a way due to economic considerations.

3. Steps needed to resolve the management problem of accident waste in Ukraine

3.1. Implementation of an advanced waste classification scheme

A new classification system focused on RW disposal was developed in 2011-2012 within the

framework of EC Project U.04.01/08-C “Improvement of the Radwaste Classification System

in Ukraine”. The project was carried out by a consortium of DBE TECHNOLOGY GmbH

(Germany, consortium leader), SKB International AB (Sweden), ANDRA (France), COVRA

(Netherlands), and ENRESA (Spain) in cooperation with REC (Ukraine).

The new classification system (Table I) is based on four final disposal options (i.e., repository

types) each corresponding to a waste class and consistent with IAEA recommendations [4].

TABLE I. DISPOSAL OPTIONS PROPOSED FOR UKRAINE ACCORDING OF NEW

CLASSIFICATION SCHEME

Waste Class Description Disposal option

Very Low Level Waste

(VLLW)

Not determined in Ukraine yet, but it is

expected that large waste volumes can

be classified as VLLW

Landfill repository (type I)

Low Level Waste

(LLW)

Corresponds to the existing short-lived

waste category - except VLLW Near-surface repository (type II)

Intermediate Level Waste

(ILW)

This corresponds to existing long-lived

waste category - except HLW

Repository at intermediate depth (type

III), or deep geological repository (type

IV)

High Level Waste

(HLW)

This corresponds to existing in

Ukraine heat-generating/HLW

category

Deep geological repository (type IV)

The new classification system in Ukraine will provide significant improvements in the

economics of waste management [6] by cost effectively allocating all existing waste,

including both operational and accident waste, as well as future waste arisings to the most

optimal repository solution for each waste class.

3.2. Updating National Safety Standards

It is necessary to eliminate, as appropriate, excessive conservatism in some of the

requirements for RW disposal in Ukraine and to bring the national norms and standards into

line with with international requirements and best practices.

3.3. Specific Waste Acceptance Criteria for Accident Waste

Special, less restrictive Waste Acceptance Criteria (WAC) for VLLW and LLW accident

waste disposal facilities within the ChEZ, could be established based on safety assessments

that take credit for the special restrictions applied to the ChEZ. In particular these restrictions


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include the requirement to limit public access to the entire area encompassed by the ChEZ

and to restricted land use over long timeframes because of the overall level of contamination

within the zone. These WAC can be derived from estimates of potential radiation exposure to

critical groups living completely outside of the ChEZ and would be applicable only to

accident waste. However, after the end of an assumed institutional period that can be assured

for only a few hundred years, the potential exposure to any individual even inside the ChEZ

should not exceed under any circumstances the intervention limit of 1 mSv per year.

3.4. Improvement of Waste Characterization

To best benefit from the new classification system improvements in sorting and

characterization processes are needed. Improved characterization and sorting of RW will

require appropriate planning; taking into account the specific characteristics of the accident

waste and the development of appropriate WAC for disposal.

4. Conclusions

To date the same approaches have been used for the management of normal operational waste

and Chernobyl accident waste in Ukraine. An assessment of the disposability of accident

waste, including the licensing of near surface repositories in the ChEZ, has shown that the

management problem for these wastes cannot realistically be resolved following the current

approaches due to the significant level of associated economic costs.

Solving these management problems requires:

A systematic review of the existing norms and regulations on radioactive waste

management with respect to best international practice

Improvement of the existing classification of radioactive waste with a focus on disposal

Improvement in safety assessment approaches for licensing the disposal of accident

waste, taking into account relevant safety features and site specific conditions at

disposal facility locations.

REFERENCES

[1] “Twenty-five Years after Chornobyl Accident: Safety for Future”, National Report of Ukraine,

KiM, Kiev (2011).

[2] The Strategy for Shelter Object Transformation, Cabinet of Ukraine, Kiev (2001) (in Ukrainian).

[3] RADIATION SAFETY STANDARDS OF UKRAINE, add. “Radiation Protection against

Potential Exposure Sources (NRBU-97/D-2000)”, Ministry of Health of Ukraine, Kiev (2000) (in

Ukrainian).

[4] INTERNATIONAL ATOMIC ENERGY AGENCY, Classification of Radioactive Waste, GSG-

1, Vienna (2009).

[5] INTERNATIONAL ATOMIC ENERGY AGENCY, Disposal of Radioactive Waste, SSR-5,

Vienna (2011).

[6] PROSKURA, M., et al., “Evaluating the Effectiveness of Implementing New Classification for

Radioactive Waste in Ukraine”, Nuclear and Radiation Safety 2(66) (2015) 39 (in Russian)


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PATH FORWARD TO THE REVISION OF THE FRENCH DOCTRINE OF WASTE

MANAGEMENT IN POST-ACCIDENTAL SITUATIONS: NEW CHALLENGES IN

THE LIGHT OF EXPERIENCE FEEDBACK AND UNVEILED AREAS DE

DEVELOPMENT

M.Tichauer, G. Mathieu, F.Besnus

Institute for radiological Protection and Nuclear Safety (IRSN), France


Abstract. The French doctrine on management in the post-accident phase of a nuclear accident or a

radiological situation has been forged since 2005 by the safety authority’s CODIRPA steering committee around

four guiding principles: anticipation, justification & optimization (in consistency with the ICRP) and association

of relevant stakeholders. A call for an evolution for such a doctrine in the area of decontamination and waste

management was identified in 2015 at the national level.

In the meantime, the French national public expert in nuclear and radiological risks IRSN has been studying

experience feedback from past accidents and existing polluted sites. On the one hand, IRSN has carried out

assessments on technical issues arising from national sites where radiological pollution calls for accurate

characterization, potential remediation actions and stakeholder involvement. On the other hand, IRSN, partners

and counterparts have worked on the territories affected in the Chernobyl area and off the Fukushima Dai-Ichi

NPP in order to acquire data and perspective from remediation and waste management issues (2013-2014).

Knowledge compiled form the basis of a proposal by IRSN (2015) for the improvement of the French doctrine

of post-accident management, with a specific focus on decontamination and waste management, stemming from

(i) the ability to quickly react after an accident with the aim of reassuring the neighboring population, (ii) the

overarching role of socio-economic factors in the efficiency of remediation activities in regions where a nuclear

accident may occur, and (iii) the identification of operational threshold values for such activities as well as

endpoints for waste streams, with a specific consideration for huge volumes of waste. On top of that, (iv) issues

on characterization of polluted sites (anticipation, strategy, and a specific focus on accuracy of measurements

and metrology…) are major areas of development as well, especially for (v) the unveiling of impact estimations

as close as possible to the real situation, e.g. in order to limit the consequences of population displacements.

IRSN’s proposal underlines that definition of post-accident strategies must primarily take into account the

involvement of the population to envision a possible return to satisfactory conditions of life; its involvement in

the decision making process is seen as a key factor for success of remediation actions.

Key Words: post-accident, decontamination, civil society involvement, waste management

1. Context

The French doctrine [1] related to post-accident management, forged under the supervision of

the safety authority’s CODIRPA steering committee, is laid around four guiding principles:

anticipation, justification & optimization (in consistency with the ICRP) and association of

relevant stakeholders.

Feedback experience from the massive decontamination and waste management actions

launched in the weeks following the F-1 accident in the frame of the Japanese “Act on

Special Measures” suggest that the main aspects of the French doctrine are relevant in terms

of strategic decision making.

In fact, the Japanese decontamination and waste management strategies is primarily intended

at people evacuated after the accident, whether by governmental decision on radiation



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protection drivers or by individual assessment of risks and opportunities arising from such a

situation: the need for the local population to envision a possible return to a normal life has

therefore to constitute the core principle guiding post-accident management plans which aim

primarily at establishing socially acceptable, stable and sustainable conditions for population

affected by contaminated territories. The notion of “territory project” in the CODIRPA

doctrine, which encompasses long term vision and policies, both local and national, for these

affected territories, seems therefore particularly relevant in the light of the feedback

experience.

In the same way, the CODIRPA approach, which consists in identifying the waste types and

their associated streams with consistent solutions for the evacuation of possible vast amounts

of waste, is somewhat similar to the strategy proponed by the Japanese government in its

general framework. It is also noteworthy that ambitious programs of decontamination and

induced waste management have been possible because of the intertwining of such a top-

down strategy and the pragmatic and massive distribution of technical guidelines as early as

the end of 2011 in a hundred municipalities in Japan.

IRSN therefore highlight that the Japanese feedback confirms the pertinence of approaches

and strategies of the French CODIRPA doctrine. Additional lessons learnt call for an update

of this doctrine, in order to reinforce the operational aspects of its implementation, should it

be necessary in the future.

