IRID Annual Research Report

2014

Printed on recycled paper.

Published: March 2015

Contents
Greetings

Main R&D Achievements: Subsidized Projects■ Key Challenge 1: R&D for Fuel Removal from Spent Fuel PoolInternational Research Institute for Nuclear Decommissioning (IRID) conducted during FY2014. I would like to

Evaluation of Long-term Structural Integrity of Fuel Assemblies Removed from the Spent Fuel Pool ............................................ 2

take this opportunity to sincerely thank all those who extend their understanding and support to IRID’s activities

Study of Methods to Process Damaged Fuel Removed from the Spent Fuel Pool........................................................................... 4

Since its establishment, IRID has been addressing the pressing R&D challenges presented by the■ Key Challenge 2: R&D for Preparation of Fuel Debris Retrieval

decommissioning of the Tokyo Electric Power Company’s (TEPCO) Fukushima Daiichi Nuclear Power Station(NPS) based on the ‘Mid-and-Long-Term Roadmap towards the Decommissioning of TEPCO's Fukushima
Daiichi NPS Units 1-4’ (‘Mid-and-Long-Term Roadmap’). IRID completed 17 subsidized projects and one
in-house R&D project during FY 2014 as part of this effort.

projects in an integrated and comprehensive manner, and collaborating with worldwide organizations:(1) Fuel Removal from Spent Fuel Pool

Development of Technology for Remotely Operated Decontamination Inside Reactor Buildings .................................................. 6
Development of Repair and Water Leakage Stoppage Technology for Leakage Points Inside the Primary Containment Vessel ..... 8
Full-scale Test for Repair and Water Leakage Stoppage Technology for Leakage Points Inside the Primary Containment Vessel... 10
Development of Technology for Investigation Inside the Primary Containment Vessel

Main Study
In-house R&D Activities
Results

IRID has been involved with the three following R&D project categories, promoting interlinkage between

Subsidized Projects

on a daily basis.

Main R&D
Achievements

This report presents a collection of the results of the Research & Development (R&D) projects which the.......................................................... 12

Development of Technology for Investigation Inside the Reactor Pressure Vessel ...................................................................... 14(2) Preparation of Fuel Debris Retrieval(3) Treatment and Disposal of Radioactive Waste
As an example, IRID developed a remotely operated robot which enabled us to verify the effectiveness oftechniques for conducting decontamination work and surveying contaminated water leakage points inside reactor
buildings at the Fukushima Daiichi NPS. Furthermore, in order to determine the location and state of the debris
look inside the primary containment vessel.
In order to efficiently and steadily undertake the unprecedented and extremely difficult – even by global
standards – of decommissioning the damaged Fukushima Daiichi Reactors, IRID is ready to play its part in thearea of R&D, working in cooperation with the government, the Nuclear Damage Compensation and

IRID will continue to steadily push forward with R&D efforts without losing a sense of urgency and with apositive attitude, we will strive to make visible progress in decommissioning efforts, enabling those people in
Fukushima forced to live as evacuees to return home at the earliest opportunity, and ensure peace of mind for theDevelopment of Technology for Non-destructive Detection of Radioactive Materials Accumulated in the Suppression Chamber... 20
Development of Technology for Retrieval of Fuel Debris and Reactor Internals .......................................................................... 22
Development of Technology for Collection, Transfer and Storage of Fuel Debris ......................................................................... 24
Development of Technology for Evaluating the Integrity of the Reactor Pressure Vessel/Primary Containment Vessel ............... 26
Development of Technology for Criticality Control in Fuel Debris Retrieval................................................................................. 28

Data Sheet

Decommissioning Facilitation Corporation, and TEPCO.

Development of Technology for Detection of Fuel Debris in the Reactor..................................................................................... 18

Othersinside the reactors, IRID has been developing devices to use cosmic ray muon detection technology and robots to

Identifying Conditions Inside the Reactor through Application of Severe Accident Analysis Code .............................................. 16

Development of Technology for Fuel Debris Characterization and Treatment.............................................................................. 30
Development of Technology for Analysis of Debris Properties .................................................................................................... 32

society as a whole.

■ Key Issue 3: R&D for Treatment and Disposal of Radioactive Waste

I strongly hope that this report helps a wider audience gain a better understanding of IRID’s R&D achievements,

Development of Technology for Treatment and Disposal of Accident-generated Waste .............................................................. 34

and ask you for your continuous guidance and support in our activities.
March 2015

Main Study Results: In-house R&D Activities
Analysis of Contaminated Objects Sampled from Reactor Buildings........................................................................................... 36

Others ‒ Data Sheet

Hirofumi Kenda

President, International Research Institute for Nuclear Decommissioning

Main Research Results Announced/Published in FY2014 ........................................................................................................... 37
Main R&D Installations/Equipment............................................................................................................................................. 39
Joint Research/Contract Research ............................................................................................................................................... 40

IRID Annual Research Report 2014

1

Main R&D Achievements
Key Challenge 1: R&D for Fuel Removal from Spent Fuel Pool

Evaluation of Long-term Structural Integrity of Fuel Assemblies
Removed from the Spent Fuel Pool
Aims

In order to determine if fuel assemblies retrieved from thespent fuel pools in Units 1-4 at the Fukushima NPS can be
stored safely or not, it will be necessary to perform corrosion
tests and investigate actual fuel in an environment thatsimulates real storage conditions. Through these tests, thestructural integrity of fuel assembly components duringlong-term storage stored in a common pool can be assessed.

Subsidized Projects

Upper end plug

Main Achievements

1. Technical development for the evaluation of the long-term integrity of the fuel assemblies
a. Development of technology to assess long-term integrity

As a large load is placed upon the upper and lower sections of fuel assemblies when they are handled, the integrity of theseparts is extremely important. To make sure no problems would arise due to the spent fuel pool water quality or rubble,
immersion tests (up to for 8,000 hours) and strength tests using a test specimen were conducted, and the integrity of the parts
was evaluated. Results showed that there was no corrosion that may affect fuel assembly integrity or degradation due tomechanical stress in the threaded sections and cladding subject to evaluation (Figure 1).

Figure 1 Results of cross-sectional (left) and external (right) observation of fuelassembly upper section test specimen (The test specimen was immersed for
1,000 hours at a temperature 90℃ and a chloride ion concentration of
2,500ppm. After that, it was further immersed for 7,409 hours at a temperature
60℃ and chloride ion concentration of 100ppm.)

This test simulated an environment in which concrete rubble was assumed to bepresented. No corrosion was observed in upper-end plugs (zircaloy) and nuts
(stainless steel).

b. Survey of condition of spent fuel stored in the common pool

In order to conﬁrm corrosion conditions under long term storage in the common pool, regarding fuel assemblies retrieved fromthe spent fuel pool in Fukushima Daiichi NPS Unit 4, surface observations of the fuel assembly, observations of the innerthreaded portion surface in crevice areas and measurement of the thickness of the cladding oxide layer were performed.
Results of surface observations showed that while a white deposit had formed on spent fuel, there were no issues with external
appearance (Figure 2). Furthermore, corrosion was found to have not occurred on the inner surface of the threaded section, andmeasurements of the thickness of the cladding oxide layer showed that when compared with oxide layer thickness in existing
fuel stored in the common pool, no increase in thickness had taken place.

Upper end plug
Nutc. Evaluation of integrity of fuel in dry storage

Tests were conducted to assess hydride precipitation behavior inside cladding material when scratches due to rubble exists andthe effect on fuel integrity of the moisture contained in rubble which has penetrated crevice areas under the assumption thatspent fuel will be stored in a dry environment. Test results showed that these had only a minimal impact on fuel assemblies indry storage. Investigations will continue on the aging of materials in the assessment of the integrity of fuel in dry storage.

Radioactivity distribution
measurement sample

IP measurement for Cl-36 distribution

2. Basic tests related to long-term structural integrity

a. Evaluation of seawater component transfer behavior on fuel components

In order to assess the seawater component transfer behavior on fuel components, tests were conducted using simulated scaling(crud) and a tracer. Results showed that the amount of seawater component transferring to the crud and material surface wasminimal. In evaluating whether there is an uptake in seawater components to the oxide ﬁlm that is formed on the fuel assembly
cladding surface or not, conﬁrmation tests that take the heat generated inside the fuel assemblies into account were conducted,
and it was found that there was no signiﬁcant uptake that occurred.

Figure 3 Distribution of Cl-36 in upper end plug screw structure (80℃, two-folddilution of artiﬁcial seawater, after 50 hours immersion)
By immersing the specimen in artiﬁcial seawater with a two-fold dilution of chloride,
it was conﬁrmed that seawater components transfer to the screw structure, butconcentration decreases afterwards as water quality improves.

Figure 2 Survey of condition of fuels stored in common pool

External observation (example) of spent fuel from Fukushima Daiichi NPS Unit
4 (upper tie plate of fuel mixed with rubble (burn-up=49.3GWd/t))

A white deposit was noted on the upper tie plate and lock nut surface, but there were
no issues in appearance.

Re-passivation potential for crevice corrosion

Fuel assemblies in spent fuel pools at Units 1-4 at the
Fukushima Nuclear Power Station (NPS) have been stored in
an environment with water quality different from normal
conditions due to the injection of seawater and addition of
rubble. To achieve long-term storage of fuel assemblies in acommon pool in future, it will be necessary to accurately
evaluate the effects of these conditions on fuel assemblycomponent materials and the integrity of the fuel assemblies
for long-term storage.

Main R&D
Achievements

Background

Nut

0.8
0.6

Room temperature Dose rate

5kGy/h
0.5kGy/h
0Gy/h

0.4
0.2
0

Steady-state
corrosion potential

-0.2
-0.4
100

101

102

103

Chloride ion concentration (ppm)104

Figure 4 Relationship diagram between re-passivation potential for crevice corrosion

in diluted artiﬁcial seawater and chloride ion concentration obtained throughelectrochemical test using crevice corrosion test specimen (testing using
crevice corrosion specimen made of a combination of Zr and SUS304L).

In room temperature water with a chloride ion concentration of less than 100ppm
and exposure to gamma radiation at 0-5kGy/h, it was conﬁrmed that crevice
corrosion would not occur as re-passivation potential for crevice corrosion was
higher than steady-state corrosion potential.

b. Evaluation of effect of corrosion derived from seawater or rubble components under radiation environment

The results of an evaluation of fuel material taken from new fuel assemblies that were in the Fukushima Daiichi NPS Unit 4 spent
fuel pool showed no signs of corrosion or loss of integrity. Additionally, results of electrochemical tests conducted underradioactive conditions conﬁrmed that in room temperature water that contains a chloride ion content of less than 100ppm,
repassivation potential for crevice corrosion was greater than steady-state corrosion potential, and that crevice corrosion did notoccur. (Figure 4) It was also conﬁrmed that even after 1,500 hours of corrosion testing, no signs of corrosion were observed.

2

Future Developments

In order to ensure that fuel assemblies retrieved from spent fuel storage pools in Fukushima Daiichi NPS Units 1-4 can be safely
stored, material taken from spent fuel in Unit 4 will be transported to a “post-irradiation test facility” and studies of white deposits
observed in surface observations and so on will be performed. Additionally, tests are planned to be performed on applicable fuelassemblies that take combined environmental factors into consideration in order to evaluate the integrity of fuel assemblycomponents during dry storage.

IRID Annual Research Report 2014

3

Aims

as seawater and concrete, spent fuel in Units 1-4 may have also

the reprocessing of damaged fuel, indicators must be

In addition to having been contaminated with impurities such

how this damaged fuel is to be dealt with, it is necessary to
conﬁrm whether reprocessing the fuel is technically feasible or
not as one of the treatment methods.

In addition to identifying the technical challenges regarding
developed to determine the feasibility of reprocessing thefuel. The aim of this project is to assess the effect on thechemical treatment process and so on, and extract andorganize the expected impact at domestic reprocessing
plants (RRP/TRP*).

* RRP: Rokkasho Reprocessing Plant; TRP: Tokai Reprocessing Plant

Main Achievements

RRP simulated waste solution(chloride ion0g/L)(1) Surface observations of immersion test piece

RRP simulated waste solution + artiﬁcial seawater(chloride ion 20g/L)

Chloride ion concentration (g/L)*Relative corrosion rate is described as corrosion rate ratio withthat of RRP simulated solution containing no impurities valued as “1”.

Subsidized Projects

been damaged due to falling debris. In order to determine

Main R&D
Achievements

Background

Relative corrosion rate

Study of Methods to Process Damaged Fuel Removed
from the Spent Fuel Pool

(2) Relative corrosion rate and chloride ion concentration(artiﬁcial seawater)

Figure 1 High-level liquid waste storage tank immersion test results (Immersion time: 960h)

In immersion tests of high-level liquid waste storage tank material (SUS316L) using a simulated solution to which impurities(artiﬁcial seawater, etc.) have been added in a simulated waste solution containing a simulated ﬁssion product component,
etc., pitting due to chloride ions within impurities was not observed, while intergranular corrosion growth was shown to haveoccurred under conditions without the addition of impurities, as surface observations indicated. A decrease in the corrosion
rate was also observed due to an increase in the chloride ion concentration, and it was determined that there was noacceleration in equipment corrosion due to the presence of impurities.

Corrosion tests (immersion, electro chemistry) was conducted on materials in high-level concentrated waste liquid storage tanks
ﬁssion products, etc. Under all conditions, results showed that while there was uniform corrosion in the form of intergranularcorrosion, pitting was not observed (Figure 1 (1)). Furthermore, an increase in chloride ion concentration was accompanied by adecrease in corrosion (Figure 1 (2)).

2. Evaluation of in-process behavior of impurities

Organic phase(solvent containing U/Pu)

Distribution ratio

Bibliographical values (1) (2)and high-level liquid waste storage tanks, using simulated high-level liquid waste solutions that took into account impurities and◆ Distribution ration of impurity
components (◇Below lower
limit of quantiﬁcation)

In order to understand the effect of the transition of impurities to uranium/plutonium (U/Pu) products and the impact of impurities
on the extraction of U/Pu, solvent extraction testing took place using a simulated solution with added impurities and ﬁssion
products in addition to a simulated solution to which U/Pu had been added. The results of these tests conﬁrmed that impurities

had a low distribution ratio of 10-2 -10-3 and they were hard to be extracted into the solvent, and that the distribution ratio of U/Pu

was not dependent on the presence of impurities (Figure 2).

3. Evaluation of the inﬂuence of impurities on waste body

To determine the composition of waste solutions assumed to be handled at domestic reprocessing facilities (pressurized water
reactor with 45,000 MWd/tU burn-up)*, a glass specimen was created from a formulation of powder materials with the major
components of seawater and mortar added as impurities, and an evaluation of homogeneity and of glass property values was

Aqueous phase (nitric acidcontaining impurities, etc.)(1) Samples left to stand after
extraction

Distribution ratio (D)1. Evaluation of corrosive inﬂuence of impure substances on re-processing equipment

No added
impurities

Simulated ﬁssion product Simulated ﬁssion product+ impurities+ impurities+ silica colloid

Additive impurities[1]Data of Inorganic Solvent Extraction (1) JAERI1047[2]PUREX Technical Manual (1980)

D = Organic phase (total aqueous phase for 5X back extraction)/Aqueous
phase (total for 2X extraction)(2) Distribution ratio of impurity components during coexistencestate with simulated ﬁssion products(3) Distribution ratio of U/Pu during coexistence state with impurity
components

Figure 2 Outcome of evaluation of the in-process behavior of impurities

In solvent extraction testing using a simulated solution obtained by adding impurity components to simulated ﬁssion
product components, the distribution ratio of impurities was of the order of a low 10-2 to 10-3, making them difﬁcult to extractfrom the solvent (30% TBP-n-dodecane) as the graph above shows. Furthermore, in solvent extraction testing using impurity
components in a U/Pu solution and simulated solutions to which simulated ﬁssion product components had been added, theU/Pu distribution ratio conﬁrms that there was no signiﬁcant change in the presence/absence of impurities. Based on theseﬁndings, it was determined that there was no effect on the transition of impure substances to the U/Pu products or on
extraction of U/Pu due to impurities.

then performed. The result was that, under all conditions, there was no phase separation precipitation, and vitriﬁcation occurred,
and there was no effect of impurities on the physical property data of the glass (Figure 3).

* This composition was used in order to make comparisons with previous tests. It was previously conﬁrmed that there is no difference between the
results of this test and in the case found at the Fukushima Daiichi NPS.

Inﬂuences expected at reprocessing facilities at the time of processing damaged fuel were comprehensively identiﬁed and thepresence or absence of required elements of research is clariﬁed. The result was that there were no new research elements outsideof the existing evaluation items; evaluating the handling of used fuel, evaluating the corrosive impact on reprocessing equipment
due to impurities, evaluating in-process behavior, or waste impact assessment.