To catch up with this objective, IRSN ongoing studies lay the basis on which a proposal [2]

for the revision of the here above doctrine was built. It stems from (i) the ability to quickly

react after an accident with the aim of reassuring the neighboring population, (ii) the

overarching role of socio-economic factors in the efficiency of remediation activities in

regions where a nuclear accident may occur, and (iii) the identification of operational

threshold values for such activities as well as endpoints for waste streams, with a specific

consideration for huge volumes of waste. On top of that, (iv) issues on characterization of

polluted sites (anticipation, strategy, and a specific focus on accuracy of measurements and

metrology…) are major areas of development as well, especially for (v) the unveiling of

impact estimations as close as possible to the real situation, e.g. in order to limit the

consequences of population displacements.

2. Time phased reactions after a nuclear accident

Since the return to satisfactory conditions of life is a key objective of a post-accident

management strategy, feedback suggests that early decontamination actions decided and

implemented directly by people on the field play a key role in strengthening reassurance

among the population. IRSN therefore considers that the identification of such early actions

(decontamination of areas where children live and play, access to public services and local

gathering commodities, etc.) is key in the anticipation of an accidental situation. The

definition of such actions which aim at opening possibilities to the re-foundation of an

acceptable day-to-day life should naturally be discussed prior with the concerned population.

In the light of the Japanese feedback, and to some extent also when it comes to the pertinence

of remediation programmes of radiologically polluted sites, the socio-economic context and

the actual needs and projects of the affected territories and their population frames post-

accident management plans, thus calling for anticipation of a possible nuclear accident,

considering the above issues.

That being highlighted, efficiency of decontamination plans should also be questioned, and

the definition of objectives should be a key aspect of anticipation as well.


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3. Anticipation

Japanese feedback suggests that both a national driver, materialized with a Law and a

framework for decontamination and waste management strategies, and a very pragmatic, “on

the field” approach, leads to notable results in terms of territory remediation. In fact, the use

of quantitative objectives related to decontamination and waste management (thresholds of

8 000 Bq/kg and 100 000 Bq/kg enshrined in the Japanese Act), associated to ambient dose

objectives (1 mSv/year in the long term) seem to be pragmatic and effective means to allow

practical decontamination works on the field. For instance, the actual use of such objectives

led to simple measurements after removal of a few centimeters of soil, where applicable.

Though, the French doctrine does not account for quantitative objective at this stage,

therefore a thorough reflexion on these could be initiated with a large participation from

stakeholders as well.

Feedback from the Fukushima prefecture and from existing polluted sites in France also

suggests that norms and standards, especially concerning radiation protection, do not fully

take into account the peculiarities of affected territories, the actual state of available waste

streams, nor explicitly the population’s concerns and expectations. The example of traditional

practices such as the use of forest products (mushrooms…) shows the importance of the

identification of quantitative restrictions that should be decided by carefully integrating their

social acceptability in the frame of a possible return to normal life strategies. In the same

way, a lack of nuance in the establishment of quantitative thresholds for the evacuation of

contaminated waste could lead to unwanted bottlenecks in existing disposal facilities,

especially for low level waste.

Moreover, IRSN notes that the difficulties faced by population who actually stayed in

contaminated territories are softened by a constant stream of information from various

authorities and most of all by their own ability to measure the impact of radioactivity on their

own health and environment. The association of both top-down strategies for remediation and

bottom-up reflexions for allowing the population to be able to understand and perform

efficient measurements could then be key to the success of remediation strategies. The role of

mediators could here be underlined too.

Finally, it is noteworthy that the absence of pre-defined evacuation solution for

decontamination-induced waste is a major limitation factor for the efficiency of

decontamination actions. Feedback shows that when a waste stream is set up and an

evacuation solution envisioned, a clear sign is given to perform decontamination actions that

play an important role in the reassurance of the population. A technical reflexion on the

ability to implement storage and particularly disposal facilities in the context of a nuclear

accident should therefore be initiated in light with the quantitative objective here above

underlined, especially considering the pre-selection of candidate sites for hosting waste

storage and disposal.

4. Characterization and impact estimations

The management of polluted sites in France relies primarily on the ability to characterize

radiological and chemical pollution but the accuracy and representativeness of such a

characterization remains a challenge. Though there are different ways of getting a good level

of confidence of the actual level of contamination in a given area (from aerial to automatic, to

on-site measurements, to alternative approaches such as geostatistics, for instance), a

thorough contextual understanding of exposition pathways, land and natural resources use,

social acceptability for population dwelling in or using patrimonial, economic, cultural, etc.


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resources offered by the affected territory is needed to guide the characterization process and

avoid inaccurate remediation plans and hiccups in their implementation.

Furthermore, it is essential to state that Japanese feedback shows that correct measurements,

have allowed to establish decontamination and waste management plans. The implementation

of such plans heavily relies on ubiquitous measurements, stressing the importance of their

accuracy. Accuracy of measurements is also key to the establishment of realistic projections

of exposition levels, especially when it comes to sketching boundaries of territories to be

evacuated. As a matter of fact, the decision to evacuate people remains one of the most

significant issues in the context of a nuclear accident.

5. Conclusion : areas of development of decontamination and waste management in

post-accident situations

Both feedback for the F-1 accident and ongoing assessments of radiologically polluted sites

and soils on the French territory suggest that different efforts can be made to upgrade the

operational aspects of the French doctrine of decontamination and waste management in post-

accident situations. All highlight the importance of civil society’s involvement in the

anticipation of such a situation, especially the population that may be affected, including

local decision-makers which may play a particular role in the reassurance of the population in

such a context. Therefore, the anticipation of such a post-accident management plan shall be

based on a large part on decentralized and associative ways, in order to integrate the socio-

economic factors which constitute key drivers for the efficiency of implemented remediation

strategies.

Relevance of management strategies relies also on technical aspects, especially those arising

from the establishment of efficient decontamination plans and waste streams, down to final

disposal facilities. Decontamination objectives and on the field threshold values then play a

significant role in the acceptability and effectiveness of such remediation programmes, as

feedback suggests.

Therefore IRSN’s roadmap encompasses two areas : on the one hand, a comprehensive

understanding of the population concerns and expectations in the context of a nuclear

accident, so that satisfactory conditions of life can be rapidly envisioned. On the other hand,

technical aspects, such as the question of the rapidity, pertinence, accuracy and

representativeness of radiological measurements, in the frame of massive waste volumes to

be managed and limited. These areas of development call for a general reflexion at different

levels, national and local, societal and technical, to help implementing remediation actions,

should a nuclear accident happen in France.

REFERENCES

[1] AUTORITE DE SURETE NUCLEAIRE, policy elements of a national doctrine for

nuclear post-accident management, Paris (2012). http://www.french-nuclear-

safety.fr/Information/News-releases/National-doctrine-for-nuclear-post-accident-

management

[2] INSTITUT DE RADIOPROTECTION ET DE SURETE NUCLEAIRE, Analyse du

retour d’expérience de Fukushima concernant la gestion des déchets contaminés en

situation post-accidentelle, Avis N°2015-00189, Fontenay-Aux-Roses (2015).

http://www.french-nuclear-safety.fr/Information/News-releases/National-doctrine-for-nuclear-post-accident-management




15


INVENTORY ESTIMATION FOR ACCIDENT WASTE GENERATED AT THE

FUKUSHIMA DAIICHI NPS

Y. Koma1,2

, D. Sugiyama 3, T. Ashida

1,2

1 Japan Atomic Energy Agency, Ibaraki, Japan

2 International Research Institute for Nuclear Decommissioning, Tokyo, Japan

3 Central Research Institute of Electric Power Industry, Tokyo, Japan


Abstract. To investigate accident waste management at the Fukushima Daiichi Nuclear Power Station, a

methodology of inventory estimation for whole waste including those generated through D&D is indispensable

and requires a calculation model. At early stage with insufficient analytical data of waste, a model should be

developed by using knowledge of past experience from literature. Uncertainty in estimation shall be decreased

based on analytical data available.

Key Words: Fukushima Daiichi Nuclear Power Station, accident waste, inventory

estimation, contamination behavior.

1. Introduction

The accident at the Fukushima Daiichi Nuclear Power Station (NPS), which is owned by

Tokyo Electric Power Co., will generate huge amount of various waste including those

already stored at the site, which requires appropriate management. In the general course of

D&D for nuclear facility, required procedure on storage and processing would be prepared in

advance [1]. However, for the case of accident waste, waste management strategy is

formulated along with their generation and planning D&D and thus, has been investigated

[2].

Waste characterization is indispensable in establishing a practical plan of waste management

[3]. Inventory estimation is important especially at early stage of investigation, and is forced

several limitations as follows;

- sampling waste for analysis is difficult due to high dose rate,

- D&D is ongoing and waste generation is hardly predicted in advance,

- development of accident events has not been clarified and estimating contamination is

difficult, and

- target nuclides for analysis were only tentatively specified according to limited

experience for waste management including disposal.

On the above premises, an inventory estimation method is discussed.

2. Calculation model for inventory estimation

Waste already generated including rubble from hydrogen explosions, felled tree, secondary

waste from the contaminated water treatment has been stored at the site. And, all buildings of

reactor/turbine and of facilities including reactor waste treatment, soil and vegetation were

contaminated. These waste and contaminated materials should be investigated for their



16

inventory. When analytical data of waste is limited, a calculation model, in which radioactive

nuclides contained in source term(s) is distributed to various materials, can be employed. A

model for waste at the Fukushima Daiichi NPS has been developed as shown in FIG. 1 [4].

The source term for contamination was assumed as the damaged fuel of unit #1 thorough #3

and radioactivity contained in the reactor buildings. The radioactivity in the source term is

distributed to several materials by way of several branchings. Therefore, the model includes

several parameters; radioactivity of the source terms, branchings and their ratios of

radioactive nuclides, such as release and leaching, waste/contaminated materials and their

classification. At early stage of estimation, calculation is made by setting parameters based

on knowledge from past accidents, and the parameters should be improved with aid of the

increasing information.

The calculated radionuclide inventory of waste involves uncertainty, which comes from

nuclear data used for source term and from supposed branching ratios. Release fraction of

radionuclides from the damaged fuel was referred to the report dealing with LWR accidents

by U.S. NRC [5] and to the evaluation on the Fukushima Daiichi NPS [6]. For example,

fraction of alkaline metals at initial release is 0.25 [5] and 0.012 [6], and this difference

should be admitted to use at the stage of insufficient analytical data. Waste inventories is

calculated by

multiplying a few

branching ratios thus, the

resulted value potentially

involves a range of

intrinsic uncertainty. For

the case of

contamination inside

reactor buildings,

transfer process by way

of both air and water

should be considered.

Chemically similar

elements are grouped

and estimated, in which

classification by NRC

[5] was initially

employed.

3. Analysis of contaminated materials and transport behaviour

When analytical data of waste is obtained, it is important to utilize it to decrease uncertainty

in waste inventory estimation. To discuss distribution of radionuclides to waste, transport

behaviour of radionuclides is quantitatively evaluated and utilized. For the purpose,

concentration ratio of a nuclide of interest to the standard nuclide was normalized with the

source term activity, which is referred to as transport ratio for convenience and utilized [7].

The transport ratio, T, presents relative transport from the supposed source term to the sample

but no information on the detail of the transport process. 137

Cs is chosen as the standard due

to its wide contamination and TCs is unity according to the definition.

Reported concentration for the accumulated water, rubble, vegetation and soil was converted

to T from the damaged fuel, and the followings were found.

– Order of T to the accumulated water was Se > I > H > Cs~Sr > Ni > Pu.

FIG. 1. A calculation model of radioactive nuclides transport from

the damaged fuel and reactors to various waste.


17

– Contamination of solid materials, such as rubble, was presumably by way of air, and

order of T was similar to that of the accumulated water except for some elements; TSr

to the accumulated water is similar to TCs, and small as approximately 10–3

to solid

materials.

– T of 3H,

14C and

90Sr to the rubble, which was generated in hydrogen explosions and

sampled inside and outside reactor buildings, were similar for each unit. 3H and

14C

might be separately transported from other fuel component.

– T to soil was affected by direction from reactor buildings.

– T to vegetation was affected by time due to redistribution in living body.

– For unit #4, which had no active fuel in the reactor, T was affected by activated

products.

From these findings, consideration on the followings should be incorporated in improving the

model of contamination.

– Contamination owing to the damaged fuel of unit #1 through #3 exhibited somewhat

different nuclide composition.

– Environmental contamination was a result of overlap of release from the 3 units.

– Dependent on material, deposit of contamination and succeeding diffusion behavior is

different.

Based on the discussion using T, element groups on contamination behaviour was arranged as

shown in TABLE 1. Number of element was increased in comparison with NRC grouping,

and category was changed; group of H–C was added, and lanthanides and actinides were got

together regardless of different stable valency. This classification can be further revised along

with accumulating analytical data.

TABLE I: CLASSIFIED (GROUPED) ELEMENTS CONCERNING TRANSPORT BEHAVIOUR.

Group Element*1

,*2

Fuel Activated products

1 Light element H C

2 Alkaline metal Rb, Cs

3 Alkaline earth Sr, Ba Ca

4 Transition metal*3

Mo, Tc, Ru, Rh, Pd, Ag, Sn, Sb Mn, Fe, Co, Ni

5 Chalcogen Se, Te

6 Rare earths, actinides*3

Y, Zr, Nb, La, Ce, Pr, Nd, Pm, Sm,

Eu, Gd, U, Np, Pu, Am, Cm

7 Halogen Br, I Cl

8 Noble gas Kr, Xe

*1 Italic element name denotes no analytical data available so far.

*2 Branching ratio may different for some contaminated material dependent on source term;

fuel and activated products.

4. Decreasing uncertainty in estimation

Uncertainty accompanied with branching ratio will be successively decreased by utilizing

analytical data. Secondary waste from contaminated water treatment is highly irradiating and

difficult in sampling. For the case, data of feed and treated water can be used to estimate

decontamination factor of treatment by calculating difference of concentration and it is

converted to branching ratio. And, transport ratio is also useful to improve branching ratio of

discharge and transport; comparison between those of initial and revised as shown in FIG. 2.

Branching ratio of Cs, Sr and Se increased from the initial values, whereas width of


18

uncertainty decreased. For transition metal ant TRU element, values decreased to reproduce

actual behaviour.

As D&D progresses, estimation of waste

generation will be improved by accumulating

analytical data and clarifying development of

accident and, consequently, uncertainty of

inventory estimation can be decreased along with

improved branchings and their ratios, and will

result in revised element classification and

statistical treatment of data.

5. Conclusion

Inventory estimation for waste generated in the

severe accident of reactor was discussed. At early

stage, in which development of accident is not

clear and analytical data is insufficient, it is

important to make a calculation model, which

describes transport of radioactive nuclides from source term to waste by using knowledge

from past accidents. Once analytical data is collected, uncertainty in estimation decreases by

improving model parameters; branchings and their ratios. As D&D progresses, waste

generation will be well predicted, analytical data will be accumulated and thus, it is expected

to decrease uncertainty in inventory estimation.

6. Acknowledgment

This paper includes the result from the work subsidized to IRID/JAEA by METI.

REFERENCES

[1] INTERNATIONAL ATOMIC ENERGY AGENCY, “Decommissioning of Facilities,”

IAEA Safety Standards Series No. GSR Part 6 (2014).

[2] SASAKI, T., et al., Research and Development Activities for Cleanup of the Fukushima

Daiichi Nuclear Power Station, 2012 Materials Research Society (MRS) Fall Meeting,

November 25–30, 2012, Boston, MA, USA (2012).

[3] INTERNATIONAL ATOMIC ENERGY AGENCY, Strategy and Methodology for

Radioactive Waste Characterization, IAEA-TECDOC-1537 (2007).

[4] OECD/NEA/RWMC, Management of Radioactive Waste after Accidents, Report of an

NEA Expert Group on Fukushima Waste Management and Decommissioning R&D (in

press).

[5] SOFFER, L., et al., Accident Source Terms for Light-Water Nuclear Power Plants,

NUREG-1465 (1995).

[6] JAPAN NUCLEAR ENERGY SAFETY ORGANIZATION, JNES-RE-2011-0002

(2011).

[7] KOMA, Y., et al., Radioactive Contamination of Several Materials Following the

Fukushima Daiichi Nuclear Power Station Accident, Nuclear Materials and Energy,

http://dx.doi.org/10.1016/j.nme.2016.08.015 (2016).

FIG. 2. Improvement of branching ratio of

transport to the accumulated water

(leaching fraction) from the initial values.

1.E-10

1.E-08

1.E-06

1.E-04

1.E-02

1.E+00

Bra

nch

ing

ra

tio

2014

不確実性ケース(最小値)

参照ケース(平均値)

No

ble

gases

Ha

log

en

s (

1)

Halo

ge

ns (

2)

Alk

ali

Me

tals

Tellu

riu

m G

roup

Ba

riu

m, S

tron

tiu

m

No

ble

Me

tals

Lan

tha

nid

es

Ce

riu

m G

rou

p

Initial from literature Uncertainty for concern Reference uncertainty


19


SAFETY ASSESSMENT APPROACH FOR DECISION MAKING RELATED TO

REMEDIAL MEASURES AND RADIOACTIVE WASTE MANAGEMENT

N. Rybalka1, S. Kondratyev

2, Z. Alekseeva

2

1State Nuclear Regulatory Inspectorate of Ukraine (SNRIU), Kyev, Ukraine

2State Scientific Technical Center of Nuclear and Radiation Safety (SSTC NRC), Kyev,

Ukraine


Abstract. The issue of management with “Legacy” (“historical”) facilities and sites for radioactive waste

management has been nowadays widely raised in many countries. The approach of remediation measures

implementation is supported also by IAEA Safety Standards including BSS GSR part 3. In Ukraine the main “Legacy” (“historical”) radioactive waste storage/disposal facilities from the past practices

are identified as [1]:

- Conserved radioactive waste disposal facilities filled to 1996 for radioactive waste, including disused sealed

radiation sources, from industries, science and medicine – so called Radon-type facilities;

- Located in Exclusion Zone the radioactive waste management facilities developed in the first years after

Chernobyl accident to store and localize the large amount of accidental radioactive waste (as primary post-

accident intrusion measures).