Future Developments

Through evaluation tests to date, it has been possible to obtain key data (if impurities will impact on the chemical treatment
process, etc.) necessary to determine if reprocessing is feasible. The necessity of investigating issues related to the handling of

damaged fuel and so on will be determined in light of the results of conﬁrmation of the state of fuel in the spent fuel pool of Unit
3 that is planned for future retrieval.

Standard concentration

X-ray intensity

4. Identifying and outlining of other inﬂuences

10X standard concentration

100X standard concentration
Standard concentration*

10X standard concentration 100X standard concentration(1) Visual observation*Standard concentration: Estimated mix concentration to vitreous waste throughconservative estimate of amount of adhered/entrained impurities (seawater/mortar
components).

(2) X-ray diffraction (XRD) measurement results.

Figure 3 Evaluation of homogeneity results using glass specimen containing seawater/mortar components.

In an evaluation of homogeneity of glass specimens created adding major components of seawater and mortar to act as animpurity in a standard waste liquid composition, no precipitation of phase separation objects was visually observed and aswith standard glass in XRD measurements, no other peaks aside from RuO2 derived from standard waste liquid wereobserved.

4

IRID Annual Research Report 2014

5

Main R&D Achievements
Key Challenge 2: R&D for Preparation of Fuel Debris Retrieval

Development of Technology for Remotely Operated Decontamination
Inside Reactor Buildings

Improvements/veriﬁcation tests of equipment for decontaminationof high places and production/ veriﬁcation tests of equipment fordecontamination of upper ﬂoors based on R&D achievements
obtained up to FY2013 is to be conducted. Through these steps,
remotely controlled decontamination techniques required for thesmooth implementation of surveys and repairs planned to takeplace inside reactor buildings will be established.

Basement area
decontamination
地下階除染装置
equipment(T/B R/B)

Lower parts
decontaminationequipment(Development
completed)

Figure 1 Locations where each type of decontamination equipment is to be placed.

Remotely operated decontamination equipment is used in (1) lower parts of the ﬁrst ﬂoor, (2) upper parts of the ﬁrst ﬂoor, (3) upper ﬂoors, and(4) basement areas (Figure 1). Prior to FY2013, (1) completing development for lower parts of the ﬁrst ﬂoor, (2) production/element testing of
decontamination equipment for upper parts of the ﬁrst ﬂoor and (3) design of decontamination equipment for upper ﬂoors were conducted.
Based on the results of these actions, the following tasks were undertaken during FY2014:Arm
movement

Head

1. Development of decontamination equipment for high places

Elevatingworkmachinea. Improvement of decontamination equipment

Improvements were made to the three types of decontamination equipment for high places produced during FY2013 (high pressure water
jet, dry ice blast, and suction/blast) in preparation for their actual application. These improvements included the production of a transfer unit
for the high pressure water jet equipment and improvements to the operability of dry ice blast during traveling or decontamination. In
addition, there was a reduction in the size of the hose reel section in the suction/blast decontamination equipment (Figure 2).

Hose reel

Cart (in traveling)b. Veriﬁcation testing of decontamination equipment

Test results conﬁrmed the effective range of decontamination of high places using the high pressure water jet and dry ice blast equipment.
The workability of the suction/blast decontamination equipment on the primary construction target of wall surfaces after obstacles have been
removed was also conﬁrmed. Through these tests, the goal of conﬁrming applicability to the actual plant was obtained.

Arm/HeadArm(set oftwo)

Main Achievements

A mockup test facility was produced (Figure 3) and veriﬁcation tests took place.
Equipment to be evaluated during actual operations was identiﬁed, and mockup equipment was produced taking the diversity and priorityof applicability into consideration. Decontamination performance, traveling performance/operability when remotely controlled and safetyfunctions were evaluated during veriﬁcation tests to conﬁrm required performance.

Subsidized Projects

In the lead-up to retrieving fuel debris from Units 1-3 atFukushima Daiichi, various tasks are planned to take place inside
the reactor buildings. In order to perform these tasks smoothly,
improving the working environment is essential. An overall
reduction in radiation levels is sought through a combination of
decontamination, shielding, and removal of radioactive sources.

Aims

Upper parts
decontaminationequipment
Main R&D
Achievements

Background

Upper ﬂoor
decontaminationequipment

High pressure water jet
decontamination equipment

Cart (in traveling)

Dry ice blast decontaminationequipment

Suction/blast
decontamination equipment

Figure 2 Improvements to equipment for decontamination of upper parts conducted during FY2014.

North
北

North

2. Development of decontamination equipment for upper ﬂoors

Plans are in place to introduce an elevating work cart from the open equipment hatch to access the upper ﬂoors of the reactor buildings (2nd and3rd ﬂoors) where decontamination of the ﬂoor and wall surface (approximately 2 m in height) will be performed. At the present time, work is
progressing on developing ways to utilize the same carts for decontamination technology developed for the decontamination of low areas (highpressure water jet, dry ice blast, and suction/blast).
Speciﬁcally, production of equipment based on methods of accessing upper ﬂoors and designs considered during FY2013 commenced.
Equipment is comprised of a shared cart (work cart/transport cart/support cart/relay cart) and each decontamination unit (Figure 4). Productionof the equipment will be completed during FY2015, and will then go through a veriﬁcation testing process.

3. Conceptual study on decontamination equipment for basement areas

In future, a drop in groundwater levels around the reactor buildings due to the operation of the bypass system and the ground freezing method
is expected, and the stagnant water level inside the reactor buildings must therefore also be lowered in response. However, this may result in arise in dust and in radiation levels inside the buildings. A conceptual study was therefore conducted on dust diffusion reduction measures forbasement areas and decontamination work when required in basements.
Future Developments
In addition to identifying challenges through veriﬁcation tests and implementing steps to deal with these, equipment for the decontamination ofhigh places will be utilized at the Fukushima Daiichi NPS according to the required purpose and timing. Equipment for the decontamination of
upper ﬂoors will be completed during FY2015 and veriﬁcation testing on mock-ups conducted. Additionally, consideration of the necessity ofdeveloping new technologies for the decontamination of basement areas will continue.

6

Figure 3 Mockup facilities simulating actual plant.

「機器ハッチ開口部からのアクセス」を想定し、上部階運用を考慮した各除染技術（吸引・ドライアイ
ス・高圧水・ブラスト）の開発を行うとともに、各除染技術の取扱いが可能な共通システムの開発（設計）を実施した。
Compressor
Decontamination unit

＜Relay cart＞＜Support cart＞＜Transport cart＞＜Work cart＞：Shared cart

上部階へのアクセスシナリオに基づき、各除染装置及び共用台車の設計を実施した。

Figure 4 Conceptual diagram of upper ﬂoor decontamination equipment (shows mounted blast equipment as an example).

IRID Annual Research Report 20147

Development of Repair and Water Leakage Stoppage Technology
for Leakage Points Inside the Primary Containment Vessel

SRV pipe
Inner catwalk

Vent pipe

Aims

not only fallen down within the reactor pressure vessel (RPV),of water leakage points in the PCV that will allow for the

It is assumed that fuel debris at the Fukushima Daiichi NPS has(PCV). A plan exists to submerge the PCV in order to retrieve(4) Chloride ion removalby reverse osmosis
membrane

Outer catwalk(5) Injection of water into the reactor(3) Nuclide removalusing absorption
vessels

This plan aims to establish technology that enables the repair
retrieval of fuel debris in a submerged state.

Vent pipe bellows

r
mete
r dia
inne
shell
S/C

the fuel debris; however, it is necessary to ensure that there is
no water leakage from the PCV prior to this step.

Main Achievements

1. Technology for strengthening the suppression chamber (S/C) support columns

In conjunction with the "Development of Technologies for Assessment of the RPV/PCV Integrity" R&D project, the reinforcement

material placing height (height to which the material needs to be ﬁlled to support the S/C in the torus room) was set to the uppersupport column pins and target compression strength was established (Figure 1). Looking ahead, casting tests and developments(2) Cooling of contaminatedwater by the heatexchanger

Vent header OP4660

Upper pin(6) Water level control
S/C center OP3570

Water stoppage boundary(buffer range)(Height to which the reinforcement
material needs to be ﬁlled tosupport the S/C)
Upper pinT-quencher OP970
Ring header(1) Water intake from the S/C(refer to following paper forwater intake locations)

Column support
Column support

Subsidized Projects

but has also made its way into the primary containment vessel

Vacuum break line

Main R&D
Achievements

Background

Vacuum break line bellows

S/C lower end (bottom)OP-485

Figure 2 Concept of the PCV circulation cooling system

The reinforcement
material placing height in the
S/C (green area)
S／C脚部補強埋設範囲（緑）Figure1 The studied reinforcement material placing height in the suppression

chamber (S/C)

made in order to execute construction of the reinforcements will continue.

2. Study on the circulation cooling system

A conceptual study of PCV circulation cooling equipment and water intake was conducted. Based on this study, a conceptual
diagram outlining construction steps will be created (Figure 2) in conjunction with an examination of the repair of each leakage
point (stopping water leakage).

3. Water stoppage technology involving injection of sealing material into the vent pipe

Tests conducted based on an auxiliary sealing material deployment improvement plan showed that there is a likelihood of further
improvements to the deployment of the material. A one-half scale model was used for water stoppage testing, and successful

R/B 1st ﬂoor

patterns of achieving a reduction in leakage are being veriﬁed (Figure 3). Looking ahead, water stoppage tests and developments

to execute construction of the reinforcement will continue. In addition, element testing for ﬁlling the S/C to prevent water leaks(stopping water from quencher/ strainer/downcomer) is being performed sequentially. Following this, tests using a full-scalemodel were conducted in preparation for the execution of actual construction.

Vacuum breakline bellows

4. Water stoppage technology involving injection of sealing material into the vacuum brake line

Expansion joint

Elbow

A study and sequential testing of methods to improve the stoppage of water leakage using cloth packing with mortar ﬁlling andsilicon-based material is underway. A ﬂexible type of guide pipe is currently being considered (Figure 4).

5. Water stoppage technology for leakage from the seal section and vent pipe bellow

Vacuum break line
Figure 3 Water stoppage testing using a one-half model of the bent pipe

Figure 4 Flexible guide pipe overview

Based on the ﬁndings and achievements of the preceding ﬁscal year, test plans for stopping water leakage from the equipment

hatch and penetrating upper part of the PCV, and sealing the personnel access lock room, are being studied. Water stoppage tests
are underway, and developments to execute these plans are progressing.

6. Boundary construction technology for PCV connector pipes

water stoppage methods are being studied. The feasibility of these methods will be studied sequentially.

In addition, water stoppage testing and developments based on study ﬁndings are proceeding towards actual construction.

9. Establishment of a plan in lead-up to ﬁlling the PCV with water

For this study, AC, RW, and RCW systems were selected. Consideration of water stoppage proposals for each system is ongoing.

7. Water stoppage technology involving injection of sealing material into clearance between
the torus room wall and penetrating pipes.
Applicability was conﬁrmed based on previous test results. Taking into account the actual environment and the current progressof each project, plans are underway to arrange relevant penetration parts taking conditions at the time of construction into

consideration, construction feasibility, selection of water stoppage materials, and the necessity of element testing in order toprevent water leakage between rooms. Water stoppage testing will continue towards actual construction.

8. Drywell (D/W) shell repair technology

Assumptions about the damage caused by such a severe accident are being made, and water stoppage material selection and
8Considering system conﬁguration at time of submersion, the processes required for planning in the lead-up to ﬁlling the PCV withwater are being examined. Scenarios at each Unit are being collated, and submergence plans clariﬁed.

Future Developments

In order to establish water stoppage technology for each repair/water leak location, ongoing tests will be performed andchallenges will be reviewed, analyzed and evaluated. A strategy to resolve these challenges that emerge in testing will be

developed, and steady R&D efforts will continue to ensure there are no issues left unresolved. Additionally, PCV submersion
scenarios will be studied for each unit separately and the establishment of water stoppage methods to be executed at the actual
plant sought.

IRID Annual Research Report 20149

Full-scale Test for Repair and Water Leakage Stoppage Technology
for Leakage Points Inside the Primary Containment Vessel

Water stoppage/
Reinforcement material line

Aims

Development of repair and water leakage stoppage technology for leakage points
inside the PCV in preparation for the implementation of the submersion method isprogressing. This project involves full-scale testing for the veriﬁcation of thedeveloped technologies (construction method, remotely controlled equipment,etc.) and the operational training applicable on site. Development is planned totake place at the Naraha Remote Technology Development Center through a joint
proposal with the Japan Atomic Energy Agency.

Main Achievements

Concrete supply facility
Waste water
purifying system

Heatingequipment(circulation)

Hot water
supply facility

Figure 1 Conceptual design of the full-scale test facility

1. Full-scale mock-up test of equipment to repair/stop leakage in lower part of the PCV

In order to conﬁrm the operability of placement equipment for suppression chamber support reinforcement developed as part ofthe ‘Suppression chamber support reinforcement technology’ component of the ‘Development of Repair and Water Leakage
Stoppage Technology for Leakage Points Inside Primary Containment Vessels’ (development of water leak repair technology)
project, this plan involves the use of water in full-scale mock-up tests. To determine test facility speciﬁcations this ﬁnancial year, theknowledge derived from the development of water leak repair technology was incorporated. Figure 1 shows a draft conceptual
design of the full-scale test facility.

2. Design/production of full-scale mock-up

In a plan for the design, production, and on-site installation of a full-scale mock-up of the lower part of the PCV at the Fukushima
Daiichi NPS Unit 2, this year saw the commencement of a compilation of speciﬁcations and detailed design take place inpreparation for actual production of the mock-up.
The primary speciﬁcations of the full-scale mock-up are as follows:■The mock is a one-eighth scale model of the suppression chamber including the vent pipes (excluding the bellows), vent head, downcomer
and surface of torus room wall at Fukushima Daiichi Unit 2.

3. Study/design/production/installation of required equipment including plumbing and turbid
water treatment systems
As full-scale testing requires the simulation of the actual environment in the Fukushima Daiichi Units 1-3 to the fullest extent
possible, including dimensions and temperature conditions, the study/design/production/installation of the following equipment
is planned. During this ﬁnancial year, speciﬁcations were collated in preparation for production and detailed designs wereembarked upon (Figure 2).

■Heating/Feed water equipment: In order to provide the heating function to produce heated water that simulates stagnant water in the

Figure 2 Designing various testing equipment

Fukushima Daiichi NPS in a full-scale mock-up test, system designs and a deployment plan for equipment that supplies heated water at therequired volume were formulated, and a boring survey at the proposed construction site required for the construction application was
conducted (Photos 1 and 2).
■Turbid water treatment equipment: As stopping water leakage at the lower part of the PCV is expected to involve the use of grouting material(including cement), water that contains cement is generated after conducting tests. A system design and deployment plan for puriﬁcationequipment outside the plant that processes the turbid water to a quality that allows for drainage was performed.
■Work ﬂoor: As the height of the ﬁrst ﬂoor of the basement in Unit 1 and Units 2 and 3 at Fukushima Daiichi differs, the design and examinationof methods of changing the height of the work ﬂoor at each Unit (ﬁrst ﬂoor of reactor building) took place.
■Mock-up transfer rail: As the weight of the mock-up is very heavy (approximately 5,400t), moving it using a crane after it has been assembled
is difﬁcult. The design/examination of the installation of a mechanism to move the mock-up (mock-up transfer rail) was performed.

Future Developments

The design, production and installation of the test facility will continue in parallel with the project to develop repair and waterleakage stoppage technology in the lead up to implementing full-scale mock-up tests. In addition to evaluating the applicability to
Unit 2 of the status of implementation and results of tests performed after the test facility is completed, the facility will be used fortraining personnel and is aimed to establish repair and water stoppage technologies for PCV leakage points towards
implementation of the submersion method.
10

Photo 1 Boring survey at the site

Photo 2

Soil sample taken during boring survey

IRID Annual Research Report 2014 11

Subsidized Projects

In order to prepare for the removal offuel debris after submerging it in water(submersion method) as part of the steadydecommissioning of the Fukushima Daiichi
NPS, the establishment of technology torepair and stop PCV leaks is required.

Feed line

Main R&D
Achievements

Background

Drain line

Development of Technology for Investigation Inside the PrimaryContainment Vessel
Aims

Video footage, information about radiation dose andtemperature have conﬁrmed that, in addition to being aharsh, highly radioactive and extremely humid environment,
visibility inside the PCV is limited due to the presence ofsteam and stagnant water and to it being very dark. Moreover
there is the possibility of unforeseen obstructions that weregenerated from the accident. In order to overcome thesechallenges, this project aims to develop technology for theinvestigation inside of the PCV.