The decisions of remedial measures for these facilities including intervention, retrieval, re-disposal,

stabilization, preservation should be made according to supporting safety reassessment.

Key Words: “legacy” facilities, “historical” radioactive waste, remedial measures, safety

assessment

1. “Legacy” Radon-type Facilities

1.1.“Legacy” Radon-type Facilities Management Decisions

In the years of sixties of 20th

century when the activity of nuclear energy use in the industries

and science had been significantly developed the level of understanding of the radioactive

waste management safety and radiation protection was quite poor. The issue of long-term

safety of radioactive waste disposal practically had not been analyzed. In former USSR

disposal facilities for radioactive waste, including disused sealed radiation sources

(hereinafter – DSRS), from industries, science and medicine were constructed and operated

according to the typical standard design developed in compliance with the regulatory

documents in force at that time. These facilities known as so-called Radon-type facilities

were located at the sites of five Regional State Specialized Enterprises of radioactive waste

management (hereinafter – RSSE RWM). Nowadays these enterprises provide the activity of

maintaining, control and monitoring of these “legacy” facilities which had been filled and

conserved during the previous period (before 1996).

The Law of Ukraine "About State Purposeful Ecological Program for Radioactive Waste

Management" envisages remedial measures including removal of radioactive waste from

conserved radioactive waste disposal facilities filled before 1996, conditioning and packaging

of this radioactive waste for further transfer for long-term storage or disposal in facilities on

the Vector Complex Site in the Chernobyl Exclusion Zone [2]. Further decommissioning



20

(dismantling, demolition, decontamination, removal) of the facilities and further remediation

of sites also should be foreseen. In the same time, according to the existing economical

situation in the country such measures can not be implemented in a short time frame for all

Radon-type “legacy” facilities. That is why the appropriate planning and optimization of

remedial measures for particular facilities and sites should be made on the basis of safety

reassessment.

1.2. Safety Reassessment for Remedial Decisions of “Legacy” Radon-type Facilities

According to the conditions of the issued licenses, RSSE RWM have started the safety

reassessment of radioactive waste facilities on their sites. The safety reassessment includes

also the conserved “legacy” facilities that were operated in the previous period. Based on the

reassessment results it is expected that:

- the time frames, while Radon-type “legacy” facilities contained “historical” waste on the

site of particular RSSE RWM could be considered safe, should be justified;

- particular remedial decisions on radioactive waste removal and/or other remedial

measures (stabilisation or strengthening of engineering barriers, monitoring system

improvement, etc.) will be planned (time frames, sequence between sites and facilities

according to their assessed safety status and/or proven urgency).

The SNRIU in cooperation with European and Ukrainian TSOs has developed under the EC

cooperation Project INSC U3.01/08 the “Guideline on Safety Reassessment of Existing

Radioactive waste Disposal Facilities and Criteria for Decision making on Further Activities

at These Facilities” [3]. The Guideline as a recommendation was submitted by regulatory

body to the RSSE RWM as a supporting methodology and approach for the implementation

of the safety reassessment.

According to the Guideline the safety reassessment is aimed at:

analysis of current safety status considering existing conditions of facilities, to what

extend Radon-type “legacy” facilities comply with the safety criteria, defined by

Ukrainian regulatory documents in force and IAEA Safety Standards;

definition of feasible measures to increase the safety level of facilities for certain time;

definition of possible options and appropriate measures of radioactive waste removal and

further management;

making decisions about terms, sequence, options of radioactive waste removal from

“legacy” facilities (e.g. urgent, after certain time, complete/partly) and scope of the other

remedial measures;

safety justification and design of particular technical decisions.

Safety reassessment of RWDS covers following issues:

characterization of facility’s site, radioactive waste in facilities, engineered barriers of

facilities;

assessment of radiological impacts for population;

assessment of occupational doses for removal of radioactive waste from facilities and

related measures, as well as for maintenance of facilities in case of leaving radioactive

waste in place;

measures for control of facilities conditions, environmental monitoring and maintenance.

Safety reassessment is carried out iteratively in the following sequence:

Stage 1. Reassessment to determine necessity for urgent removal of radioactive waste

from facilities and making respective decisions.


21

Stage 2. Reassessment to determine necessity for delayed removal of radioactive waste

from facilities and making respective decisions.

Stage 3. Reassessment to determine expediency of removal of radioactive waste from

facilities or their leaving in place and making respective decisions.

2. Management of “Legacy” facilities of the Chernobyl accident origin

Within the Chernobyl Exclusion Zone territory there are legacy radioactive waste

management facilities developed in the first years after Chernobyl accident to store and

localize the large amount of accidental radioactive waste (as primary post-accident intrusion

measures). Now these facilities require remedial actions such as measures on stabilization,

safety upgrading or waste retrieval (partly retrieval), liquidation. The scale and content of

remedial actions should be planned and implemented on the basis of safety assessment.

These facilities are so-called:

- Radioactive waste disposal points (RWDP) “Pidlisniy” and “Chornobyl NPP Stage

III” which have some engineering barriers and contain the most hazardous intermediate-level,

high-level and long-lived radioactive waste and even fuel-contain materials (in RWDP

“Pidlisniy”). Most of this radioactive waste because their activity belongs to disposal in

geological repository. The RWDP facilities not considered as final disposal. So this

radioactive waste sooner or later needs to be removed. Before it could be possible (the

appropriate management techniques and geological repository are in place) the measures for

stabilization, monitoring, surveillance to ensure the safety level and barriers containment

properties are needed.

Prior to making decisions on further management of long-lived and high level radioactive

waste in RWDP “Pidlisniy”, RWDP “Chornobyl NPP Stage III”, the designs for their

stabilization and safety improvement are under implementation. The Operator of these

facilities envisages safety assessment of these facilities to substantiate period of further safe

radioactive waste storage taking into account engineering barriers improvement and

stabilization.

- Temporary radioactive waste localization points (TRWLP) - trenches/clamps without

engineering barriers located at territories adjacent to ChNPP with total area of 100 000 m2.

There was localized mostly low-level radioactive waste such as contaminated materials, soil,

rubbish, structures, demolished country houses, etc. arisen from immediate decontamination

works after accident. The total amount of TRWLP trenches/clamps is estimated as 800-1000.

The remediation measures for TRWLP may include removal and redisposal of radioactive

waste from trenches/clamps or maintenance/upgrading of some trenches/clamps without

removal of radioactive waste [2].

Currently the primary decisions of TRWLP trenches/clamps liquidation measures are taken

for the most dangerous of them according to qualitative judgments (flooded, adjacent to water

sources, located at the places of current construction activity). The corresponding technical

decisions include excavation and redisposal of radioactive waste, examination and

recultivation of the site.


22

2.1 Safety Assessment for Remedial Decisions of TRWLP in Exclusion Zone

Regarding radioactive waste placed in TRWLP of the Exclusion Zone, decisions on

additional intervention measures taken to decrease their hazard (retrieval, re-disposal,

stabilization, preservation, etc.) will be made after investigations of radioactive waste in

TRWLP trenches/clamps and their territory, and according to the safety assessment results.

Appropriate investigations and assessment currently are in progress under EC INSC

U4.01/10-D industrial project.

The SNRIU in cooperation with European and Ukrainian TSOs developed under EC INSC

U3.01/10 project the “Guideline for safety assessment of emergency RAW temporary

localization (storage) sites in Chernobyl exclusion zone” [4]. The Guideline as a

recommendation was submitted by regulatory body to the Operator as a supporting

methodology and approach for the implementation of the safety assessment of TRWLP.

The long-term safety assessment of TRWLP is based on the considerations that access of

public to Exclusion Zone will be prohibited as long as its memory exists, no population is

residing there, only restricted activities are carried out and access of staff of nearby facilities

to the TRWLP territory is limited. According to the President Order [5] the zone of special

industrial use considered forever unsuitable for living should be established within the

Exclusion Zone.

According to the Guideline safety assessment approach does not include systematic detailed

assessment of condition of each individual trench/clamp. Cumulative radiological impact of

RWTLP on humans without detail assessments for each specific point should be assessed.

Taking into account uncertainties, related to incompleteness of study of TRWLP sites, RW

characteristics, as well as insufficient knowledge about possible evolution and extreme

change of conditions in the future, iterative approach of assessment is promoted in three

stages of assessment.

Stage 1 includes

- systematic identification, collection, systematization and validation of available data;

- develop conservative conceptual model based on existing initial data of each TRWLP;

- development of scenarios and conservative calculations of radiological impacts;

- comparison of the results of calculations with ranking between TRWLP.

Stage 2 goes to upgraded assessment of radiological impacts of TRWLP if at Stage 1

conservative assessments of radiological impacts exceed admissible criteria (or if uncertainty

of the results obtained does not allow to make a conclusion about TRWLP safety level).

Stage 2 assessment is based on additional studies of TRWLP, upgraded model, scenarios and

carrying out more realistic calculations of radiological impacts. As a result ranking the

TRWLP in order to decide on some remedial measures to take after comparing the results of

the safety assessment with the dose criteria.

Stage 3 is carried out if at Stage 2 upgraded assessments of radiological impacts exceed

admissible criteria. Stage 3 includes:

- Determination of needs for implementation of remediation measures based on

ALARA principle.

- Overview of applicable remedial measures;

- Develop of remedial measures at TRWLP;


23

- Adjustment of the model, scenarios and carrying out detail assessments of

radiological impacts taking into account applicable remedial measures;

- Comparison of the results of calculations with criteria of radiological impacts.

- The radiological criteria, safety assessment approach and decisions making on

remedial measures for existing TRWLP could be based on the consideration of the

“existing exposure situation” [6]. The radiological criteria for the public

recommended as:

- 0,3 mSv on the border and outside of the Exclusion Zone;

- Acceptable value of permissible dose in the range of 1-20 mSv/year based on ALARA

principle at the TRWLP territory including the period after possible termination of

Exclusion Zone existence and loss of memory in long-term period;

- Total annual exposure dose for public from all sources of the Exclusion Zone at any

time shall not exceed 1 mSv (outside Exclusion Zone, presuming existence of

restricted approach territory).

REFERENCES

[1] National Report of Ukraine on Compliance with Obligations under the Joint Convention

on the Safety of Spent Fuel Management and on the Safety of Radioactive Waste

Management, Kiev, 2014.

[2] The Law of Ukraine № 516-VI from 17/09/2008 "About State Purposeful Ecological

Program for Radioactive Waste Management".

[3] Project INSC U3.01/08 (UK/TS/39). Task 1 Report. RISKAUDIT Report № 1773.

Luly 2013.

[4] INSC Project U3.01/10. Subtask 3.1. “Guideline for safety assessment of emergency

RAW temporary localization (storage) sites in Chernobyl exclusion zone”, revision 3

September 2015.

[5] The Order of President of Ukraine № 141/2016 of 28.04.2016, Kyev, Ukraine.

[6] Radiation Protection and Safety of Radiation Sources: International Basic Safety

Standards / General Safety Requirements, IAEA Safety Standards Series, No. GSR Part 3,

Vienna, 2014.


24


POTENTIAL NANO-FE/CA/CAO COMPOSITE ENABLED ENVIRONMENTAL

REMEDIATION TECHNOLOGIES FOR RADIOACTIVE WASTE

S.R. Mallampati1, Y. Mitoma

2, C. Simion

3

1Department of Civil and Environmental Engineering, University of Ulsan, Ulsan 680-749,

Republic of Korea 2Department of Environmental Sciences, Prefectural University of Hiroshima, Shobara, Japan

3Politehnica University of Bucharest, Department of Organic Chemistry, Bucharest, Romania


Abstract. Radioactive cesium (137

Cs) contaminated soil obtained from Fukushima was treated with nano-

Fe/Ca/CaO/[PO4], approximately 27.3 wt% of magnetic and 72.75% of non-magnetic soil fractions were

separated. The highest amount of entrapped 137

Cs was found in the lowest weight of the magnetically separated

soil fraction (i.e., 80% in 27.3% of treated soil). Results show that 137

Cs either in the magnetic or non-magnetic

soil fractions was 100% immobilized. While, 137

Cs contaminated fly ash (containing an initial 14,040 Bq kg−1

137Cs concentration) obtained from burning wastes from Fukushima were reduced to 3,583 Bq kg

−1 after

treatment with nanometallic Ca/CaO methanol suspension. Furthermore, both soil/ash content and eluted

solution concentrations of 137

Cs were much lower than the Japanese Ministry of the Environment regulatory

limit of 8,000 Bq kg−1

and 150 Bq L−1

respectively. The nano-Fe/Ca/CaO/[PO4] enabled treatment can be

considered as environmental remediation of radioactive waste.

Key Words: Radioactive cesium; Nano-Fe/Ca/CaO composite; Remediation; Contaminated

soil/fly ash.

1. Introduction

In Japan, the major environmental concern on the radioactive cesium (137

Cs) deposition and

its contamination due to the emission from the Fukushima Daiichi Nuclear Power Plant

showed up after a massive quake on March 11, 2011. Many fallout of 137

Cs generated a large

amount of various wastes (soil, water, ash, sediment/wide range of concentration).

Fukushima prefecture exceed the government’s safe limit of 2500 Bq/kg of cesium137

in soil.

According to the Japan science ministry about 8 percent of the country’s land has been

contaminated with levels higher than 10,000 Bq/m2 of cesium

137 a threshold that Japan’s

science ministry defines as affected by a nuclear accident [1,2]. On the other hand, by the end

of March 2012, incineration ash containing 100,000 to 140,000 becquerels per kilogram (Bq

kg−1

) of (137

Cs) was recorded. High levels of 137

Cs are also present in incineration ash from

normal garbage [3,4]. Temporary disposal sites for incinerated ash containing 137

Cs are

rapidly filling up. No alternative landfills are available. During and after the 30 years it takes

for 137

Cs to decay by half, each time it rains, 137

Cs deposited will be washed down to where

people live [4]. Therefore, the 137

Cs extraction and immobilization in contaminated soil/ash is

recognized to be one of the most difficult problem solved by taking advantage of suitable

technologies [1-4]. For radiological contamination, the separations of interest may be the

removal of the radioactive species from its host matrix (i.e., an extraction), or it may involve

processes such as fixation or stabilization, in which the radioactive material is separated from

any mobilization and/or transport pathways so that the risk it poses is reduced or eliminated

altogether by preventing it from being made available to a receptor [1-4]. In recent years,



25

nanotechnology has risen to the forefront and the new properties and enhanced reactivities

offered by nanomaterials may offer a new, low-cost paradigm to solving complex

environmental and engineering problems [1-4]. This study assessed the synthesis and

application of nano-Fe/Ca/CaO composite/ nanometallic Ca/CaO methanol suspension for

use as a separation/extraction and immobilizing treatment of 137

Cs contaminated soil/fly ash.

2. Materials and methods

2.1. Soil sample and fly ash collection and nano-Fe/Ca/CaO and nanometallic

Ca/CaO methanol suspension preparation

Radioactive Cs, contaminated soil samples collected from areas affected by the Fukushima

Daiichi NPP accident. On the other hand, radioactive cesium (137

Cs) contaminated fly ash

samples were collected from disaster stricken area of Iwaki city, Fukushima prefecture, and

were used as such for treatment experiments. Nano-Fe/Ca/CaO/Nano-Ca/CaO (dry system)

systems were prepared respectively by a ball milling process [2,3]. The suspension of

nanometallic Ca/CaO methanol was prepared by magnetic stirring the mixture of 10 g

nanometallic Ca/CaO and 50 g methanol for 30 min. After stirring, the samples were kept for

10 minutes to settle and the supernatant of methanol suspension contains nanometallic

Ca/CaO (about 25-30%) (nMCaS) was separated and used for treatment experiments.

2.2.Radioactive cesium (137Cs) contaminated soil and fly ash treatment

Radioactive cesium (137

Cs) contaminated soil samples (10 g) were mixed with 0.5 g nano-

Fe/Ca/CaO, or 0.5 g NaH2PO4 ([PO4]). These mixtures were ground in a 500 mL ceramic

tumbling mill, along with 10 ceramic balls (10-mm diameter) for 2 h at 100–150 rpm in open

atmosphere. After the treatment, the samples were analyzed and compared with the untreated

soil samples. The magnetic and non-magnetic soil fractions were separated easily through a

balance of magnetic forces and gravity. On the other hand, about 1 g of 137

Cs contaminated

fly ash and 3 g of extraction solvent (i.e. nMCaS) were thoroughly mixed for 4 h at 200 rpm

followed by filtered through a 0.45 μm membrane filter. The total 137

Cs concentrations in

extracted solutions and/or soil/fly ash residues were measured using a high-purity germanium

detector with gamma-ray spectrometry (GMX-20P4-70; Seiko EG & G), according to the

standard Ministry of Education, Culture, Sports, Science & Technology Japan (MEXT)

method [2,3].