Shapechange

Photo 3 B1 investigation device (shape-changing robot)
To be used outside Unit 1 pedestal.

Reactor pressure vessel
Grating

X-100BCRD rail

Main Achievements

1. Formulation of a plan for investigation inside the PCV and development of investigationequipment

Needs analysis were conducted for other related projects (Development of Technology for Fuel Debris Retrieval/Reactor Internals,
Development of Repair and Water Leakage Stoppage Technology for Leakage Points inside the PCV, etc.) and the results weresummarized. To respond to these needs, the outline of PCV access methods for internal investigation and investigation methods
were studied. Prospects emerging from the results of these studies and plans were shared with other related projects, andadjustments were made accordingly.

X-6

2. Development of investigation devices, etc.
a. Technology for accessing inside the pedestal

Development of devices that enter the PCV in the Fukushima Daiichi Unit 2 and then penetrate into the pedestal to conduct
investigations (‘Investigation on platform inside the pedestal’ (A2 Investigation)) is currently underway (Photo 1).

：Access route
Inside primary containment vessel (PCV) (image)

Photo 2 Shielding block removal equipment

To be used in front of X-6 penetration in Unit 2.

Pedestal

A self-propelled miniature robot will enter via the opening (diameter about 115 mm) of the X-6 penetration (PCV penetratingpart), and after traveling through a guide pipe inserted into the PCV, traverse the CRD rail into the pedestal interior. Upon arrivalin the pedestal, it is planned that onboard measurement devices (camera, etc.) will gather various types of data. Veriﬁcation tests
are expected to commence in the ﬁrst half of FY2015.

b. Technology for removal of the shielding block
Prior to installation of the development unit required for the A2 Investigation, removal of the shielding block placed in front of
the X-6 penetration via remote control is required. Technology for this purpose (Equipment for removing the X-6 shielding
block) is currently being developed (Photo 2).
Preparations are currently underway in preparation for veriﬁcation testing at Unit 2 in the ﬁrst half of FY2015.

c. Technology for accessing the pedestal exterior

A shape-changing robot to investigate the pedestal exterior inside the Unit 1 PCV (an investigation device that works on thegrating outside of the pedestal (‘B1 Investigation’) was developed (Photo 3).
This robot takes a tubular form as it travels along the existing guide pipe (inner diameter about 100 mm) placed at the X-100B
penetration, and changes into a U shape once landing on the grating* on the 1st ﬂoor inside the PCV in order to achieve stablemaneuvering. It is expected that this will be the ﬁrst robot to investigate the inside of the PCV of Unit 1, and veriﬁcation testingis planned for the ﬁrst half of FY2015.
* Grating: grid-like ﬂoor

d. Technology for measurement of fuel debris
Development of fuel debris measurement equipment is underway to understand the position and distribution of moltenmaterials assumed to be fuel debris in extremely harsh environments; darkness, high radiation, rain drops and fog. Taking thissevere environment inside the PCV into account, this equipment will employ an optical cutting method and development is
progressing towards veriﬁcation tests after FY2016.

12Photo 1 A2 investigation device

To be used inside Unit 2 pedestal.

Figure Devices currently being developed and where they will be used.

Future Developments

Development will continue in the lead up to veriﬁcation tests from the X-100B penetration in Unit 1 and the X-6 penetration in Unit
2 scheduled for the ﬁrst half of FY2015. In addition, device development to further investigate both the interior and exterior of thepedestal inside PCVs will be ongoing.

IRID Annual Research Report 2014 13

Subsidized Projects

It is estimated that reactors cores in Units 1-3 have melted
and fuel has partially fallen with reactor internals into the RPV
and PCV. In particular, it is possible that after fuel debris
melted through the bottom of the PCV in Unit 1, they
emerged from the inside of the pedestal (which supports thePCV) and spread out of the pedestal opening. However, theactual condition has not been conﬁrmed.

Main R&D
Achievements

Background

Traveling on grating mode

Traveling through guide pipe mode

Development of Technology for Investigation Inside the Reactor
Pressure Vessel
Multistage cylinder

Aims

regarding the location, shape and condition of fuel debris andbe optimized based on related projects and after compiling

In order to develop equipment for their retrieval, information

addition to the complex reactor internal structure, radiationlevels are extremely high, creating a situation that makes it
currently difﬁcult to directly obtain information from the insideof the RPV.

The overall plan for the survey of the interior of the RPV willneeds on the construction site. Speciﬁcally, in addition to

Subsidized Projects

core internals must be understood in advance. However, in

Main R&D
Achievements

Background

Extension cylinderconsidering at an early stage survey methods in order toaccess the RPV, the development of technology of survey
equipment and devices will take place.

Main Achievements

1. Formulation of plan for investigation inside the RPV

Based on related projects such as “Development of technologies for retrieving fuel debris and core internals” and the collating ofneeds (including fuel debris sampling) from the site (TEPCO), survey locations and items, and the time required for and necessityof surveys were identiﬁed.

(1)水圧式走行装置

Hydraulic traveling device

Figure 1 Hydraulic traveling device

As a result, survey needs (amount and scope of fuel debris located in the core or lower part of the reactor, etc.) were consideredbased on the progress of research and development for each project. In addition, the selection of the method of fuel debris

retrieval from the site and survey needs (temperature inside the RPV/cooled state of fuel debris, etc.) towards each construction
method and equipment design milestone were clariﬁed.

In regard to the sampling of fuel debris, the timing of implementing this task and its purpose and positioning were set out andcoordinated with the 'Analysis of Debris Properties' project, etc., along with a clariﬁcation of the division of roles.

2. Study on investigation methods and Formation of development plan for investigationequipment

A survey of technology that can be applied to investigation inside the RPV stemming from technology developed in the‘Development of Technology for Investigation inside the PCV’ project and from a FY2013 IRID Request for Information (RFI) was
conducted.

Based on this survey, a request for technologies as below that can be applied to investigations inside the RPV was made both

Crawler
Figure 3 Shield plug hole boring device (image)within Japan and internationally, and the feasibility of these technologies considered:(1) Transfer technology of investigation equipment (small diameter pipe expansion/transfer technology)(2) Investigation support technology (radio communication in highly radioactive environments)

3. Development of investigation devices/equipmenta. Survey technology that utilizes existing routes (pipes, etc.)

Prototype testing of survey devices using existing large-diameter pipes took place. During this ﬁnancial year, veriﬁcation ofaccessibility was conducted by studying travel motion (running devices in horizontal/vertical/elbow situations or through pipes

PCVヘッ
ド
PCV
head

of various diameter), ability to grip surfaces (maintenance of position and posture), and handling of direction change (running

Multi-node link
(pantograph)through T-junctions, directional control). By assuming that devices will operate in an environment that will expose them to up to
1000 Gy/h, movement mechanisms of a hydraulic (Figure 1) or electric (Figure 2) type was studied.

b. Survey technology that utilizes new routes (opening of holes, etc.)

Element testing took place to verify the feasibility of drilling holes via remote control for the purpose of accessing the inside of
the RPV by opening a hole at the top of the reactor containment vessel (Figures 3 and 4).

Future Developments

Based on ﬁndings from survey of onsite needs in FY2014, development of technology will continue in the lead-up to early
investigation of the reactor well in FY2016 and survey of the RPV interior in and after FY2018.

14

Figure 2 Electric traveling deviceノズル
Nozzle(2) Electric traveling device
Cut fragment
adsorption
mechanism
切断片吸着機構
RPV保温架台
RPV
insulation
frame

RPVヘッ
ノズル
RPV
headドnozzle
Figure 4 RPV head hole boring device (image)

IRID Annual Research Report 2014 15

Identifying Conditions Inside the Reactor through Application of
Severe Accident Analysis Code
Aims

process of formulating measures for fuel debris retrieval and(MAAP, SAMPSON) in order to elucidate accident progression

Identifying conditions inside reactors is essential in theinvestigate the inside of Units 1-3 at the Fukushima Daiichi
NPS directly due to the extremely high radiation levels
present.

This project aims to enhance severe accident analysis codesand promote decommissioning efforts through the sharing of

Subsidized Projects

developing safety plans. However, it is extremely difﬁcult toMain R&D
Achievements

Background

Fuel/structural material
that has melted andsolidiﬁed as fuel debris

information required for identifying conditions inside reactorsthat are gained from accident progression analysis, testing,and so on.

Main Achievements

1. Improvement of the MAAP code and accident analysis

Weight distribution of fuel debris

MAAP
is an analysis code that enables the consistent evaluation of thermal-hydraulic（Modular Accident Analysis Program）

MAAP
SAMPSONphenomena/ﬁssion product behavior in the RPV or PCV during severe nuclear power plant accidents.

Improvements were made on a physical phenomenon model assuming that the molten core ﬂowed down and was deposited in

BSAF（9 organizations)

the RPV lower plenum after the core was damaged, based on accident progression at the Fukushima Daiichi NPS and analyses of

on-site surveys. Then, an accident progress analysis was conducted (Figure 1). From the results of this analysis and on-site
investigations, an estimate was made of the distribution of fuel debris in the RPV and PCV (Figure 2).

2. Improvement of the SAMPSON code and accident analysis

Core (%)
00
0〜2

RPV bottom (%)
Approx. 10
00〜5PCV (%)
Approx. 90
100
70〜100Figure 1 Improvement of MAAP Analysis Model

Figure2 Results of estimation of conditions inside Unit 1

Improvements were made to a physical phenomenon model that treated the
behavior of molten core ﬂowed down, deposited in the RPV lower plenum, andreleased to outside of the RPV after the core was damaged.

It is estimated that most of the fuel debris in Unit 1 has fallen down to the ﬂoor of thePCV. (Values shown represent ratio compared to the weight of all structural materialpresent in the reactor core prior to the accident. The sum of all percentages of lessthan 100% indicates that a certain proportion did not become debris and remains
intact.)SAMPSON (Severe Accident analysis code with Mechanistic, Parallelized Simulations Oriented towards Nuclear ﬁelds) is a model
constructed using multi-dimensional mechanistic models and theoretical-based equations that is suited to analyzing the
In order to improve the accuracy of estimates regarding the dispersion of debris contained in the lower part of the RPV, functional

improvements to models that combine the RPV and PCV, and improvements to the model analyzing thermal-hydraulic behavior

Initial cavity

Initial cavity

the bottom of the RPV evaporates, in-core monitor housing is the ﬁrst site of damage (Figure 2). In order to verify this, tests using
real-size penetrating pipes and real corium is planned for the next ﬁnancial year.

3. Analysis and evaluation of conditions Inside the RPV and the PCV

a reference the Molten Core Concrete Interaction (MCCI) analysis module in the SAMPSON code and also through reviewing theﬂuidic cell boundary condition setting model. CCI test analyses from the OECD/MCCI projects were performed in order to verify
the developed model. The results of an analysis of the depth and shape of concrete erosion match test results well, as Figure 3
shows, and conﬁrmed the predictive performance of the developed model.

4. Tests to simulate the conditions inside the reactor during the severe accident

Initial cavity

Debrissurface
Corrosion
surface

DebrisConcreteSpace

temperature (corresponding to cladding tube surface temperature) and ﬂuid temperature grew with an increase in the amount ofheat per unit area (heat ﬂux), as Figure 4 shows. It was also found that this difference grew in relation to the increase inconcentration of a sodium chloride solution.

5. International cooperation concerning identifying the condition inside the reactor

The Benchmark Study of the Accident at the Fukushima Daiichi Nuclear Power Station (BSAF) Project commenced in 2012. A
project report containing summarized results of an analysis conducted by the eight participating countries that included a study ofthe location and amount of fuel debris was released this year (Figure 2). The scope of the BSAF Project will be expanded to include
an analysis of the migration behavior of ﬁssion products in Phase 2, which will also see 12 countries participating.

流
体
温
度
[K]

熱流束(kW/m
Heat ﬂux (kW/m2)2)

Testing was done to determine the effect that injecting seawater into a reactor, as took place in the Fukushima Daiichi NPS
accident, has on thermal-hydraulic behavior. As an example test result, it was found that the difference between the wall

N aC l水溶液濃度0% (pure water)
0%(純水)
3.5%
7%
10%
15%
20%

NaCl aqueous solution concentration

壁
面
温
度

were made. An analysis using a molten core relocation analysis module yielded prediction results that suggest that, after water in

A 3D analysis model that enables analysis of debris lateral erosion inside the concrete ﬂoor like sump pit was developed, using as

Wall temperature – Solution temperature (K)dispersion and properties of fuel debris.

Figure 3 Pit erosion behavior evaluation example

Results of an analysis of concrete erosion depth and shape agree with the ﬁndingsof the OECD/MCCI project CCI-2 test, and conﬁrmed the prediction performance of
sophisticated models.

Figure4 Figure 4 Example of test results using sodium chloride solution (Temperature

difference between wall and solution)

As heat ﬂux was raised, wall temperature increased and the temperature differencewith the solution became greater. In addition, the temperature difference between
the wall and the solution also rose as the concentration of the sodium chloride
solution increased.

Future Developments

The sophistication of severe analysis codes will be further advanced, and an improvement in accuracy of predicting the location offuel debris and migration of ﬁssion product nuclides will be sought. In addition, utilizing data obtained from the actual reactors,
comprehensive analysis and evaluation of conditions within reactors will be executed, the results of which will be used to conﬁrm
and implement methods of fuel debris retrieval at each unit.

16

IRID Annual Research Report 2014 17

Development of Technology for Detection of Fuel Debris
in the Reactor
Aims

This project uses cosmic-ray muons to image and map the distribution of fueldebris inside the RPV from outside the reactor buildings. Using the muontransmission method, it is hoped that fuel debris at Unit 1 can be identiﬁed at the
spatial resolution of about 1m and the data obtained will be used to study
methods of how fuel debris can be retrieved. A measurement system that uses
the muon scattering method will be used to conﬁrm the location of fuel anddebris remained in the reactor at Unit 2 (the spatial resolution of about 30 cm).

10-day measurement
Figure 1 Muon detector which uses transmission method to be installed at

the Fukushima Daiichi NPS Unit 1
Three-layered scintillation detector is housed in an enclosure with iron walls
approximately 10cm thick to shield against background gamma rays. Temperatures
are kept at a constant level inside the enclosure through an air conditioning unit.

60-day measurement

Figure 2 Result of transmission method simulation

Using the Monte Carlo method, measurement days and image clarity were
examined. As the image contrast is proportional to the density of muon travelinglength, the longer the measurement time, the clearer the image contrast becomes.

Main Achievements

Testing conﬁrmed appropriate shielding thickness that enables the effective measurement of muons even in radioactive
environments (0.4mSv/h). In addition, a measurement system that can reduce the effects of background gamma radiation by
conducting coincidence counting using three-layered 2D resolution scintillation detectors to determine muon transmission
position was developed. This measurement system was placed outside the Unit 1 reactor building and the measurement to image
inside the reactor was conducted. Furthermore, an evaluation of the transmission method measurement performance wasconducted (Figures 1 and 2).

Upper muon tracking device

Counting rate (relative value)

1. Development and measurement of a detection system for the transmission method underhigh radiation

Filler gas containing hydrocarbon

Converted period under the condition of 50μSv/h(Month)2. Development of a large muon tracker under high radiation

Large-scale muon tracking devices with 7 x 7 m2 active sensor areas that use the muon scattering method to image the inside of
the RPV from outside the reactor building were developed. The muon trackers are comprised of 3.5-meter drift tube detectors thatare joined in two end-to-end to make chambers of 7 m in length. There are 140 of these tube chambers that are arranged to forma 7 x 7 m2 active sensor area in a six-layered stack. The gas encapsulated in the detectors is of a type that does not containhydrocarbons that can be broken down by gamma rays, ensuring resistance to radiation during the period of measurement(Figure 3).

Lower muon tracking device

Figure 3 Production of highly radiation-resistant drift tube detectors.

Figure 4 Muon tracking device system performance test platform

Manufacturing of the drift tube detectors (effective length of 3.5m, as shown at right)
employing an encapsulated gas mixture (Ar:CO2:N2=96:3:1) that enables operation
without any reduced sensitivity, even after exposure to gamma radiation over theequivalent of a four-year period at a maximum design dose rate of 50μSv/h.

Muon tracking devices with a 7 x 7m2 active sensor area comprised of 3,360 drift
tube detectors (3.5m long) were constructed and placed one above the other. Leadand other materials were placed between them in order to evaluate system
performance.

Two of the muon trackers were placed apart, one above the other, in order to conﬁrm measurement performance (Figure 4).

A method of estimating the distribution of fuel debris from measurement values taken at Unit 2 by muon tracking devices wasdeveloped. Speciﬁcally, taking angular resolution, position resolution and the effect of shielding material into account, the muon
scatter angle is evaluated, and fuel debris density contrast images are created by the relationship between the structurecomposition and the scattering angle of muons as they pass through the structure. In order to evaluate the accuracy of thisdevelopment approach, the simulation test took place using a 3D model of the reactor; results were then used to predictmeasurements of the actual site (Figure 6).

Future Developments

While expanding the scope of measurements using the transmission method at Unit 1, detection of fuel debris at Unit 2 (lower
part of the RPV and the reactor core) will be conducted using the scattering method. Furthermore, by improving processing
algorithms and measurement techniques, the visual range will be expanded, measurement time reduced, and more detailed
information provided by resolution improvement. This will contribute to the establishment of measurement technology for the
distribution of fuel debris inside the reactors and the selection of efﬁcient methods of fuel removal.
18

Gamma ray reaction

4-layer coincidence

Pattern + 4-layer coincidencea) Coincidence counting gamma ray removal rate

Event count

4. Development of an estimation technique of fuel debris distribution inside the reactor bymuon tracking

Muon reaction

Event countAs detectors that utilize the scattering method are large in scale, time-coincidence logic incorporated into the electronic circuit toremove high background gamma rays was developed. A combination of coincidence counting in more than four layers and linear
reaction patterns allows muon measurements without shielding even at the dose rate of approximately 50 μSv/h. Additionally, interms of measurement performance, time resolution of less than 2ns (position resolution of less than 0.1 mm) and a coincidence
counting gate width of less than 1μ were achieved.

Muon identiﬁcation rate (%)

3. Development of a muon measurement system under high radiation (less than 50 μSv/h)

Pulse widthb) Time resolutionc) Angular resolution

Figure 5 Results of circuit system performance evaluationa) Even under a gamma radiation ﬁeld of 50μSv/h, it was possible to detect over 90%
of muons using coincidence counting and linear reaction patterns integrated into
the circuitry.
b) With a 200ns pulse width, was achieved the gate width of less than 1μs required
to remove gamma radiation.
c) A time resolution (FWHNM) of under 1ns was conﬁrmed, against the target lower
than 2ns.

Figure 6 Results of scattering method simulation

Predicted results from 7m x 7m muon tracking devices placed above one another inproduction level conditions to place them outside turbine or reactor buildings(measurements at 20, 40 and 80 days)

IRID Annual Research Report 2014 19

Subsidized Projects

In order to select effective methods offuel retrieval, it is important to understandconditions inside the reactors, including
the distribution of fuel debris. However,as the radiation level inside the RPV isextremely high, it is very difﬁcult to verify
the condition directly.

Main R&D
Achievements

BackgroundDevelopment of Technology for Non-destructive Detectionof Radioactive Materials Accumulated in the Suppression Chamber
Aims

In order to gather information required for
repairs and water leakage stoppage in the
S/C, in addition to estimating the conditionof accumulated radioactive material present
inside the S/C, a method of measurement
will be developed.

Main Achievements

1

Design/production of measurement systems

2

Design/production of access devices

3

Combination performance testing

4

Design and production of perforating apparatus

5

Veriﬁcation tests

6

Non-destructive detection work

Table 1Analysis target
regionRadioactive material deposits
in S/CS/C lower portion enlarged view
Figure 1 Computational scheme

Development/Work items

In the formulation of development plans, development/work items
required for the detection of radioactive materials were identiﬁed inadvance.

1. Formulation of development plan

Developments and work items required for the detection of radioactive materials believed to be indispensable were extracted(Table 1) and development plans were formulated.

2. Radioactive materials migration scenarios

Consideration was given to scenarios involving the migration of radioactive materials into the S/C and torus room (Figure 2). As itis thought that the likelihood of an inﬂow of radioactive debris that exceeded the acceptable limit was low, by taking
measurements of the lower part of the S/C and the region around the sand cushion drain pipe outlet where there is a relatively
high probability of radioactive materials having been deposited, it can be conﬁrmed that there are no radioactive materials thatexceed the acceptable limit.

3. Evaluation of impact of radioactive substances on water stoppage material

Among the inﬂuence factors arising from remaining radioactive materials, there is a concern that even the smallest deposit mayheat water to a temperature of 80°C, and the heat generated within water stoppage materials may cause cement deterioration.
Calculations were made based on a conservative estimate of the weight of uranium in this scenario.

Form

S/C

Torus Room

Meltedradioactive
substancesRadioactive substances in the drywell reach
the S/C vent pipe entrance and ﬂow into the S/C.

Radioactive substances in the drywell erode the PCV
shell and sand cushion and ﬂow in via the drain pipe.

Pulverizedradioactive
substances

Material moves due to the ﬂow resulting from the injectionof coolant water and ﬂows in via the S/C vent pipe.

Material moves due to the ﬂow resulting from the injectionof coolant water and ﬂows in via the sand cushion drain pipe.

Aerosols

Material moves due to the ﬂow of gases,
and ﬂows in via the SR piping or S/C vent pipe, etc.

Material moves due to the ﬂow of gases, and traversing
the S/C, ﬂows in via damaged vacuum break line, etc.

Table 2

Scenario where there is a relatively high probability of an inﬂow of highly radioactive material

4. Development of technology for detection of radioactive material
a. Evaluation on nuclide composition and distribution of radioactive material

Nuclides originating from fuel (nuclides selected for measurement (Cm-244, Eu-154, etc.), background nuclides, and nuclides in shielding
material) were determined using the ORIGEN code. The mix ratio of nuclides originating from fuel and structural materials has been set basedon MAAP code analysis results. The S/C and torus room calculation model (Figure 1) is a 3D 1/16 scale representation of the S/C and torus
room; this model was used to evaluate neutron and gamma ray ﬂux in the area around the bottom of the S/C (Figure 2).

b. Evaluation of background radiation

An assessment of the effect of background gamma radiation (Cs-134, Cs-137) on stagnant water was conducted (Figure 3).

c. Selection of optimum techniques for detection of radioactive substances

To measure acceptable background radiation level and radio-sensitivity, a B-10 neutron detector and CdTe/LaBr3(Ce) gamma-ray detectors
were selected as the best choice for the task and their responses evaluated at the locations in a. and b. above.

Accumulated
water levelin the torus
room

ﬁlled with waterd. Study on method of estimating the amount deposited in the S/C

A comparison between measurements done in the vicinity of locations with a high probability of radioactive material deposits and thresholdlevels set through the analysis in a. will decide the presence or absence of radioactive material that exceeds acceptable levels.

Future Developments

The possibility of radioactive material that exceeds acceptable levels having ﬂowed into the S/C or torus room is considered low,
and it was conﬁrmed that it is technically feasible under the assumed conditions to verify the presence or absence of this materialusing non-destructive methods. The implementation of measurement systems or design/production of devices to access theseareas will be decided based on the results of developments of methods to repair and stop water leakage.

20

Cm-244 measurements

Eu-154 measurements

Cs- 137 measurements

Figure 2 Distribution of radiation originating from fuel in the vicinity of the S/C (Unit 1)

Figure 3 Background radiation (Unit 1)

With the radiation source established as fuel debris, neutron ﬂux and gamma ray ﬂux at the bottom of the S/C
were evaluated. It was conﬁrmed that, even in the background environment (Figure 3), Cm-244 (neutrons) andEu-154 (gamma radiation) could be measured.

After evaluating background gamma radiation instagnant water, it was found that radiation levels
were high (red zone: around 10-7/cm2/source)
around the stagnant water in the torus room. It
was also conﬁrmed that gamma ray ﬂux was
almost of a uniform level in both the S/C andtorus room.

IRID Annual Research Report 2014 21

Subsidized Projects

In order to repair and stop water leakage in the suppression chamber (S/C), it isnecessary to determine the condition of radioactive material that has accumulated
inside. However, a method of evaluating conditions has not yet been established. The
optimum solution would be a non-destructive examination technique, but it is not clear
if deposits of radioactive material that need to be detected can be evaluated using a
non-destructive method. Another important issue is to establish approaches that candetermine the amount deposited throughout the entire area, including the S/C.

Content

Main R&D
Achievements

BackgroundNo.

Development of Technology for Retrieval of
Fuel Debris and Reactor Internals
Partialsubmersion

Partiallyin airCompletely
in air

Feasibility ofwatertightness
is a challenge

Representative method 2

Representative3

Identiﬁcation ofchallenges whenrequired

Feasibility of cooling air is a challenge

1 Situation prior to commencement of work
2 Removal of shield plug

In addition to the technical catalogs*, the applicable technologies for debris retrieval were extracted from the types of equipmentused at TMI-2, Sellaﬁeld and Paks, and existing technologies and proven technologies from outside the nuclear energy ﬁeld (e.g.,
medicine).

・Negative pressure management (inside well)

Dose rate on
operation ﬂoor:
under 1mS/h

Dust scattering

Radiation

・Negative pressure management

・Airtight function of shielding
Scatterprevention ・Isolation of areas on transfer routes

・Application of anti-scattering agent/sheet

4 Removal of lagging material and RPV head・Evaluation of required shielding amount・Shielding structure (e.g., installation of heavy
structure, opening and closing mechanism)
・Feasibility study on transfer methods and procedures・Negative pressure management (inside well)Figure 2 Use of isolation sheet for preventing the spread of

Equipment

・PCV repair (water stoppage/maintenance)(The extent of the repair depends on target level)
・Water circulation system/Water boundary
construction
・Ventilation system/Negative pressure boundary
construction

Criticality

・Installation of criticality monitoring system
・Installation of neutron absorber injection
equipment

Investigation ・Identifying position of debris

・Evaluation of required shielding amount・Shielding structure (e.g., installation of heavy
structure, opening and closing mechanism)
・Feasibility study on transfer methods and procedures・Negative pressure management

Scatter ・Airtight function of shielding
prevention ・Isolation of areas on transfer routes

・Application of anti-scattering agent/sheet

Decontamination ・Development of remote decontamination equipment (if necessary)

Equipment

・Radiation-resistant equipment
・Remote maintenance/replacement method
for equipment/parts
・Method of equipment recovery in emergency

6 Removal of core internals/fuel debris
Shielding・Evaluation of required shielding amount・Shielding structure (e.g., installation of heavy
structure, opening and closing mechanism)
・Feasibility study on transfer methods and procedures・Negative pressure management

Scatterfunction of shielding
prevention ・Airtight・Isolation of areas on transfer routes

Transfer・Prevention of criticality in retrieved debris (canister)・Material accountancy of retrieved debris

・Development of remotely operated equipment
・Processing equipment for heterogeneous/very
hard objects
・Retrieval technology for secondary products such
Equipment as aerosols
・Radiation-resistant equipment
・Remote maintenance/replacement method forequipment/parts
・Method of equipment recovery in emergency

Criticality

・Installation of criticality monitoring system
・Installation of neutron absorber injection equipment

7 Removal of lower core internals/fuel debris
ShieldingShielding・Evaluation of required shielding amount・Shielding structure (e.g., installation of heavy
structure, opening and closing mechanism)
・Feasibility study on transfer methods and procedures

Scatterprevention

・Negative pressure management
・Airtight function of shielding
・Isolation of areas on transfer routes
・Application of anti-scattering agent/sheet

contamination

・Development of remotely operated
equipment
・Study of cutting procedure/method

Equipment ・Radiation-resistant equipment

・Remote maintenance/replacement method forequipment/parts
・Method of equipment recovery in emergency

Criticality

・Installation of criticality monitoring system
・Installation of neutron absorber injection equipment

9 Removal of components in pedestal/fuel debris
Shielding・Evaluation of required shielding amount・Shielding structure (e.g., installation of heavy
structure, opening and closing mechanism)
・Feasibility study on transfer methods and procedures・Negative pressure management

Scatter・Airtight function of shielding
prevention

・Isolation of areas on transfer routes
・Prevention of criticality in retrieved debris (canister)・Material accountancy of retrieved debris

Transfer

Equipment

Criticality

In addition, measures to address issues identiﬁed in ‘2. Designing a plan for ﬁnalizing debris retrieval methods’ were reviewed anda development plan was formulated.

Transfer

5 Removal of steam dryer/ separator
Shielding

Decontamination ・Development of remote decontamination equipment (if necessary)

8 Removal of RPV bottom

Retractable shieldwork area

・Actualization of equipment design (removal ofequipment interfering with work, installation of
cranes/work cells, etc.)
・Ensuring debris/waste material destination
・Feasibility of transfer routes (opening dimensions,shielding, prevention of scatter, etc.)

・Airtight function of shielding
Scatterprevention ・Isolation of areas on transfer routes

・Application of anti-scattering agent/sheet(4) The material selection and handling tests of isolating sheet for preventing expansion of contamination (Figure 2).
(5) Prototyping and operational testing of remotely operated arm for supplementary work under very high radiation (Figure 3).

・Development of remotely operated equipment (for work 34m
below the operation ﬂoor )・Processing equipment for heterogeneous/very hard objects
・Retrieval technology for secondary products, such as aerosols
・Radiation-resistant equipment
・Remote maintenance/replacement method for equipment/parts
・Method of equipment recovery in emergency

・Installation of criticality monitoring system
・Installation of neutron absorber injection equipment

・Evaluation of required shielding amount・Shielding structure (e.g., installation of heavy
structure, opening and closing mechanism)
・Feasibility study on transfer methods and procedures・Negative pressure management

Scatter ・Airtight function of shielding
prevention

・Isolation of areas on transfer routes

Transfer・Prevention of criticality in retrieved debris (canister)・Material accountancy of retrieved debris

・Development of remotely operated equipment
・Processing equipment for heterogeneous/very
hard objects
・Retrieval technology for secondary products such
Equipment as aerosols
・Radiation-resistant equipment
・Remote maintenance/replacement method forequipment/parts
・Method of equipment recovery in emergency

Criticality

・Installation of criticality monitoring system
・Installation of neutron absorber injection equipment

10 Removal of components outside pedestal and fuel debris
Shielding・Evaluation of required shielding amount・Shielding structure (e.g., installation of heavy
structure, opening and closing mechanism)
・Feasibility study on transfer methods and procedures・Negative pressure management

Scatter ・Airtight function of shielding
prevention

・Isolation of areas on transfer routes

Transfer・Prevention of criticality in retrieved debris (canister)・Material accountancy of retrieved debris

Equipment

・Development of remotely operated equipment (accessfrom above)
・Processing equipment for heterogeneous/very hard objects
・Retrieval technology for secondary products, such as
aerosols
・Radiation-resistant equipment
・Remote maintenance/replacement method forequipment/parts
・Method of equipment recovery in emergency

Criticality

・Installation of criticality monitoring system
・Installation of neutron absorber injection equipment

Challenges and ways to proceed in development

Shielding and scattering (spread of contamination) prevention are issues common to all steps and of high priority.
-> The issues are related to processing devices, lifting equipment, work cells and negative air pressure control equipment used in each work step.
Review of entire work process is required.
(Will be effective if carried out conﬁrming method feasibility, identifying problems, and points to improve at each work step and overall, using a model, etc.)

Future Developments

22

・Evaluation of required shielding amount・Shielding structure (e.g., installation of heavy
structure, opening and closing mechanism)
・Feasibility study on transfer methods and procedures

Shielding(2) Prototyping of simulated debris for use in cutting evaluation tests, etc.
(3) Position control characteristics evaluation test for access equipment for use in remote controlled work.

(3) Study of fuel debris retrieval method/system/equipment

・Radiation-resistant equipment
maintenance/replacement method
Equipment ・Remote
for equipment/parts
・Method of equipment recovery in emergency

Equipment ・Positive/negative pressure switching methods inside PCV(1) Evaluation test for cutting considering fuel debris collection.

(2) Formulation of development plans for related element technologies and equipment
i. Investigation/study of existing technologies
ii. Formulation of development plans for related element technologies and equipmentiii. Element test/technology development
iv. Feasibility evaluation on the actual unit/mock-up

Decontamination ・Development of remote decontamination equipment (if necessary)

Decontamination ・Development of remote decontamination equipment (if necessary)

Isolation sheet

・Evaluation of Cs decontamination distribution

reduction)
Decontamination ・(dose
Dose reduction including decontamination of・Application of anti-scattering agent/sheet

3 Removal of PCV head

Element testing for technical issues that the twelve scenarios/methods have in common was implemented.

(1) Setting of conditions towards fuel debris retrieval method policy decision

・Airtight function of shielding
Scatterprevention ・Isolation of areas on transfer routes

Shielding

4. Formulation of development plans for related element technology and equipment

Centered on the important issues of shielding/preventing spread of contamination and remote automation, technology
development for the following items will be conducted, looking towards the presentation of a debris retrieval method (procedure)plan in late FY2016.