3. Results and discussion

3.1.Radioactive cesium (137Cs) contaminated soil treatment

Once the principal steps for the 133

Cs immobilization/separation process using artificially

contaminated soil with 133

Cs have been established (Fig. 1a), the procedure was applied to

actual soil samples contaminated with 137

Cs obtained from Fukushima prefecture, the region

most affected by emitted radionuclides. The treatment process is presented as a simplified

flowchart in (Fig. 1b). The initial radioactive cesium (137

Cs) concentration in contaminated

soil was 6,201 Bq kg−1

. The radioactive cesium concentrations in various soil size fractions

were measured. As presented in Table 1, a high concentration of 137

Cs was found in the fine

fraction of soil (i.e. <0.125 mm size 4446 Bq kg−1 and >7 mm size contains only 436 Bq

kg−1), which corresponds to about 40 wt% (i.e., 0.5 to < 0.125 mm size) fine fraction of soil

contained about 67% of total radioactive cesium (137

Cs) [3]. Any isotope of cesium, whether

radioactive or stable, will preferentially adhere to the surfaces of soil and other organic


26

particles. After treatment with nano-Fe/Ca/CaO/[PO4], the sum of 137

Cs allocations among

the magnetic and non-magnetic soil fractions were, respectively, 5,419 and 1,438 Bq kg−1

(much lower than the regulatory limit of 8,000 Bq kg−1

established by the Japanese Ministry

of the Environment) [3]. The best results were obtained using the nano-Fe/Ca/CaO/[PO4]

dispersed mixture, where approximately one-fifth of the contaminated soil, containing

approximately 88.1% of the amount of Cs, was magnetically separable. Cs, either in the

magnetic or non-magnetic soil fractions, was immobilized completely because no leaching

was detected. This method concentrates radio-isotopes into a smaller volume of insoluble

material that is easy to store.

FIG. 1. A possible pathway for separation/immobilization of (a) stable cesium (133

Cs) and (b)

radioactive cesium (137

Cs) with the nano-Fe/Ca/CaO/[PO4] treatment [3].

TABLE 1. RADIOACTIVE CESIUM CONCENTRATIONS IN VARIOUS SOIL SIZE

FRACTIONS

Soil size

(mm)

Soil ratio (Wt%)

137Cs

Concentration (Bq kg−1)

Soil Ratio (Bq %)

>7 4.8 436 0.7

02-7 8.6 724 3.3

2-0.5 46.6 1227 30.1

0.5-0.25 17.1 1893 17.8

0.25-0.125 10.5 3187 18.3

<0.125 12.4 4446 30.8

Total 100

100

3.2. Removal and immobilization of radioactive cesium (137Cs) in fly ash

Once established the principal steps for the 133

Cs removal/immobilization process using

artificially contaminated fly ash with 133

Cs (Fig. 3a), the present methodology applied to

actual radioactive cesium (137

Cs) contaminated fly ash obtained from a municipal waste

incinerator from Fukushima prefecture, the region most affected by radionuclides emitted

after the Fukushima Dai-ichi Nuclear Power Plant accident. The first extraction of 137

Cs from

fly ash was carried out using methanol to remove the large amount of 137

Cs followed by the


27

second step treatment of fly ash with n-MCaS to further remove the residual (137

Cs), i.e.

trapped inside the fly ash particles. During the process, the dispersed mixture of nano-

Ca/CaO is act as a sealing agent for fly ash particles that facilitate the filling of channels and

pores created by methanol. The process is presented as a simplified flowchart in (Fig. 3b).

Thus, from an initial concentration 14,040 Bq kg−1

in the Fukushima area fly ash, after the 1st

methanol extraction, the remaining total 134

Cs and 137

Cs concentration was only 3,583 Bq

kg−1

, representing 74.5% removal efficiency. After the 2nd

extraction with n-MCaS, elution

test conducted on the treated fly ash gave only 100 Bq L−1

total 134

Cs and

137Cs, a lower value

than the regulatory limit of 8,000 Bq kg−1

and 150 Bq L−1

, respectively, established by the

Japanese Ministry of the Environment. This study demonstrates that nanometallic Ca/CaO

methanol suspension extraction effectively extracted and immobilized radioactive cesium

from contaminated fly ash [2]. These results suggested that nanometallic Ca/CaO methanol

suspension extraction can be regarded as a sustainable remedial strategy for extraction and

immobilization of 134

Cs and 137

Cs from actual contaminated fly ash.

FIG. 3. Schematic representation of possible mechanisms determining the (a) Stable cesium (133

Cs)

(b) radioactive cesium (134

Cs and 137

Cs) extraction and immobilization in contaminated fly ash with n-

MCaS extraction [2].

4. Conclusions

In this study, nano-Fe/Ca/CaO composite has act as a promising agent for simultaneous

removal and immobilization of radioactive cesium (137

Cs) from contaminated soil/fly ash.

The highest amount of entrapped 137

Cs was found in the lowest weight of the magnetically

separated soil fraction (i.e., 80% in 27.3% of treated soil). Both soil/ash content and eluted

solution concentrations of 137

Cs were much lower than the Japanese Ministry of the

Environment regulatory limit of 8,000 Bq kg−1

and 150 Bq L−1

respectively. These results

suggest that simple treatment with nano-Fe/Ca/CaO/[PO4]/ nanometallic Ca/CaO methanol

suspension is a highly potential amendment for the remediation of radioactive cesium-

contaminated soil/fly ash.

5. Acknowledgment

We are thankful to National Research Foundation of Korea (NRF) and Japan Society for the

Promotion of Science (JSPS), for providing financial support for this study through the

collaborative research project (2016K2A2A4003740).


28

REFERENCES

[1] Mallampati et al., “High immobilization of soil cesium using ball milling with nano-

metallic Ca/CaO/NaH2PO4: implications for the remediation of radioactive soils”,

Environ. Chem. Lett. 10 (2012) 207.

[2] Mallampati et al., “Preferential removal and immobilization of stable and

radioactivecesium in contaminated fly ash with nanometallic Ca/CaO

methanolsuspension”, J. Hazard. Mater. 279 (2014) 59.

[3] Mallampati et al., “Solvent-free synthesis and application of nano-Fe/Ca/CaO/[PO4]

composite for dual separation and immobilization of stable and radioactive cesium in

contaminated soils”. J. Hazard. Mater. 297 (2015) 82.

[4] Mallampati et al., “Dynamic immobilization of simulated radionuclide 133Cs in soil by

thermal treatment/vitrification with nanometallic Ca/CaO composites”, J. Environ. Radio.

139 (2015) 124.


29


THE TREATMENT OF HIGHLY RADIOACTIVE LIQUID WASTE ORIGINATING

FROM A SEVERE ACCIDENT AT A VVER 440 NPP

V. Havlová1. V. Brynych

1, L. Szatmáry

1, P. Franta

1, M. Gajdoš

2

1ÚJV Řež, a.s., Husinec, Czech Republic

2Slovenské elektrárne, a.s., Bratislava, Slovak Republic


Abstract. The paper provides a summary of a design concept for a modular unit that can be used for the

treatment of large volumes of radioactive waste produced as a consequence of a severe accident at a VVER 440

NPP. The concept was designed as one of a number of post-accident measures following the Fukushima Daichi

accident. The modular unit is made up of three main parts: a sorption unit, a vitrification module and a gaseous

contaminant capture module.

Key Words: severe accident, liquid radioactive waste, safety measures, radionuclide waste

1. Introduction

Requirements surrounding specific post-accident recovery measures were tightened

significantly following the Fukushima Daichi accident. One of the main challenges concerns

the treatment of large volumes of highly radioactive waste water generated during the severe

accident mitigation. Therefore UJV Řež in the cooperation with Slovenské elektrárne, a.s. has

started a review process of the concept and design of radioactive contaminated water

treatment device, dedicated to the VVER 440 NPP severe accident [1]. The contaminated water treatment apparatus for use in case of damage sustained by the core

of VVER-type reactors is being developed as an integral part of a programme aimed at the

implementation of post-Fukushima safety measures.

2. Design basis

The development of a modular system to be employed for the purpose of liquid radioactive

waste processing commenced at UJV Rez, a.s. as early as in the late 1980s [12]. With

concern to the post-Fukushima safety measures presently, previous experience has been

subjected to re-evaluation and extended by means of the consideration of information

concerning both the Three Mile Island and Fukushima Daichi accidents and complemented

with short-term experimental research in order to propose the treatment procedure described

further.

2.1.Modular system requirements

The decontamination apparatus must ensure the following requirements:

1. A decrease in radionuclide concentration to the required limit levels or to the levels

required for water re-use for reactor cooling purposes following the relevant time

period

2. Vitrification of spent sorbent to the desired form for temporal or final

storage/disposal.



30

3. Decrease of specific radioactivity of gaseous emissions from decontamination

apparatus to levels required for NPP air conditioning.

4. Decontamination of approximately 10 m3 liquid waste every 24 hours.

5. Equipment of the apparatus according to regulations for nuclear waste handling -

manipulators, shielding etc.

6. The unit has to consist of several mobile modules so it can be fast assembled on

required place. It will be easily scalable if necessary.