・Evaluation of required shielding amount・Shielding structure (e.g., installation of heavy
structure, opening and closing mechanism)
・Feasibility study on transfer methods and procedures

Shielding

In reviewing debris retrieval, methods were classiﬁed into twelve types in terms of the PCV water level and the direction that debris
would be accessed from; issues were identiﬁed and organized for three representative methods thought to be the most practical(Figure 1). Decision items at the time when debris retrieval methods are used at each plant were pointed out, and eachcorresponding situation arranged (Table 1).

3. Survey and review of existing technologies

Method step
diagramFigure 1 Representative methods and scheme of challenges extraction

1. Setting of conditions for deciding on fuel debris/reactor internals (hereinafter, ‘debris’)
retrieval technology

2. Designing a plan for ﬁnalizing debris retrieval methods

Challenges/risk
evaluation for each
representative
methodRepresentative method 3
Representative
2

Main Achievements

Consideration of the method of debris retrieval must be based on the situation at each plant and the information obtained fromother related projects. Therefore, questions as to what kind of requirements or plant-related information existed and by whenthese would be clariﬁed were set out. Additionally, in preparation for the commencement of actual fuel debris retrieval in
December 2021, the required timing and accuracy of this information was clariﬁed.

Representative method 1

Subsidized Projects

In the lead-up to the commencement of the mock-up test offuel debris retrieval methods (planned for FY2018) based on
the Mid-and-Long-Term Roadmap (June 2013 revision),
speciﬁc R&D will be undertaken. Under this project, plans for
retrieval methods at each plant will be drawn up, anddevelopment plans for required equipment and element tests
created. In addition, element tests of common technology will
be conducted and included in the FY2015 plan.

Complete
Representative
submersion
1

Identiﬁcation of representative method challenges

From side From bottom Identiﬁcation ofrepresentative
method challenges

Compared to other devices, the highly radioactive environment around the reactor core presents requirements of remote automated devices thatare more severe than with other devices and a higher degree of difﬁculty.
There is a need to develop essential technology, etc., through partial mock-up testing, and to improve device feasibility at an early stage.
(Preparation for actual devices is, from looking at method and device feasibility prospects, without waste and efﬁcient.)
Preparation for mock-ups needs to proceed as planned for the purpose of device testing and worker training.
Development of safety-related design criteria, examination of safety equipment, and safety evaluations (accident assumption, risk evaluation) are required to be undertaken.

Figure 3 Handling test on remotely operated working arm(artiﬁcial muscle robot)

Main R&D
Achievements

It is believed that fuel debris in the RPV and PCV at the
Fukushima Daiichi NPS has not reached criticality at thepresent time. However, reactor buildings, RPV and PCV
suffered damage during the accident, and the plant itself
remains in an unstable condition. It is therefore vital that
re-criticality is prevented while retrieving fuel debris, and theplant is kept stable to avoid the dispersion of radioactive
materials.

Aims

From top
Water level

Background

Access direction

Table 1 Important issues and development procedures for each work step towards retrieval of fuel debris.

Required work processes for the methods of fuel debris retrieval (three representative methods) weredivided into ten steps and challenges identiﬁed. Measures to resolve these challenges were examined andorganized, and ways forward combined into four categories.

* The Agency for Natural Resources and Energy will conduct surveys and call for technical proposals publically for the FY2011 ‘Technical catalog for the developmentof equipment to prepare for fuel debris removal towards the Decommissioning of the TEPCO Fukushima Daiichi Nuclear Power Station’ as a project related to theSubsidy for Development of the Technological Foundation for Decommissioning and Safety of Nuclear Reactors for Power Generation.

IRID Annual Research Report 2014 23

Development of Technology for Collection, Transfer and Storage ofFuel Debris

to be carried out of reactor buildings and stored until the timedebris will be based on the established spent fuel transportation

Fuel debris retrieved from the Fukushima Daiichi NPS is expected

when the method of ﬁnal disposal is determined. For this reason,technology required for the collection, transfer and storage of fueldebris must be established at an early stage.

Development of storage canisters for the housing of retrieved fuel3. Preparation for transfer
outside (categorization)

4. Transfer (outside building)TMI-2
experience

Submerged from above

Transferin the air

No Categorization/
No volume reduction

Half-dry transfer

1F planned
scenario(1) Submerged
from above(1) Transfer
in water(1) Categorization/
Volume reduction(1) Wet transfer(outside)(2) Transfer
in the air(2) No Categorization/
No volume reduction(2) Half-dry transfer(outside)and storage technologies. In FY2014, development requirements
for storage canisters at the Fukushima Daiichi NPS will beInterim storage(wet storage)(1) Categorization/
Volume reduction

organized, and canister design approaches and basic structuraldesign concepts will be derived.

Main Achievements(2) Non-submerged
from above(same as above)(3) Non-submerged
from side(same as above)

1. Study on a fuel debris collection, transfer and storage system(same as above (1))(2) No Categorization/
No volume reduction

Temporarystorage(wet storage)

Long-term
direct storage(dry storage)

Subsidized Projects

Aims

2. Transfer (inside building)Main R&D
Achievements

Background

1. Fuel debris retrieval(same as above (2))

Figure 1 Scenario for collection, transfer, and storage of retrieved fuel debris (partial)

Scenarios for the collection, transfer and storage of retrieved fuel debris studied were as follows: Collection (Submerged/Non-submerged

The measures in the red boxes represent an example of scenarios that are expected to be of merit in terms of safety, work
efﬁciency, and feasibility. However, there are many unknowns about fuel debris characteristics and as knowledge increases, it isexpected that the degree of difﬁculty and merits of each scenario may change. It is for this reason that technology development
will involve taking technical elements in common without narrowing them down, and reviewing the order of priority.

from above/Non-submerged from side) -> Transfer (Wet/Dry/Half-dry) -> Storage (Wet/Dry/Half-dry). In addition to conﬁrmingthat all scenarios can be feasible, scenarios that are expected to be advantageous were selected in terms of safety, work efﬁciency
and feasibility, etc., for each individual step (Figure 1).

2. Establishment of a storage canister design concept

After studying the work involved in each step of the scenarios described above, it was found that as the collection of fuel debris
would take place in highly radioactive environments or in narrow spaces, emphasis would need to be placed on operability in thedesign process. For this reason, it was thought that the safety function of the storage canisters should be minimum, and the safety
would be guaranteed by the surrounding equipment and installations instead. In future, adjustments will progress taking
surrounding equipment and installations into consideration (Table 1 and 2, Figure 2).

3. Development of safety evaluation methods

Evaluation methods for determining the shape/conﬁrming applicability of storage canisters were examined. In principle, thedesign methods for spent fuel transportation casks or the plant design can be used. However, it was found that for part of the safety
evaluation, veriﬁcation data would need to be further expanded. Additionally, it was seen that there are many items dependent on
the characteristics of the fuel debris itself (temperature limits, etc.) (Table3).

4. Investigation on the transfer and storage of damaged fuelIn order to carry out 1-3 above, technical information related to the transfer and storage of fuel debris from Three-Mile Island
Reactor 2 (TMI-2) in the US were collected, and sub-criticality evaluation technology, drying technology and technology to deal

Safety function

Relevant considerations

Basic storage canister concept
Simple natural heat radiation

Adjustments to be made to cask baskets and facility air
conditioning, etc., in order to ensure temperature environment
surrounding fuel debris canisters.

Structures

Weight reduction focused on minimum safetyfunctions to maintain sub-criticality

Where required, make adjustments to utilize cushioning
material, etc., to avoid large stress on canisters in the event ofthem falling, etc.

Shielding

Weight reduction through non-expectation ofshielding function.

Adjustments to minimize worker exposure using
peripheral equipment and facilities.

Conﬁnement

Sealed construction not employed from theperspective of remote lid closure.

Suppression of radioactive materials into the environment managed
through improvement of airtight performance of transfer cask anduse of ﬁltering on exhaust vents at storage facility.

Sub-criticality

It is assumed that sub-criticality will be maintained
in isolation, and dimensions were kept large to theextent possible from a receiving perspective.

Sub-criticality at time of arraying storage casks to be secured
through racks and transfer container basket.

Heat removal

Hydrogen

Reduction of hydrogen amount using catalyst
process and adoption of hydrogen releasemechanism using ﬁlters, etc.

Consider dealing with hydrogen outside storage canisters
through use of catalysts or vent in the transfer containers.

Material property

Application of materials/corrosion inhibitors
suited to expected environment.

Consider materials/corrosion inhibitors. Where necessary,
exercise option to manage water quality on environment side, etc.

Fire prevention

Prevention through water injection or use ofinert atmosphere (nitrogen, etc.)

Adjustment by injecting water or deploying gas
supply facility for creation of inert atmosphere.

FilterLid

Drain pipe

Catalyst
Long type

Short type(for fuel debris retrieval in narrow spaces)

Table1 Storage canister basic concept

Figure 2 Storage canister basic concept (example)

Assuming that the storage canisters will be used in the collection (→transfer→temporary storage) →transfer→storage process, each step was ﬁrst examined from a work environment safety perspective before establishing areasonable storage canister concept.

A stylized example concept based on Table 1. Reviewed
based on handling capability in construction process.

with hydrogen gas were studied during a visit to U.S Idaho National Laboratory (INL) where fuel debris from TM-2 are stored.

Safety evaluation items

Future Developments

In order to solve issues which may become problems in developing the basic design of storage canisters, various analyses, evaluations
and element tests will be performed. Achievements in other projects will be reﬂected to fuel debris collection/transfer/storage

Issue
Parts
optimization

FlangeStructure
Filterscenarios and storage canister design.

Drain
Internal structure

Future plan

(1) Evaluation offunctions to maintainsubcritical state.

Consider prerequisites, such as determining the nuclear fuelmaterial that clouds pool water, etc.

Formation of speciﬁc plan that links with construction side forremote closing of lid, etc.

(2) Evaluation of
structural strength

Determined based on amount of radioactive material that is released
through the ﬁlter outside canisters and structural integrity to handle canister
accidents, etc.

Adjustments to prerequisites related to hypothetical accidents, etc.
Complex storage canister components veriﬁed through detailed
analysis and drop tests.

(3) Evaluation of aging
Selected based on collected case examples and testing ofdegradation of materials rust prevention, etc., through a combination of stainless-steel(including corrosion)and metal.
(4) Evaluation concerning Close examination of water absorption effects through testing, etc.
management of
Evaluation of catalyst recombination effect.
hydrogen gas

Future plan

Formation of speciﬁc design plan that links with remote waterdrainage and fuel debris removal.
Formation of speciﬁc plan of internal structure that adapts toshape of fuel debris and retrieval method.

Dimensions etc.

Determination of each dimension based on safety evaluation.

Materials

Design of materials based on study of corrosion inhibitors.

Table2 Proposed way forward for basic storage canister design (partial)(5) Evaluation of radiation
shielding function

No particular challenges.

(6) Evaluation of decay
heat removal function

Reﬂect latest ﬁndings regarding fuel debris temperature
limitation by linking with other projects, etc.

(7) Evaluation of
conﬁnement function

Reﬂect latest ﬁndings regarding amount of radiation emitted
from fuel debris by linking with other projects, etc.

Table3 Proposed way forward for safety evaluation of transfer/storage (partial)

24

IRID Annual Research Report 2014 25

Development of Technology for Evaluating the Integrity of
the Reactor Pressure Vessel/Primary Containment Vessel
Aims

Based on seismic evaluations that take the impact ofage-related deterioration due to corrosion and fallen fueldebris into consideration, an evaluation of the structural
integrity of the PCV/RPV will be made. Corrosion control
measures will be studied considering methods of fuel debris
retrieval and PCV repair/water leak prevention from a seismic
intensity perspective and used in maintaining the structural
integrity of the PCV/RPV.

Main Achievements

H26-1(In air (current
water level))

H26-2(Full submersion)・Assumed future state: Parameter(in 15 years, 40 years)
・Building damage model
・D/W water level: Well ﬁlled withwater
・S/C interior: Concrete OP3570
・Torus room: Concrete (column
support upper pin position:
OP2140)・Vent pipe: Considering repair
・Vacuum break line: Considering
repair・Additional equipment for
operations ﬂoor: approx. 6100t
・Injection of the water stoppage
material into individual room: Yes
・Attenuation constant: Parameter(1) Concrete 7%, metals 4%(regulatory guide)(2) Concrete 5%, metals 1%(for design during construction)
・Seismic waves: Current Ss

1. Evaluation of submersion method feasibility based on the seismic integrity of the PCV/RPV

A seismic response analysis model was developed for reactor building and large-sized equipment interaction that took both in air(current water level) and complete submersion methods of fuel debris retrieval into consideration under the plant conditions
reﬂecting latest plans such as repairs (Table 1). Utilizing this model, the seismic load (shear and moment) of the PCV/RPV regions
for evaluation is calculated and, assessment of fatigue strength at each region will take place and the feasibility of the submersion
method etc., examined.

2. Simple evaluation of equipment seismic resistance considering repairs (water leakage
stoppage) and water level rise in the PCV

Based on the conditions of reactor buildings and PCV/RPV following the accident and on the methods of building repair and fueldebris retrieval, the necessity of changes to the seismic response analysis model or to weight was examined, and relationships
between plant conditions and seismic response analysis parameters was compiled. In future, calculation of seismic load variation
coefﬁcients and consideration of methods to combine these will be made, and development of a simple method of evaluation willcontinue.

3. Development of anti-corrosion measures

In addition to rust-preventive agents (sodium tungstate, sodium pentaborate, etc.) identiﬁed previously, phosphates (phosphatemixed with zinc, sodium molybdate/sodium zinc oxide mixed phosphate) and sodium metavanadate were also selected as
candidates, and validation tests, including those to evaluate rust inhibiting effects under high levels of radiation exposure,commenced (Table 2). After selection of an effective rust-preventive agent, an evaluation as to whether there are any secondary
effects, such as adverse radiolysis effects or negative effects on the function of water treatment equipment, will be carried out.

4. Advancement of predictability of long-term corrosion thinning amount

1F-3 representative

Phosphate
Rust inhibitor

Sodiumtungstate

Sodiummolybdate

Sodiumpentaborate

sodiumnitriteRust inhibitor effectconﬁrmation test
results

Completedin FY2013

Completedin FY2013

Completedin FY2013

Completedin FY2013

Anticorrosion
coating types

Oxide ﬁlm type

Oxide ﬁlm type

Oxide ﬁlm type

Oxide ﬁlm type

Precipitation
ﬁlm type

Input required
for anticorrosion
effectMedium

Large

Large

Small

Small

Zinc
phosphate mix

Zinc/molybdenum
acid sodium
mixed phosphate

Sodiummetavanadate

Due to be completed Due to be completed Due to be completed
in FY2014-15
in FY2014-15
in FY2014-15

Oxide ﬁlm type +Precipitation ﬁlm Oxide ﬁlm type
type

MediumBeing
conﬁrmed

Table 2 Rust inhibitor selection result

Organization of study conditions(materials used, environmental conditions)

5. Evaluation of pedestal erosion impact(1) High temperature
heating and exposure
testing using cylindrical
specimen (concrete
constitutive law)

Testing samples were manufactured in order to conduct concrete and rebar basic materials tests under high temperature
heating/underwater exposure conditions, simulated block tests with the same thickness as found in actual RPV pedestals, and thetolerance tests with the small-scaled mock up. Tests and analyses will be performed following the RPV pedestal examination ﬂow(Figure 2), and the integrity of RPV pedestals will be evaluated.

(2) Rebar adhesion
deterioration test
(cohesion strength,
tension stiffening)(3)’ Simulation analysis of the RPV
pedestal strength evaluation test
(impact on overall behavior of
the miniature mockup)

Future Developments

26

・Assumed future state: Parameter(in 15 years, 40 years)
・Building damage model
・D/W water level: Well ﬁlled withwater (approx. 35m, OP39920)
・S/C interior: Concrete OP1900
・Torus room: Concrete(column support upper pin position: OP-100)
・Vent pipe: Considering repair
・Vacuum break line: Considering repair
・Additional equipment for
operations ﬂoor: approx. 4710t
・Injection of the water stoppage
material into individual room: Yes
・Attenuation constant: Parameter(1) Concrete 7%, steel 4%(regulatory guide)(2) Concrete 5%, steel 1%(for design during construction)
・Seismic waves: Current Ss

Table 1 Seismic response analysis cases and conditions

A long-term corrosion test (target of 10,000 hours) was initiated in order to build a corrosion thinning amount prediction model(Figure 1). Based on the data obtained from this test, a corrosion thinning amount prediction model equation will be constructed
and the corrosion thinning amount in the PCV/RPV up to the time of fuel debris retrieval will be calculated. These ﬁndings will be
reﬂected in seismic strength evaluation.

In addition to accelerate development of corrosion prevention measures, simple methods of evaluating seismic strength
developed as part of this project will continue to be leveraged. Furthermore, the examination of an evaluation model equivalent
to the method to retrieve fuel debris in air from the side and the effects of boric acid injection on the long-term integrity ofequipment will be properly evaluated.

・Assumed future state: Parameter(in 10 years, 15 years, 40 years)
・Building damage model
・D/W water level: approx. 6.5m
・S/C interior: Filled with water・Torus room: OP3200
・Vent pipes: Filled with water・Additional equipment for
operations ﬂoor: Parameter (None,approx. 4710t)
・Injection of the water stoppage
material into individual rooms:
None・Attenuation constant: Parameter(1) Concrete 7%, steel 4%(regulatory guide)(2) Concrete 5%, steel 1%(for design during construction)
・Seismic waves: Current Ss

・Assumed future state: Parameter(in 15 years, 40 years)
・Building damage model
・D/W water level: approx. 0.6m
・S/C interior: Concrete OP-1050
Water level OP3100
・Torus room: Concrete (column
support upper pin position:
OP-100)・Vent pipe: Running water in lower
portion・Additional equipment for
operations ﬂoor: approx. 4710t
・Injection of the water stoppage
material into individual rooms:
None・Attenuation constant: Parameter(1) Concrete 7%, steel 4%(regulatory guide)(2) Concrete 5%, steel 1%(for design during construction)
・Seismic waves: Current Ss(3) RPV pedestal strength
evaluation test
(miniature mockup
strength evaluation)(4) Extended immersion
corrosion rebar test
(decrease of rebar
yield strength)(5) High temperature heatingand exposure test using blockspecimen (intensity distributionof extra thick cross-section)(6) Development of strength
evaluation method for structures
after SA (evaluation of the actual
RPV pedestal)

Construction of 3D-FEM analysis model
Integrity evaluation of the actual RPV pedestal

Figure 1 Long-term corrosion testing
Figure 2 RPV investigation workﬂow

IRID Annual Research Report 2014 27

Subsidized Projects

The deterioration of the RPV and PCV structural materials
at the Fukushima Daiichi NPS resulting from the hightemperatures, injection of seawater and falling debris that
occurred due to the severe nature of the Great East Japan
Earthquake is a cause for concern. In the period leading up to
the retrieval of fuel debris from the reactor core, a plan formaintaining the structural integrity of the PCV/RPV over thelong term is required.

・Assumed future state: Parameter(in 10 years, 15 years, 40 years)
・Building damage model
・D/W water level: approx. 2.9m
・S/C interior: Filled with water・Torus room: OP3680
・Vent pipes: Filled with water
・Vacuum break line: Filled withwater
・Additional equipment for
operations ﬂoor: Parameter (None,approx. 5100t, approx. 6100t)
・Injection of the water stoppage
material into individual rooms:
None・Attenuation constant: Parameter(1) Concrete 7%, steel 4%(regulatory guide)(2) Concrete 5%, steel 1%(for design during construction)
・Seismic waves: Current Ss

Main R&D
Achievements

Background

Plant/CaseDevelopment of Technology for Criticality Control in Fuel Debris
Retrieval

In the lead-up to fuel debris retrieval, in addition to preventing
re-criticality in each process, criticality safety controls toprevent excessive exposure to the general public or workers
on site are to be established. In conjunction to developing
criticality safety controls for the PCV submerging process,neutron absorption materials and fuel debris re-criticality
detection and monitoring technology will be developed.

Frequency

Aims

Total ﬁssion number

After an examination of criticality controls at the time of PCV submersion based on this evaluation, it was conﬁrmed that in criticality
prevention where sodium pentaborate is used as a soluble neutron absorption material, the required boron concentration wasapproximately 12,000 ppm. However, when the highly enriched 235U in fuel debris is set at approximately 4.0 wt.%, higher than the
average value in the fuel assemblies, and in light of fuel combustion efﬁciency and the effectiveness of neutron absorption material
such as gadolinia, the required concentration is expected to decrease.
Additionally, even in the event that criticality occurs, if the speed of water submergence is closely monitored, it can be expectedthat the total ﬁssion number can be suppressed to stay within the target range (Figure 1).

2. Technology for reactor re-criticality detection

The applicability of gas sampling ﬁssion product gamma radiation detector systems to actual reactors to detect re-criticality was
conﬁrmed. Speciﬁcally, a sub-criticality estimation algorithm that uses the differences in the yield of ﬁssion product nuclides
produced in nuclear ﬁssion was evaluated.
Furthermore, detectors were placed in the vicinity of fuel debris retrieval positions and development commenced of an in-core
criticality approach detection system to detect abnormalities before criticality is reached. In addition to extracting candidate
criticality approach detection methods and making evaluations (Table 1), a study of test methods for system veriﬁcation was
conducted.

3. Technology for criticality prevention

Development of non-soluble neutron absorbing material is currently in progress. This technology prevents criticality when it ispossible to access the fuel debris by supporting soluble neutron absorbing material through the direct application on debris.
Irradiation tests that compared materials to candidate materials chosen in the previous ﬁscal year were executed, and candidates
were selected based on elution characteristic radiation-resistance performance (Figure 2). In addition, an applicability evaluationof the fundamental physical properties of newly-devised candidate materials (underwater curing epoxy/Gd2O3) was carried out,
and the required functions of non-soluble neutron absorbing material and input pathway for its application, etc., were examined.
Soluble neutron absorption material will be widely used to prevent criticality between PCV submersion and the retrieval of fueldebris. In order to conﬁrm the effect the application of this material will have on the integrity of RPV/PCV, corrosion testing using aboric acid + pH adjuster and sodium pentaborate was performed. Results showed that signiﬁcant corrosion was not observed withthe concentration required for the suppression of criticality (Figure 3) and applicability for the prevention of criticality during PCV
submersion was conﬁrmed together with the results of a rough evaluation of insertion equipment.

Future Developments

In regard to criticality controls for the PCV submersion process examined during this year, in addition to achieving reﬁnement by reﬂecting
the latest ﬁndings (on debris location and so forth), further development of criticality control methods will take place while multiple methodsof fuel debris retrieval involving extraction in air and in water are considered. The feasibility of the in-core criticality approach detector system

28

Figure 1 Evaluation of total ﬁssion number on the assumption of a criticality accident

occurring at the time of PCV submersion

Using calculations that simulate behavior in the period between when criticality isreached and when it is detected and criticality suppression commences, the totalﬁssion number that affects exposure level is evaluated. Frequency distribution wassought with unknown conditions set as parameters. By limiting submersion speed(blue -> purple distribution), the total ﬁssion number can be kept under 1019 times, aﬁgure that keeps the exposure level under an allowable range.

Category

Feynman-αmethod- Passive method that does
not require an external
neutron source.

- Neutron multiplication
factor is estimated basedon intensity of the neutron
counting rate.

Time Interval
Distribution(TID)
method- Passive method that does
not require an external
neutron source.
- Neutron multiplication
factor is estimated basedon neutron ﬂuctuating
signal (noise) statistics.

△

Evaluating detection
efﬁciency of debris isa challenge.

○○

Applicable to shallower
sub-criticality with aneutron multiplication
factor of 0.7 or more.
Through preliminary
investigations, a prospect todistinguish between
spontaneous ﬁssion and ﬁssion
multiplication was obtained.
Detailed examination ofneutron multiplication factor
estimation is required for the
next step.

Table 1 Criticality approach detection technique candidates

Focusing on proven conventional methods, applicability based on conditions at thetime of fuel debris retrieval was examined and candidate methods were selected. In
future, selection will be made based on applicability to multiple retrieval methods,
including in-air and under water, and conﬁrmation of feasibility through veriﬁcation
testing.

Neutron absorber candidates
B4C/Sintered metal material

Solid

B or Gd-containing glass material
Gd2O3 particle

Liquid -> Solid
LiquidB concentration in test solution(mg/L)

The most recent ﬁndings regarding conditions inside the reactor were reﬂected in criticality scenarios for each process between
PCV submersion and the retrieval of fuel debris. Additionally, evaluation of criticality of fuel debris attached to CRD piping wasadded, and the risk of criticality for each part, such as fuel debris at the RPV lower head and the effect of submerging fuel remaining
in the reactor core, was evaluated.

- Active method that uses
external neutron sources.

- Neutron multiplication
factor is estimated basedon neutron ﬂuctuating
signal (noise) statistics.

Main Achievements

1. Evaluation of criticality

Neutron source
multiplication
methodApplicability

Cement/Gd2O3 particle

Water glass/Gd2O3 particle

Underwater curing epoxy/Gd2O3
Slurry/ Gd2O3

Gd or B-containing glass material

B4C/Sintered metal material

Gamma ray irradiation dose (kGy)
Figure 2 Non-soluble neutron absorption material dissolution test result (example)

Candidate materials that were chosen based on conﬁrmation of basic physicalproperties in the previous year were singled out by conﬁrming their characteristics
under irradiation.

Figure 3 Soluble neutron absorption material corrosion test result (example)

Image shows results of 500-hour test on sodium pentaborate of 10,000 ppm fornon-coated base material.
Stable passivation ﬁlm was formed and it was conﬁrmed that no signiﬁcant corrosion
was observed. Likewise, corrosion did not occur even when conditions werechanged, such as testing in the presence or absence of boric acid + pH adjuster orof rust inhibitors.

applicable to this process will be conﬁrmed through the performance of veriﬁcation tests. Furthermore, collection and selection of candidate
non-soluble neutron absorbing materials based on conﬁrmation of nuclear characteristics will take place and methods of material
application will be examined. The plan is to utilize these technologies in the process of selecting safe and efﬁcient methods of fuel retrieval.

IRID Annual Research Report 2014 29

Subsidized Projects

It is assumed that fuel debris is not currently in a critical state.
However, as it is posited that changes in the shape of fueldebris or in water level may occur during the process of fueldebris retrieval in future, the development of criticality
scenarios, criticality evaluation, monitoring technology, andtechnology to prevent criticality to prepare for such a situationis required.

Provisionaltarget value

Characteristics

Main R&D
Achievements

Background

Detection method

Development of Technology for Fuel Debris Characterization andTreatment④Metallic phase①Molten corium mass

In order to examine fuel debris retrieval methods or obtain/share thedata or information required for examination in other projects suchas that related to development of equipment or devices for retrieval
and technology for collection, transfer and storage of fuel debris,
testing and analysis using simulated debris will be conducted andestimates made of fuel debris characteristics.

Main Achievements

1. Analysis of the characteristics using simulated debris

Simulated corium (in this case, compacted Zr+SUS)
Concrete③Oxide corium/
concrete interface

Before heating

During heating

After heatinga. Understanding debris characteristics data

(U,Zr)O2, in tetragonal and monoclinic phases, and that includes ZrO2-stabilizer, was fabricated. Hardness, elasticity ratio and fracture
toughness as mechanical properties were measured and evaluated, and data on oxide-based fuel debris characteristics was
expanded. At the same time, Fe2 (Zr,U) was fabricated and evaluation of metallic characteristics commenced. In addition, boring tests
on multiple cold materials with differing characteristics were performed, clarifying the impact each property had on boring
performance.
U-Zr-O phase relations under conditions where hyper oxides in U-Zr-O-based oxides occur were investigated experimentally, anddata was expanded in order to predict the chemical form of U. Using two kinds of melting methods – arc melting and focusedradiation melting (Figure 1) – basic data related to reactions with concrete was obtained. Properties data on ﬁne debris generated
after immersion in cooling water for an extended period was acquired. In addition, solidiﬁed oxide melt debris were produced andby going through the melting process, the impact on debris characteristics could be clariﬁed. Calculations of physical property data
for systems that assume the presence of debris generated by fuel containing Gd or composite systems containing structural material(Fe) were performed.
International cooperative efforts related to molten core concrete interaction (MCCI) products enabled the use of MCCI test products
stored by the French Nuclear International Agency (CEA) (Figure 2) to obtain properties data such as chemical composition andhardness. Additionally, knowledge concerning the interfacial morphology between metals and ceramics was obtained by studying
metallic/ceramic solidiﬁed products using UO2 produced by the Kazakhstan National Nuclear Energy Center (NNC) (Figure 3).

⑤Oxide corium②Upper crust

Corium/concreteinterface(vertical direction)③

Corium/concreteinterface(horizontal direction)⑤②

Upper oxide crust①

Corium oxide
phase

④

Continuous
metallic phase
Silica-based
concrete

Figure 1 State of simulated MCCI products manufactured through light condensing

Figure 2 Sample selection of the MCCI test product from the French Nuclear

MCCI phenomenon was simulated on a small lab-scale in the following method: , asimulated corium disk (photos show compacted Zr+SUS) was placed on a piece ofconcrete and the simulated corium only was melted by heating only the disk using alight condensing and heating device. Using this method, data that focused on
characteristics of heterogeneous parts of MCCI that is thought to have occurred in
the Fukushima Daiichi NPS accident was acquired.

A few campaigns were selected from previously formed MCCI test products held by
the CEA, based on the criteria that conditions of the campaign should closely reﬂect
concrete components and the ratio of fuel and structural materials found after theaccident at the Fukushima Daiichi NPS. Five samples from these test products weretaken and subjected to chemical composition and hardness tests. As a result, datathat could be used for predicting the general characteristics of MCCI products in ashort period was obtained.

and heating

Stainless steel plate

Fuel debris

Tray

Stainless steel plate

International Agency (CEA)

Fuel debris

Trayb. Comparison with TMI-2 debris

In order to evaluate the characteristics of fuel debris from the Three Mile Island Unit 2 (TMI-2) accident and compare and examine it
against data obtained in the step ‘Understanding debris characteristics data’ above, sample processing and ceramography wereconducted (Figure 4) and Vickers hardness measured using three different samples of TMI-2 debris held by the Japan Atomic
Energy Agency (JAEA). In addition, in order to establish speciﬁc properties analysis techniques using actual debris samples, an alkali
fusion technique was applied and analyzed. Assuming that existing facilities can be effectively utilized for further actual debris
sample analysis in future, a transport study using TMI-2 debris as an example was conducted.

2. Development of debris treatment technology

While sharing information with other projects related to the collection, transfer and storage of fuel debris, a research and development
plan based on future technology development needs was formulated. Additionally, in regard to hydrous/drying property data required
for estimation of hydrogen generation by water radiolysis (problematic from a safety perspective), testing using porous ceramics wasundertaken and factors that affect drying properties set out. Further to this, simulated MOX fuel debris consisting of UO2, ZrO2 and PuO2
was manufactured, and redox behavior and properties changes in redox process experimentally were investigated.

Figure 3 Result of fabrication of metal and ceramic solidiﬁed products

produced at the Kazakhstan National Nuclear Energy Center(NNC)(1) Appearance of molten debris
after dropping

Area where stainlesssteel plate has melted
Metal which has
solidiﬁed after melting(2) Appearance after powdery surface
debris is cleared away

Future Developments

Figure 4 Cutting of TMI-2 debris sample and SEM/EPMA analysis of cutting

In regard to the mechanical properties of fuel debris particular to the Fukushima Daiichi NPS, in addition to evaluating simulated
debris, we will continue international cooperative efforts with France’ s CEA to evaluate the characteristics of MCCI products andcooperate with Kazakhstan’ s NNC to produce solidiﬁed products containing UO2, and compile data related to fuel debris
characteristics by the end of FY2015.
In the development of treatment technologies, based on a plan formulated in FY2014, evaluation of hydrous/drying properties
and so on required for the development of retrieval and storage technology will be performed.
30

In cooperation with Kazakhstan’s NNC, a UO2 + Zr + B4C mixture of theorder of 10kg was heated and melted before dropping into a pan placedon stainless steel of the same type used in reactor construction in orderto fabricate large simulated debris. As a result of the removal of powdery
surface debris, the metallic and ceramic solidiﬁed product could be
conﬁrmed (Figure 3 (2)), and knowledge of the metallic and ceramic
mixture state and interface form, and of product physical properties,
could be obtained.

dust

Cross section

M11-P10 (Crust) Ceramic

M11-P10 (Crust) U-Zr oxide

Photo at upper left shows pretreatment (cutting) of TMI-2 debris held by
the JAEA before being subjected to hardness testing and analysis before
treatment testing. The photo at lower left is an example of crust portion
from the TMI-2 debris. The results of SEM observations and EPMA
analysis of external appearance and cutting dust (right) showed thatdebris taken from the crust portion was primarily an oxide consisting of U
and Zr.