2.2.Solution composition

The volume of treated water is expected to be approximately 10 000m3 assuming that the

treatment of the contaminated water commences 6 months following the successful

stabilization of the progression of the accident based on the response of the Emergency

Response Organisation (ERO) according to plant-specific severe accident management

guidelines (SAMG). Current coolant volume limitations are related to SAMG strategies

adopted with concern to the long-term removal of residual heat; however, the volume of

treated water can be increased by means of the scaling up of the apparatus.

It is presumed that the major contaminants within the coolant solution will consist of

radioisotopes of caesium and strontium together with actinides. The total activity level

following an NPP severe accident is assumed to be in the range of 1015

- 1017

Bq for beta and

gamma radionuclides and 1013

-1014

for alpha radionuclides; 6 months after the accident, the

activity level will have decreased by one order. The composition of the accident coolant

water can be approximated at: H3BO3 15g/l, N2H4 0.8g/l and KOH plus NaOH together up to

3.3g/l, leached from the walls of the NPP: K approx. 0.2g/l and Ca approx. 0.3g/l and

mechanical impurities up to 0.7mm in diameter 0.1 – 0.2g/l.

3. Modular treatment system

The modular treatment unit concept was proposed on the basis of the requirements identified

above and consists of three basic modules - the sorption module, vitrification module and the

gaseous contaminant capture module (see Fig. 1). The design of the apparatus further

includes transportation and manipulation machinery, a cooling system, a decontamination

module and shielding.

3.1. Sorption module

Individual vessels are utilized for both sorption and vitrification purposes; therefore, the

sorbent is mixed with a glass substrate prior to the initiation of the sorption process thereby

allowing for vitrification following sorption without the need for further sorbent handling.

The disadvantage of this approach, however, is that the sorbent dilates by up to 40%

following vitrification leading to the creation of undesirable free space within the vessels.

Two vessel sizes were considered - 25 and 50 litres. Moreover, the vessels contain an

integrated system that allows solution flow from the top of the vessel to the bottom and back

to the top towards the next vessel, thus providing for the prevention of the leakage of the

radioactive solution from the vessel even in case of tube rupture.

Various types of inorganic sorbents have been tested in previously conducted studies (e.g. [3,

4, 5]) in order to determine sorption properties. Sorption experiments have been performed

with respect to solutions containing H3BO3 and NaOH or KOH in order to model the various

waste solutions that can be presumed to originate from a severe accident. It was determined

that zeolite sorbents exhibit the best sorption properties for caesium and, subsequently,


31

mordenite was selected as the best sorbent type for this element. It was further determined

that clinoptilolite exhibited the best sorption affinity for strontium. Furthermore, it was

discovered that zeolite sorption ability can be improved via modification using either

hexacyanoferrate (with respect to caesium sorption) or titanium (strontium sorption).

However, it will be necessary to devote further research to sorbent efficiency so as to ensure

future availability in large volumes at an acceptable price. Importantly, however, long-term

storage also leads to sorbent degradation and changes to sorbent properties. In either case, it

is advisable that an up-to-date sorbent database accompanied by information on local market

availability be maintained at respective NPPs.

FIG. 1. Modular decontamination system concept.

Following the pre-treatment stage, the decontaminated water is re-directed to a control tank

and, if deemed necessary, towards an alpha treatment vessel, the function of which consists

of the treatment principally of alpha radionuclides by means of precipitation with either

ferrous hydroxide or hydrated titanium dioxide. Further research is required concerning the

optimisation of the sorption materials employed so as to ensure that the activity of radioactive

wastes is acceptable for transport.

3.2. Vitrification module

After flowing through the sorption module, the solution is filtered and stored prior to standard

liquid waste treatment or further utilized for nuclear reactor cooling purposes and the spent

sorption vessels are further vitrified in the vitrification unit with respect to which low

temperature glass with the required leaching levels is recommended (up to 1050ºC; the

release of gaseous Cs is expected to be less than 1%). However, it is recommended that the

transportation and storage of the vessels and resulting vitrified product be subjected to further

detailed study due to the expected high activity levels of 1010

-1013

Bq.

3.3. Gaseous contaminant capture module

Gases from the vitrification units are captured and returned to the first sorption vessel;

decontaminated gases are then released into the NPP’s air conditioning system (see Fig.1)


32

4. Conclusions

The modular treatment system described above takes the form of a design concept only; there

remains a very low probability of the occurrence of severe accidents at NPPs. The actual

construction of such a modular system and its introduction to NPPs will have to be preceded

by additional engineering research. Moreover, the long-term storage of such a device and the

required sorbent materials may well lower the level of effectiveness due to ageing and

potential damage over time. Notwithstanding, requirements concerning the implementation of

safety measures and overall preparedness coupled with the substantial costs associated with

further research open a window of opportunity for potential international cooperation

between nuclear operators. In order to comply with requirements set in place following the

Fukushima Daichi accident, all nuclear plant operators should be able to have such a device

ready for deployment within a specified time following the occurrence of a severe accident.

Therefore, a potential proposal for future cooperation concerning the development of a waste

treatment modular device is set out below:

VVER + EU NPP operator joint programming and potential research

The preparation of the detailed design of a treatment module the construction of

which would be feasible during 6 months following an NPP severe accident

Sorbent database that would be subject to regular updating and which would fully

consider the availability of large volumes within acceptable cost limits

Further research on the vitrification and transportation of treatment products

5. Acknowledgement

The study was conducted with the support of Slovenské elektrárne, a.s.

REFERENCES

[1] FRANTA P., HAVLOVÁ V., SZATMÁRY L., BRYNYCH V., AND SÁZAVSKÝ P.

Zpracování velkých objemů kontaminované vody po havárii na JE s reaktory VVER 440.

Koncepční studie přepracování kapalných odpadů a nakládání s upravenými RAO. (in

Czech). ÚJV Řež, a.s. Report, ÚJV Řež a.s., Řež (2014).

[2] FRANTA P, VOJTĚCH O., KUČA I., KEPÁK F., NENIČKA P., STUCHLÍK S.,

SCHULA M., AND ŠTOLA M. Příprava vývoje a výroby mobilních zařízení pro

zneškodňování atypických radioaktivních odpadů. ÚJV Řež, a.s. Report, ÚJV Řež a.s.,

Řež (1991).

[3] VEJSADA J., HRADIL D., ŘANDA Z., JELÍNEK E. AND ŠTULÍK K. Adsorption of

cesium on Czech smectite-rich clays—A comparative study. Appl. Clay Sci., vol. 30, no.

1, pp. 53–66, Aug. 2005.

[4] FRANTA P. AND DUŠKOVÁ D. Aktualizace navrhovaného postupu dekontaminace

kapalných havarijních RAO”, ÚJV Řež, a.s. report, ÚJV Řež a.s., Řež (1993).

[5] FRANTA P. Návrh a bezpečnostní aspekty zařízení pro zpracování kapalných havarijních

RAO. ÚJV Řež, a.s. report, ÚJV Řež a.s., Řež (1994).


33

04 – 07 / ID 139. Post-Accident Management – Lessons Learned and Preparedness

APPROACHES OF SELECTION OF ADEQUATE CONDITIONING METHODS FOR

VARIOUS RADIOACTIVE WASTES IN FUKUSHIMA DAIICHI NPS

Y. Meguro1,2

, A. Nakagawa1,2

, J. Kato1,2

, J. Sato1,2

, O. Nakazawa1,2

, T. Ashida1,2

1 Japan Atomic Energy Agency (JAEA), Tokai, Japan

2 International Research Institute for Nuclear Decommissioning (IRID), Tokyo, Japan


Abstract. A variety of radioactive wastes, of which chemical composition and radioactive inventory are

different from those of wastes from conventional nuclear power stations in Japan, have been generated in

decommissioning of Fukushima Daiichi Nuclear Power Station. It is necessary to evaluate feasibility of

conditioning methods to these wastes, in which waste forms that meet the criteria in Japan can be produced and

which have been used to real radioactive wastes at home and abroad, because the majority of such wastes have

not been solidified in Japan. The authors investigated an approach for screening of conditioning methods for the

Fukushima wastes on the basis of the findings of the existing methods and results of fundamental solidification

tests using synthetic Fukushima wastes, only in terms of technical practicability. Here five solidification

methods such as cementation, vitrification and sintering were selected, and also 13 wastes with different

chemical composition are solidified, and hardening characteristics and compressive strength and leachability of

the solidified form are studied. A screening flow of the methods was proposed, and evaluation criteria on each

step in the flow was set up. The screening has not been finished yet, but in this presentation a trial result was

opened for a waste and improvements of the screening flow found in the trial evaluation was described.

Key Words: Conditioning method, screening of methods, Fukushima wastes.

1. Introduction

In decommissioning of Fukushima Daiichi Nuclear Power Station (1F), a variety of

radioactive wastes are being generated, and these differ in type and show different chemical

composition and radioactive inventory from those of wastes generated by operation and

decommissioning of the conventional commercial nuclear power plants in Japan [1]. For

example, more than 30 types of radioactive wastes are generated from contaminated water

treatment facilities that were introduced to decontaminate contaminated water accumulated in

nuclear reactor buildings and their turbine buildings of 1F.