IRID Annual Research Report 2014 31

Subsidized Projects

In order to develop, design and manufacture equipment,
tools and containers for the retrieval and storage of fueldebris, the underlying data that is comprised of fueldebris mechanical properties, such as hardness, thermal
properties, chemical stability or moisture characteristics
must be fully understood.

Aims

Main R&D
Achievements

BackgroundReceiving debris
輸送容器による受け入れin the transfer cask

Development of Technology for Analysis of Debris Properties
Aims

The ﬁnal goal of this project is to analyze Fukushima debris anddetermine their physical properties in order to deploy and use theobtained information in other projects. For that purpose,
preparation is to be made in an examination of the overall analysis
workﬂow, the formulation of an analysis and measurement
technology development plan, and development of individual
analysis and measurement technologies.

Main Achievements

1. Establishment of development plan for fuel debris analysis and measurement technology

Solidsamples
固体試料(test
piece preparation)（試験片作製）

液体試料
Liquidsample(dissolution)（溶解）
Nuclidequantitative analysis
核種定量分析
(gammaspectrometry, etc.)（γ線スペクトロメトリ）

固体表面分析
Solidsurface analysis（(SEM/WDX,
SEM/WDXetc.)
等）
Measurement of
硬度等機械
mechanical properties,
such as hardness
的物性測定

Measurement of
熱量等熱的
thermal properties,such
as heat capacity
物性測定

quantitative analysis
Elemental
quantitative Nuclide
核種定量分析
元素定量分析
analysis(alpha spectrometry)（αスペクトロメトリ等）(ICP-A
ESetc.)等）(ICP-AES,(liquid scintillation, etc.)（液体シンチレーション等）

Figure 2 Debris dissolution test

Figure 1 Overall debris analysis workﬂow plan (outline)

The above ﬁgure shows an outline of the overall analysis workﬂow considered
to enable the efﬁcient implementation of debris analysis. Under this workﬂow, valuesof debris with unknown structural components are roughly analyzed by thenon-destructive method and then a work schedule of overall analyses is created.

The photo shows situation immediately after removal from electric furnace during
alkaline fusion. The test involved reacting simulated debris with sodium peroxide at
850℃ and then, after cooling, dissolving the debris using concentrated nitric acid.
No solids were observed in the resulting solution, and it was conﬁrmed thatcomplete dissolution had occurred.

Histogram

Phase separation results

Cross section 1

Analysis items and the overall workﬂow of debris analysis were examined in line with the needs of other projects and technical challenges
were identiﬁed. Items that include basic analysis such as observation of debris shape, element/nuclide quantitative analysis for composition
evaluation, and mechanical/thermal property measurements have been selected (Figure 1). The establishment of procedures for dissolving
debris, the inﬂuence of ﬂux components in quantitative analysis, and the actualization of test measurement methods have been identiﬁed
as the main technical challenges. These potential challenges were collated and future development plans have been formulated.

2. Development of analysis and measurement technology

Cross section 2

a. Development of dissolution technique using fusion method, etc.

A fusion method used as a pre-analysis processing method is being examined for the purpose of elemental analysis of fuel debris samples. In
addition to Zr cladding and reactor core internal structural materials being mixed together, it is thought that MCCI products, debris reacted athigh temperatures with concrete, exist and are extremely insoluble. In response to this, it was conﬁrmed that by using sodium peroxide,simulated debris could be completely dissolved in nitric acid through the “alkaline fusion method,” known as a method for the dissolution of low

Cross section 3

solubility samples (Figure 2).

b. Study on chemical state analysis method

To investigate the composition of solid debris (including chemical state), simulated solid products from assumed components were produced
and a quantitative analysis using SEM/WDX solid state analysis methods was studied. The quantitative analysis method involved themeasurement of composition area ratio using SEM/WDX reﬂection electron imaging and using spot analysis and so on as a method of
identifying chemical states (Figure 3). In testing of simulated U+UO2 solidiﬁed product, almost identical results were seen in quantitative results

yielded from powder X-ray diffraction (Rietveld analysis).

c. Maintenance of debris analysis equipment

Void

Identiﬁcation by spot analysis, etc.

Figure 3 Results of phase analysis using backscattered electron images and gradation distribution taken by SEM/WDX

The proportion of each phase of the simulated solidiﬁed product (U+UO2) was measured using gradation distribution of backscatteredelectron images taken by SEM/WDX. The chemical form of the phase was identiﬁed through SEM/WDX spot analysis.

High-dose rate debris is required to be handled within a cell. For the purpose of analysis of debris, analysis equipment (SEM/EDX/WDX) andthe cell to contain them were conﬁgured, and cell speciﬁcations and the required modiﬁcation of equipment were examined. From a safety andoperability standpoint, all of the equipment is required to be placed within the cell, which was required to be 2.5 m X 2.8 m X 2.8 m in size as aresult. In regard to the modiﬁcation of the equipment, locations that required modiﬁcation and steps involved were clariﬁed in terms ofIron cell

operability, reduction of the impact of debris radiation, and maintenance (Figure 4).

Contains 244Cm, 240Pu

The fuel debris produced at the Fukushima Daiichi NPS is of a type that contains internal structural components. If the appropriate analysis
and classiﬁcation can be employed at the time the debris is retrieved, the subsequent storage process and selection of storage location canbe streamlined. Classiﬁcation of debris was therefore examined in terms of ‘nuclear material management’, ‘streamlining of storage’ and‘safety at time of disposal’. In addition, a proposed sorting method for the measurement of nuclear material (Figure 5) was examined, andthe sorting performance (lower detection limit) of devices using the active neutron method was analytically studied.

The technically-challenging fuel debris dissolution procedures will be established and technology to deal with ﬂux components inquantitative analysis will be developed. These outputs will be reﬂected in the development of the overall workﬂow for fuel debris analysis.
Working closely with other projects, in addition to further clarifying required physical property values and content, test and quantitativeevaluation methods that can produce appropriate property values will be explored. Furthermore, in preparation for early analysis,containers for transportation of fuel debris for analysis will also be studied.
32

Minimal/No containing
244
Cm, 240Pu

Neutron

3. Study on properties analysis for the proper treatment and disposal of debris

Future Developments

Passive neutron method

Determined that there isa high possibility ofcontaining nuclear fuelMaterial accountancy

Figure 4 Cell composition and equipment modiﬁcation points

The required modiﬁcations are as below:- Automation of objective aperture (A)- Stroke extension of sample introduction part (B)- Addition of EDX drive unit (C)- Installation of WDX shield (D)- Installation of control device (E)Considered range

Active neutron method(ﬁssile nuclide detection method)

Determined
as fuel debris

Fission neutrons

Below detection limit
Determined that there is a low
possibility of containing nuclear fuelFigure 5 Proposed sorting method related to the presence or absence of nuclear material

When a signiﬁcant amount of neutrons are measured using the passive neutron method, material isclassiﬁed as containing nuclear fuel. When there is less than a meaningful quantity, it is dealt withusing the active neutron method. When a signiﬁcant amount of ﬁssion neutrons are detected,
material is classiﬁed as containing nuclear fuel. When they are not detected, it is assumed thematerial has a low possibility of containing nuclear fuel.

IRID Annual Research Report 2014 33

Subsidized Projects

Determining the properties of debris is crucial to understanding
the situation inside the reactor core, stability management, andensuring that retrieval and storage operations proceed safely andsteadily. In order to provide useful information without delays,
analysis-related technologies, including pre-analysis processing
and wastewater treatment technologies, must be established inadvance so as to handle a wide variety of material properties.

基礎特性
Basic
properties(observation of shape, measurement of mass, etc.)（形状観察、
質量測定等）
Main R&D
Achievements

Background

非破壊測定
Non-destructive
measurement(X-rayCT, etc.)（X
線ＣＴ等）

Main R&D Achievements
Key Issue 3: R&D for Treatment and Disposal of Radioactive Waste

Sr radioactivity concentration (Bq/g)

Fallen tree
Standing tree around Unit 3

Inside building (Unit 1)
Inside building (Unit 3)

Lower detection limit

Main Achievements

Figure 2 Distribution of radioactivity in standing trees (branches and leaves) at theplants (felled and standing trees) taken inside and outside reactor
buildings

Fukushima Daiichi NPS

Cement kneading solidiﬁcation

Technology No.

Process overview This is a technique to solidify and encapsulate waste on the basis ofa property of cement material to be hardened by hydration reaction.
Applicablewaste
material

Spent ion exchange resin, sludge, incinerated ash, etc.
Cement silo
Water added toconcentrated
liquid waste

2. Study on long-term storage method of accident-generated waste

Concentrated liquid
waste weighing tank

In regard to the long-term storage of cesium absorption vessels used to separate and recover cesium from contaminated water, a simulation test
using a 1/4 scale mock-up was used to determine conditions inside the vessels (Figure 3). Observations showed that insoluble components
containing chlorine are transferred to the part that generates heat by absorbing Cs. Based on this knowledge, estimations were made bycalculating temporal changes inside the vessels. Additionally, experiments also sought to determine absorption vessels material corrosion
conditions.
Slurry dehydration/drying tests were conducted in order to stabilize iron hydroxide and carbonate slurry generated by the Advanced Liquid
Processing System (ALPS). In addition, the amount of hydrogen gas expelled by stable products due to exposure to radiation was also evaluated.

3. Study on treatment of accident-generated waste

In-drum mixer
The above diagram represents in-drum mixing method: the placingof weighed concentrated liquid waste together with cement andkneading water, etc., in a waste canister and kneading conducted togenerate waste materials.
In addition to the in-drum mixing method, out-drum mixing andvacuum injection methods are also available, See Chart 1-1(Appendix) for more details.

Based on the results of surveys on existing disposal concepts and safety assessment techniques (scenarios, models, and parameters),
basic concepts and characteristics were arranged. In addition, safety evaluation method was developed provisionally consideringcharacteristics of accident-generated waste assuming that existing disposal concepts would be applied to accident-generated waste.
Furthermore, analysis cases for assumed scenarios were created, and in addition to estimating the safety of each waste product, sensitivity
analysis to determine parameter variation was implemented and information that contributes to the investigation of disposal
classiﬁcation and relates to important nuclides that affect disposal safety were extracted.

5. Study on prerequisites for R&D

In addition to classifying accident-generated waste characteristics and contamination history, a series of examples (waste management
stream) looking at how storage, processing and disposal processes are to be handled was drawn up (Figure 5). Information items (waste
amount, radioactive concentration, storage conditions, etc.) were also organized together with this action. Related water analysis result
data was published in the FY2013 edition of the Database on Analyses of Contaminated Water (JAEA-Data/Code 2014-016）.
Additionally, the relationship between main work items and information items in the development of processing and disposal
technologies was organized as a step towards the development of tools designed for information management.
34

Domestic licenses (normal Portland cement, high early strengthPortland cement, ultra high early strength Portland cement,
Special remarks moderate heat Portland cement, sulphate resisting Portland cement,
blast furnace slag cement (A-type, B-type, C-type)
Admixture sometimes added during kneading (See Chart 1-2 in Appendix.)

apparatus

Incinerated ash: Substances that generate hydrogen (aluminium, etc.)
are required to be removed beforehand.
Waste liquid containing borate: Adding Ca2+, borate concentrationmust be kept below 5wt%
Waste liquid containing sulphuric acid: SO42- concentration must be
reduced by adding Ca(OH)2.
8% 3CaO/Al2O3 cement: Use SO42- concentration of 150-1,000ppm
5% 3CaO/Al2O3 cement: Use SO42- concentration of greater than 1,000ppm(1) Processing conditions
Temperature: Room temperature
Pressure: Atmospheric pressure(2) Processing conditions
Filling rate: 2-66 wt%
Domestic: International: See Chart 1-3 in Appendix(3) Waste characteristics
Compressive strength: 3.2-70 MPa
Density of solidiﬁed waste: 1.5-2.3 ton/m3
Radiation resistance: No change to mechanical properties at irradiation of around 108Gy.
G(H2): 0.05-0.13
Heat resistance: Resistant to temperatures of below 200℃ during long-term storage
See Chart 1-4 in Appendix for heat reaction.
Nuclide leaching rate: See Chart 1-5 in Appendix.
Partition coefﬁcient: 21 cm3/g(Cs), 11 cm3/g(Sr)(1) Technology maturity (Actual, etc.)

Domestic: Many results at nuclear power stations (See Chart 1-1 in Appendix.)
International: Many results at nuclear power stations or research facilities (See Chart 1-3 in Appendix.)(2) Operational safety

High degree of safety as operations carried out at room temperature and dangerous material not used.

(3) Serviceability

Serviceability is high as operations are not required for heating etc. and device is simple.

(4) Generated amount of secondary waste(5) Applicability to waste matter generated by accident (organic matter, etc)
As CN progresses from decontamination equipment sludge, it isnecessary to perform pre-treatment and the decomposition of CN.

Figure 4 Example of technical catalog

Basic testing and the creation of a technical catalog were conducted in the initial selection of candidate treatment and conditioning
techniques. Targeting the slurry generated by the ALPS and waste absorbents, solidiﬁcation tests were performed by using various typesof solid-type materials and the curing process and waste characteristics were explored. Additionally, resin-based absorbent volume
reduction testing and radioactive material gas phase migration experiments using high-temperature processing were conducted. In
addition to making a technical catalog that includes a waste treatment, conditioning technique technical outline, applicable waste,
process workﬂow, processing capacity, and technology readiness (Figure 4), organization of technical requirements commenced in theprocess of narrowing down initial candidate technologies.

4. Study on disposal of accident-generated waste

Cement weighing
equipment

Process ﬂow
and equipment
overview

Figure 3 Simulated tests to determine conditions inside cesium absorption

Pretreatment

Radioactivity level readings are being sought through the analysis of a variety of accident-generated waste. Contaminated water, plants
and rubble (Figure 1) have been analyzed and in regard to the analysis of standing trees, 3H (tritium) and 14C have been detected alongwith 137Cs and 90Sr, revealing the distribution of radioactivity at the power station site. At the same time, development of analyticaltechniques intended for nuclides difﬁcult to analyze at low concentrations, collation of data of the characteristics of secondary waste
stemming from treatment of contaminated water (zeolite for Cs absorption, etc.), and evaluation of radioactivity levels (inventory) for
waste currently difﬁcult to analyze were executed.

Figure 1 Relationship in concentration of 90Sr and 137Cs detected in rubble andProcessing ability and waste characteristics

1. Characterization of accident-generated waste

Cs radioactivity concentration (Bq/g)

137

Applicability evaluation

Through characterization by inventory assessment based on waste productanalysis and analytical techniques, the study of long-term storage strategies forstable management up to processing and disposal, investigations and basic tests
related to technologies for processing and packaging waste, and research into andorganization of existing disposal concepts and safety assessment techniques,
technologies for the safe treatment of accident-generated waste and for its
disposal will be developed.

Around Unit 4

Subsidized Projects

Waste generated through the accident at the
Fukushima Daiichi NPS contains radionuclides derived
from fuel, may contain seawater components, is highly
radioactive, wide-ranging in contamination level, andhigh in volume, making it different from radioactive
waste generated at typical nuclear power plants.

Aims

Around Unit 3

Main R&D
Achievements

Background

Dose rate

Around Unit 1

90

Development of Technology for Treatment and Disposal of
Accident-generated Waste

Waste generation
Contamination classiﬁcation: Ⅵ, Ⅶ

Rubble inside building
Rubble outside

Storage

Preprocessing/Processing/Reprocessing

Outdoor accumulation

Covering soil/
Temporarystorage facilities/
CanisterMaterial
Low
Dose rate

Storage

Waste conditioning

MediumDisposal/Recycling
Reuse

Non-recyclable
Grinding

Storage

Flexible container storage

Waste conﬁrmation(1-30mSv/h)

Solid waste
storage facility

Storage

Recyclable
Reprocessing
SortingHigh

Grinding

Basic ﬂow
High (Equivalent to L1)
Grinding

Container storage

Encapsulation

Storage

Solidiﬁcation

Waste conﬁrmation

Encapsulation

Storage

Solidiﬁcation

Waste conﬁrmation

Trench disposal (L3)

Pit disposal (L2)

Sub-surface disposal (L1)

(Dose category value) Upper bounds of radioactive concentration for burial of low-level radioactive solid waste based on Nuclear Reactor Regulation Law

Dose rate (TRU equivalent): >10 (Bq/t) α concentration
Dose rate (High L1 equivalent): >1015 (Bq/t) 60Co, >1014 (Bq/t) 137Cs
Dose rate (High): <1015 (Bq/t) 60Co, <1014 (Bq/t) 137Cs
Dose rate (Medium): <1010 (Bq/t) 60Co, <108 (Bq/t) 137Cs
Dose rate (Low): Requires study
11

Figure 5 Example of series of handlingprocesses for storage, processing
and disposal of waste

Future Developments

In addition to continuing to analyze accident-generated waste characteristics, inventory assessment using analytical techniques will be studied. In
regard to the study of long-term storage strategies, engineering tests related to the stabilization of slurry generated by the ALPS will be conducted
to evaluate the entire system. Data related to the processing and packaging of waste will continued to be obtained in the study of processing
techniques. Meanwhile in the study of waste disposal, existing evaluation methods will be reviewed and classiﬁcation evaluation and indication ofchallenges regarding accident-generated waste will be undertaken. Finally, in the examination of prerequisites for R&D, organization of waste
management stream candidates and information on accident-generated waste will be ongoing.
IRID Annual Research Report 2014 35

Main Study Results: In-house R&D Activities

Others ‒ Data Sheet

Analysis of Contaminated Objects Sampled from Reactor Buildings

Main Research Results Announced/Published in FY 2014

Background

In order to develop remote decontamination
devices, obtaining detailed data on contamination
conditions inside the Unit 1-3 reactor buildings isessential. An understanding of the contamination
situation and acquisition of basic data throughanalysis of contaminated samples from eachreactor building is therefore required.

Aims

Contaminated samples taken from high does rate areas in reactor
buildings (Units 1-2) will be analyzed to understand in detail the type ofradioactive materials present and their distribution (particularly theirpermeation into concrete walls exposed to a steam environment). Steps
will be taken to expand basic data that will contribute to the developmentof remotely controlled decontamination equipment and the formulationof a decontamination plan.

Pub lis her

Date

Details

2 Japan Society of Maintenology

July 24, 2014

Remote decontamination technology demonstrated at the Fukushima Daiichi NPS(blast/suction decontamination using MEISTeR, core sampling)

3 Atomic Energy Society of Japan

August 22, 2014

Development of technology for criticality control in fuel debris retrieval

4 Atomic Energy Society of Japan

August 22, 2014

Evaluation of long-term structural integrity of fuel assemblies removed from the spent
fuel pool

Main Achievements

5 Atomic Energy Society of Japan

August 22, 2014

Development of technology for remotely operated decontamination inside reactor buildings(Results of dosage survey taken at southern end of Unit 1 ﬁrst ﬂoor and operation ﬂoor of Unit 2)

6 Atomic Energy Society of Japan

August 22, 2014

Development of technology for remotely operated decontamination inside reactor
buildings (High pressure water jet)In a sample taken from the central part of the well plug, the surface of the curing sheet covering the ﬂoor showed ﬁxed
contamination of 137Cs radioactive material, etc., at 2.6×106Bq/cm2, and ﬁxed contamination of 1.4×102Bq/cm2 was also found on
the epoxy coating surface. While there were some signs of degradation of the thin epoxy coating surface, contaminants were
found to have not permeated into the epoxy coating and concrete, and it was conﬁrmed that the epoxy coating suppressed thepenetration of the contaminants into the concrete (Figure 1). Fixed contamination of 1.0×104Bq/cm2 was found on thethinly-layered surface coating on the ceiling of the upper part of the well plug (deck plate), and by organizing cut test pieces andperforming simple decontamination tests, the effectiveness of decontamination techniques were conﬁrmed.

7 Atomic Energy Society of Japan

August 22, 2014

Localized corrosion behavior of stainless steel in zeolite containing diluted seawater
under gamma ray irradiation

8 Atomic Energy Society of Japan

September 8, 2014

Identifying conditions inside reactor through application of severe accident analysis
code (MAAP5 revised version)

9 Atomic Energy Society of Japan

September 8, 2014

Identifying conditions inside reactor through application of severe accident analysis
code (analysis code for casting process simulation)

10 Atomic Energy Society of Japan

September 8, 2014

Identifying conditions inside reactor through application of severe accident analysis
code (SAMPSON code)

11 Atomic Energy Society of Japan

September 8, 2014

Identifying conditions inside reactor through application of severe accident analysis
code (sophisticated MAAP)

12 Atomic Energy Society of Japan

September 8, 2014

Development of technology for remotely operated decontamination inside reactor
buildings

13 WRFPM2014

September 14, 2014 fuel pool

14 WRFPM2014

September 14, 2014 Development of technology for criticality control in fuel debris retrieval

2. Analysis of contaminated samples taken at the southern side of the ﬁrst ﬂoor of Unit 1

Future Developments

Basic data obtained through this research on the contamination situation at each Unit will be used in the selection of
decontamination technology for decontamination work at the south side on the ﬁrst ﬂoor at Unit 1 and the operation ﬂoor at Unit
2, and will be used for such things as the improvement of remotely controlled decontamination equipment in other projects andwill be reﬂected in decontamination plans.

Sample appearance
Surface

Peeling between
coating ﬁlms

Peeling between coating ﬁlms

15 Radiation Research Symposium

October 9, 2014

Evaluation of radiation resistance of solidiﬁed inorganic specimen simulating sludge
generated from water treatment in the Fukushima Daiichi NPS

16 NuMAT2014

October 27, 2014

Identifying conditions inside reactor through application of severe accident analysis
code

17 ICMST-Kobe2014

November 2, 2014

Development of technology for remotely operated decontamination inside reactor
buildings

18 ICMST-Kobe2014

November 2, 2014

Evaluation of radioactivity inventory of 137Cs in waste generated from water treatment
system at the Fukushima Daiichi NPS

The 9th Takasaki Advanced

19 Institute Meeting to Report Results November 20, 2014

Development of technology for fuel debris characterization and treatment of fueldebris/Development of technology for treatment and disposal of accident-generated waste

20 Japan Society of Corrosion Engineering November 26, 2014

Development of technology for evaluating integrity of the RPV/PCV

5th Applied Laser Technology

Japan Atomic Energy Agency

Longitudinal section of sample

21 Report Meeting

November 28, 2014

Development of technology for treatment and disposal of accident-generated waste

23

Workshop on Radiation Exposure Management
at the TEPCO Fukushima Daiichi NPS

December 11, 2014

Development of technology for remotely operated decontamination inside reactor
buildings

24

University of Tokyo The First Human Resource
Development Seminar

December 11, 2014

Radioactivity analysis for the treatment and disposal of accident-generated waste

February 12, 2015

Sector of Fukushima Research and Development Meeting to Report Results for 2014 Results of evaluation of fuel debris characteristics,
study on method of processing damaged fuel, identiﬁcation of conditions inside reactors, and treatment/disposal of waste

Figure 1 Imaging plate measurement results of contamination

26 IAEA International Specialists Meeting February 19, 2015

Further investigation of the fabrication and production phases of simulated fuel debris
by melting and solidiﬁcation of fuel and B4C control rod material.

Through measurements of sample surface and longitudinal
sections, it was found that while ﬁxed contamination was present
on the surface of the epoxy coating, contamination had notpermeated into the concrete. (Removed paint and contaminants
found were assumed to have come from when the test sample
was taken).

27 National Laboratory (INL)

Workshop with the Idaho

March 2, 2015

Development of technology for collection, transfer and storage of fuel debris

28 Atomic Energy Society of Japan

March 20, 2015

Development of technology for remotely operated decontamination inside reactor
buildings

29 Atomic Energy Society of Japan

March 20, 2015

Treatment of contaminated water at the Fukushima Daiichi NPS and the current state ofwaste products generated from the water

25 Japan Atomic Energy Agency

Longitudinal section of sample (illustration)

sample (Unit 2 operation ﬂoor: Floor of upper part of well)

Appearance of longitudinal section of sample

Imaging plate

JAEA efforts towards technology for the decommissioning of the Fukushima Daiichi
NPS

2014 Specialists’ Meeting onRadioactive Waste Management

Coatingﬁlm

ConcreteNovember 27, 201422

Contamination

Partial peeling

Evaluation of long-term structural integrity of fuel assemblies removed from the spent

IRID Annual Research Report 2014 37

Data Sheet

In a sample taken from the area in between the X-6 penetration and piping, ﬁxed contamination on the surface of the epoxy
coating covering the ﬂoor was found to be 1.4×102Bq/cm2, but was not found to have permeated and caused internal
contamination. Samples taken from the vicinity of water stains at the AC pipe base showed ﬁxed contamination on the surface ofthe epoxy coating covering the ﬂoor at 2.7×102Bq/cm2. While permeation into the coating was not found, in places where cracks
had formed in the coating, penetrating contamination in the lower concrete layer was observed to be around 1.7×102Bq/cm2. It
was assumed to be caused by moisture included 137Cs that had passed through the cracks in the epoxy coating and had
permeated into the concrete below.

OthersJuly 4, 2014

Main Study
In-house R&D Activities
Results

1 Atomic Energy Society of Japan

Evaluation of long-term structural integrity of the fuel assemblies removed from thespent fuel pool

1. Analysis of contaminated samples taken from the operation ﬂoor of Unit 2

36

No .

Main R&D Installations/Equipment

Main Research Results Announced/Published in FY 2014

(Over 1 million yen)

No.

Date

DetailsNo .

Pr o j ect Na m e

1

Evaluation of long-term structural integrity of the fuel assemblies removed from the spent fuel pool

Ultrapure water making apparatus

Item Name

2

Evaluation of long-term structural integrity of the fuel assemblies removed from the spent fuel pool

Multi water quality meter3

Evaluation of long-term structural integrity of the fuel assemblies removed from the spent fuel pool

Automatic tritator

March 20, 2015

31 Atomic Energy Society of Japan

March 20, 2015

Preparation and characterization of simulated MCCI products (Characterization ofinterface between concrete and molten core)

32 Atomic Energy Society of Japan

March 20, 2015

Study on dissolution technique for damaged fuel in the Fukushima Daiichi NPS(Dissolution test of fuel debris sample from TMI-2)

4

Evaluation of long-term structural integrity of the fuel assemblies removed from the spent fuel poolSpectrophotometer

33 Atomic Energy Society of Japan

March 20, 2015

Impact evaluation of seawater on corrosion of materials used in reprocessing process

5

Study of methods to process damaged fuel removed from the spent fuel poolPressure reduction type immersion corrosion test equipment

34 Atomic Energy Society of Japan

March 20, 2015

Development of technology for remotely operated decontamination inside reactor
buildings (Analysis of concrete core sample from south side of the ﬁrst ﬂoor in Unit 1)

6

Development of technology for investigation inside the PCVInvestigation equipment for pedestal interior

35 Atomic Energy Society of Japan

March 20, 2015

7

Development of technology for investigation inside the PCV

X-6 penetration hole boring equipment

8

Penetration cable removal equipment

March 20, 2015

Development of inventory evaluation methods for the radioactive wastes of
the Fukushima Daiichi NPS

Development of technology for investigation inside the PCV

36 Atomic Energy Society of Japan

9

Development of technology for investigation inside the PCV

Sediment removal device

37 Atomic Energy Society of Japan

March 20, 2015

Study on long-term storage method of spent zeolite absorption vessel

10

Development of technology for investigation inside the PCVAlternate shielding body

38 Atomic Energy Society of Japan

March 20, 2015

Current status of secondary waste stemming from contaminated water treatment andefforts towards processing and disposal

11

Development of technology for investigation inside the PCV

Pre-conﬁrmation Equipment for pedestal interior

12

Development of technology for investigation inside the PCV

Mock-up of PCV internals

13

Development of technology for investigation inside the PCVShield block removal equipment

14

Development of technology for investigation inside the PCV

Shape-changing device for investigating pedestal exterior

15

Development of technology for investigation inside the PCV

Scattering prevention device for use in equipment to investigate pedestal exterior

16

Development of technology for investigation inside the PCV

Equipment for measurement of fuel debris shape

17

Development of technology for detection of fuel debris in the reactor

Small-scale muon tracking system using scattering method18

Development of technology for detection of fuel debris in the reactor

Muon tracking system using scattering method for 1F

19

Development of technology for retrieval of fuel debris and reactor internals

Equipment for evaluating sheet to prevent spread of contamination

20

Development of technology for retrieval of fuel debris and reactor internals

Remotely operated arm (artiﬁcial muscle robot)

21

Development of technology for fuel debris characterization and treatment

Ceramography image processing equipment

22

Development of technology for fuel debris characterization and treatment

Carbon coater

23

Development of technology for fuel debris characterization and treatment

Fuel debris sonic speed measuring device

24

Development of technology for fuel debris characterization and treatment

Fuel debris pressure test equipment

25

Development of technology for fuel debris characterization and treatment

High capacity thermobalance and simultaneous heat analysis equipment

26

Development of technology for fuel debris characterization and treatment

Crystal piezoelectric four-component dynamometer

27

Development of technology for fuel debris characterization and treatment

SEM elemental analysis system

28

Development of technology for fuel debris characterization and treatment

Hydraulic auto mounting machine

29

Development of technology for fuel debris characterization and treatment

Inverted metallurgical microscope

30

Development of technology for fuel debris characterization and treatment

Vacuum arc melting furnace

31

Development of technology for analysis of debris properties

Ball mill

32

Development of technology for analysis of debris properties

Small thermostatic dryer

33

Development of technology for analysis of debris properties

Measurement control system for waste sorting measurement technology

34

Development of technology for treatment and disposal of accident-generated waste

Small-scale ﬁlter press

35

Development of technology for treatment and disposal of accident-generated waste

Full-scale ﬁlter press

36

Development of technology for treatment and disposal of accident-generated waste

Slurry conditioning tank equipment

37

Development of technology for treatment and disposal of accident-generated waste

Particle size analyzer

38

Development of technology for treatment and disposal of accident-generated waste

Atomic absorption spectrophotometer

39

Development of technology for treatment and disposal of accident-generated waste

Water puriﬁcation system

Radiochemical analysis of trees collected from the Fukushima Daiichi NPS

IRID Annual Research Report 2014 39

Data Sheet

30 Atomic Energy Society of Japan

Preparation and characterization of simulated MCCI products (Phases andmicrohardness of molten solidiﬁed samples prepared by arc melting)

Others38

P u bl isher

Joint Research/Contract ResearchNo.

P r o j e ct N ame

Type

Title

Partner

PeriodRPV internal investigation. Element
technology feasibility study (transporttechnology)

Tokyo Institute ofTechnology

Dec. 2014 – Feb. 2015

Contract
research

RPV internal investigation. Element
technology feasibility study (transporttechnology)

Meijo University

Dec. 2014 – Mar. 2015

3 Development of Technology forContract
research

Measurement of fuel debris inside Unit 1
reactor at the Fukushima Daiichi NPS using
muon transmission method

High Energy
Accelerator
Research
Organization (KEK)

Jun. 2014 – Mar. 2015

4 Development of Technology forContract
research

Evaluation of characteristics of debris which
has melted and solidiﬁed

Central Research

Nov. 2014 – Feb. 2015

5 Development of Technology forContract
research

Study related to technology for evaluating
waste product inventory

Central ResearchInstitute of ElectricPower Industry

Sep. 2014 – Jan. 2015

6 Development of Technology forContract
research

Study on evaporation of remaining water
through zeolite

Central ResearchInstitute of ElectricPower Industry

Oct. 2014 – Jan. 2015

7 Development of Technology for

Joint
Research

Study on precision technology for analysis ofactinide analysis reagents using capillaryelectrophoresis

Saitama University

Aug. 2014 – Jan. 2015

8 Development of Technology for

Joint
Research

Zeolite thermal conductivity measurement
and analysis

Fukushima National
College ofTechnology

Sep. 2014 – Jan. 2015

9 Development of Technology forContract
research

Study on improvement of remote operabilityof manipulator

Kobe University

Dec. 2014 – Aug. 201510 Development of Technology forContract

Study on 3D recognition technology
surrounding robots

University of Tsukuba

Dec. 2014 – Aug. 2015

Study on camera calibration method for
simulated bird’s eye view image generation
system mounted on remotely operated
robot

University of Tokyo

Nov. 2014 – Aug. 2015

1 Development of Technology forContract
research

2 Development of Technology for

Investigation Inside the RPVInvestigation Inside the RPV

Detection of Fuel Debris in theReactorFuel Debris Characterization andTreatment
Treatment and Disposal of
Accident-generated Waste
Treatment and Disposal of
Accident-generated Waste
Treatment and Disposal of
Accident-generated Waste
Treatment and Disposal of
Accident-generated Waste

Remotely Operated Decontamination
Inside Reactor Buildings

Remotely Operated Decontamination research
Inside Reactor Buildings

11

Contract
Development of Technology for
Remotely Operated Decontamination research
Inside Reactor Buildings

Institute of ElectricPower Industry

5F, 3 Toyokaiji Building, 2-23-1 Nishi-Shimbashi,
Minato-ku, Tokyo 105-0003, Japan
Phone:+81-3-6435-8601
URL http://irid.or.jp40