Conditioning methods of the wastes have to be developed toward disposal. Here a

conditioning method consists of a pre-treatment method and a solidification method with or

without an immobilization method. Because most of wastes from the water treatment

facilities in 1F have not been generated until now in Japan, anyone has never solidified these

wastes. Therefore at first it is necessary to confirm feasibility of conditioning methods, which

have practical results to conventional radioactive wastes at home and abroad. Of course it is

necessary for the characteristic of the waste form to be met to criteria of Japan. In addition, it

is not realistic that separate solidification method is used for individual waste each, because

there are a variety of the wastes and this idea leads to construction of many solidification

facilities. Therefore it is important to select suitable conditioning methods of high general

versatility and to apply them to the wastes as many as possible.

The authors have studied existing waste conditioning methods, evaluated characteristics of 1F

wastes, prepared their synthetic wastes, and solidified them in fundamental scale, in order to

screen conditioning methods. Unfortunately, enough data from fundamental solidification



34

tests for screening has not obtained yet. Therefore, in the present paper, we describe a

screening way that was newly proposed to narrow down the several applicable conditioning

methods from the viewpoint of technical feasibility using the findings of the existing methods

and the results of fundamental solidification tests.

2. Conditioning Method Screening Approach

At first proven conditioning methods used at home and abroad are investigated and also the

findings of the existing methods are summarized. On the basis of the findings, several

methods are chosen for fundamental tests to confirm technical feasibility. Typical wastes are

picked from the wastes of the contaminated water treatment facilities in 1F, and their

synthetic materials are prepared and used to the fundamental conditioning tests, in which

characteristics of a waste form are obtained. A screening flow consists of several steps, and

criteria in each step are devised to evaluate feasibility of the methods, in which the findings

and the test results are used. This screening approach is not intended to choose the most

suitable conditioning method for each waste and is intended to judge that each conditioning

method is feasible to each waste. As the next, the conditioning method that is applicable to

waste as much as possible is chosen. The choice of the final method will be decided after

future engineering tests.

3. Fundamental Solidification Test

Based on the findings of the investigation of several solidification methods, five methods;

cementation, geopolymerization, vitrification, melt-solidification, sintering solidification

were chosen for fundamental solidification tests. The wastes more than 40 kinds have been

generated from operation of about 10 contaminated water treatment devices. These are

sludges, inorganic and organic adsorbents, filters and so on. It is hard to investigate even

fundamental solidification for all these waste, because it is limited about time. In generally

the main chemical composition of the waste is more important than radioactive inventory for

the solidification. And so 13 wastes having different chemical composition were chosen from

the water treatment wastes. In the choice, combustible wastes that did not leave any ash after

incineration were not chosen under the assumption that all combustible wastes might be

incinerated before solidification. The chosen wastes were a zeolite, a silicotitanate, a

ferrocyanide adsorbents for removal of 137

Cs, a titanate for 90

Sr, a titanium oxide and a

cerium oxide for 125

Sb, a barium sulfate, an iron hydroxide and a carbonate precipitates for

sedimentation decontamination, and incineration ashes of a Cs filter, a Sr filter, a chelate

resin and an ion-exchange resin. In the solidification tests, synthetic wastes were prepared on

the basis of characterization studies for the real wastes and obtaining the adsorbents that were

used in 1F by any means possible.

4. Screening Flow

A pilot screening flow was proposed to evaluate feasibility of the conditioning methods, these

are pre-treatment, solidification and immobilization methods, to the 1F wastes and criteria in

the steps in the flow were determined. The flow was prepared for only technical evaluation

and not for engineering judge. The flow is shown in Fig. 1. The flow consists of two

treatment points and six evaluation steps. The first point is pre-treatment at A and the second

is conditioning at B. When the wastes need pre-treatment prior to solidification, pre-treatment

methods are evaluated at step A-1, and then the treated waste is characterized at step A-2. In

the case of no pre-treatment or after the pre-treatment, the need of solidification is evaluated

in the point B. When the waste should be solidified, at first the solidification methods are


35

evaluated at B-1, in which the methods divides two groups depending on their property. The

methods involved in the step B-1-1 are not able to immobilize the waste into a container for

disposal, and the other methods in the step B-1-2 can directly immobilize the waste in a

container as a waste form. After the step B-1-1, the solidified waste is evaluated at the step B-

1-1-1. At last, the characteristics of the waste form is evaluated at the step B-3 following the

feasibility study of immobilizing method at the step B-2 in some cases.

In the wastes from the contaminated water treatment devices, there is a possibility not to be

directly solidified these wastes. And then the pre-treatment of them is investigated at the

point A. Judge items are followings. The waste includes much water, hazardous substances

such as ferrocyanide compound, materials influencing the waste form’s stability and organic

substances, the waste amount can be reduced, and it is difficult to be directly solidified. The

evaluation criteria at the step A-1 are that whether the above issues can be solved or not. The

pre-treatment methods to be evaluated here are general dehydration, drying, incineration or

heating, gasification, compaction, melting, cutting, and crushing. The items obtained at the

step A-2 are the treated waste’s conditions such as weight, particle size, specific gravity, their

chemical composition, and their radioactivity inventory.

FIG. 1 Evaluation flow of conditioning method.

Criteria for judging at the point B are followings. It is necessary from current Japanese

regulations, and there is possibility to leak radioactive substances, and lack of stability and

safety of the waste. For the solidification method included in the step B-1-1, evaluation items

are to solidify the waste, to deal with the solidified one easily, to reduce the volume, not to

include any substances influencing disposal. At the step B-1-1-1, chemical composition,

strength, H2-gas generation, chemical stability, leaching ability, and crushability are

evaluated. The solidified waste through B-1-1-1 and the waste that has not to be solidified at

the point B are immobilized in a container as a waste package and then feasibility of

immobilization methods are evaluated at the step B-2. The evaluation items in the step B-2

are to make a stable waste package, to increase safety, and to reduce risk of leakage of

radionuclides. At the step B-1-2, evaluation items are followings. The waste is easily mixed

with solidified materials, the filling rate of the waste is high enough, the mixture is harden

within one day, the solidified waste shows high stability for long time, any substances that


36

influence disposal are not included, the safety during storage and transfer increases, and risk

of leakage decreases. At last the characteristics, such as compressive strength, temperature

and H2 generation by radiolysis, leaching ability of the main radionuclide, of the waste

package are evaluated at the step B-3.

5. Screening Test

According to the evaluation flow as shown in Fig. 1, feasibility of the pre-treatment and

solidification methods were evaluated for the sludge generated from the decontamination

device in 1F using fundamental experimental results. At the point A, it was judged that the

sludge should be pre-treated because the sludge included water, a ferrocyanide compound

and sulfate. At the step A-1, dehydration without heating, drying at around 100 ºC or more,

and heating with high temperature, were chosen as pre-treatment methods. Because the

ferrocyanide compound in the sludge was decomposed over 350 ºC or more in the

atmosphere [2], it was decided that the heating method over 350 ºC was suitable. The

radioactive inventory and main chemical composition of the treated sludge were estimated

from characterization studies of the 1F wastes [1], but the other properties such as its grain

size and specific gravity were not. To obtain them the engineering test using the real sludge

waste is necessary.

Next at the point B, it was decided that the treated sludge have to be solidified mainly from

Japanese regulation. Although experimental results of fundamental solidification teste were

not described here, the treated sludge was solidified by vitrification and geopolymerization

methods that were nominated in the step B-1-2. The tests by other solidification methods

have been conducted. Both the methods showed enough high solidification ability to meet the

criteria of the step B-1-2. After this screening test, missing information for the evaluation was

summarized and these will be used feature experimental plan.

6. Conclusion

The evaluation flow was proposed to confirm the feasibility of the conditioning methods to

the wastes generated in 1F, and a screening test was conducted for the sludge. Although it has

not finished the screening of the solidification method to the 1F water treatment wastes,

several issues and missing information were extracted from the trial test. For example when

the feasibility was determined at each step, only one item was used for judgement in some

case, and it was concluded that priority should be considered in the evaluation. The priority

of the issues may be different for each waste, and therefore the priority has to be decided

individually.

7. Acknowledgment

This paper includes the results of “Development of technology for treatment and disposal of

accident waste” subsidized to IRID/JAEA by METI.

REFERENCES

[1] KOMA, Y., et al., “Research and Development on Waste Management for the

FUKUSHIMA DAIICHI NPS by JAEA,” Int. Nuclear Fuel Cycle Conf. GLOBAL 2013

(Proc. Symp. 2013), (2013) 736.

[2] MIMURA, H., et al., “Physicochemical Properties of Potassium Nickel

Hexacyanoferrate(II)-Loaded Chabazites”, J. Nucl. Sci. Technol., 35, (1998) 5.

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