R&D on Fast Reactor Cycles and Role of Monju and Joyo

February 17, 2016

Director General, Advanced Fast Reactor Cycle System Research and Development Center,

Sector of Fast Reactor Research and Development, Japan Atomic Energy Agency (JAEA)

Hideki Kamide

International Symposium on “Present Status and Future Perspective for Reducing Radioactive Wastes~ Challenge for the Relief in the Next-generation ~,” Feb. 17, 2016, Tokyo

Today’s significance to hold fast reactor cycle technologies and the role of "Monju" in R&Ds

Future planning of R&Ds Reflection of R&D results into the demonstration and

commercialization R&D on “Monju” Safety enhancement of FBR / FR Development of demonstrated technologies of FBR / FR Development status of separation and transmutation technology that

utilizes FBR / FR Situation for the irradiation test resumption of “Joyo”

Direction of the fast reactor cycle R&Ds should be headed

1 Contents

Today’s significance to hold fast reactor cycle technology

Drastic strengthening of efforts to resolve the SF problem Efforts on final disposal of high-level radioactive wastes Expansion of the storage capacity of the SF Technology development for reducing the

volume and radiotoxicity of radioactive wastes

Promotion of the nuclear fuel cycle policy Promoting such as reprocessing and plutonium-use in LWR Effective use of recovered Pu, etc. (R&D on

Nuclear fuel cycle, Fast reactors, etc.) Monju: aggregation of research results as an

international research center Flexibility of medium- and long-term response (Uncertainty)

Spent fuel (SF) Problem

Japan stores about 17,000 ton of SF.

SF and radioactive wastes are generated continuously by the nuclear power generation and the decommissioning.

Responding to sustainability and uncertainty

The basic policy is to promote the nuclear fuel cycle.

To ensure a wide range of choices for uncertainties in future is also important from the energy security in Japan.

Holdings of fast reactor cycle technology is important. Contribution to Japan's energy security (effective utilization of uranium

resources) Reduction of high-level radioactive waste (HLW) generated (environmental

load reduction) Step-by-step R&D using "Joyo", "Monju", etc. is essential.

2

Efforts to make steadily progress without Postpone

FR Reprocessing (Vitrified waste)

LWR Reprocessing (Vitrified waste)

3

Source：JAEA made by using the Uranium 2009: Resources, Production and Demand.

-100 years

【Effective use of uranium resources】

0

0.2

0.4

0.6

0.8

1

LWR SF

volume decreases to one-seventh

Volu

me

of H

LW p

er u

nit p

ower

ge

nera

tion

(rel

ativ

e va

lue)

【Reduction of environmental impact】

Efficient use of Resources The use of Pu enables Energy Independence without depending on overseas

Uranium resources Deployment of FBR allows the U use of more than a thousand years.

U resource reserves are about 100 years depending on nuclear power generation. A steady R&D of FR Cycle is required. The development takes a long time.

Eco-friendly Reduce the amount of HLW by reprocessing SF and vitrifying HLW. Higher reduction efficiency by Shifting to the FR Cycle

High thermal efficiency of Power Generation Lower heat generation of vitrified wastes by removing Minor Actinides.

Features of the Fast Reactor (FR) Cycle

＜Reduction of generated wastes＞

In the case of LWR use

In the case of FBR use

Several thousand years

Number of years until used up uranium resources

The role of "Monju" in R&D

Actual scale test is essential for Fuel Assembly. Full-scale irradiation test in “Monju”

The world's first to obtain characteristics of Core with Am* in the entire region.

Monju

Safety Enhancement

Safety technology system of FBR

SDC and SDG of Gen-IV reactors

Accident management measures in Severe Acci.

Comprehensive safety evaluation by PSA

Prevention of re-criticality in CDA

Diversification of stable cooling measures of a damaged reactor core fuel

Implementation of the SA counter measures

R&D outside of “Monju”

R&D in "Monju"

“Monju” is an actual plant to provide R&D field for Safety technology system of the entire FBR

Pu-3

CPF AGF

Fuel development and irradiation tests

Reactor characteristics/ rector system

Fuel Fabrication

Reprocessing

Am*: typical long-lived radionuclides in the waste

Compilation of the R&D results of the FBR technology

Joyo

"Monju“ is a large-scale FBR power plant and aggregation of our country own technology. Know-how obtained by operating our own plant (design, manufacture, construction) is key.

FMF PIE tests

4

Holdings of our own FBR technology (design, construction, operation, decommissioning)

Reducing the waste volume and radio-toxicity

CPF

Evaluation of waste reduction

by FR cycle

Fast Reactor (FR)

Development of test field for R&D “Monju” Early release from the NRA’s order on safety treatment “Joyo” Apply to NRA for permission to restart Pu-3 fuel production facility Commercialization of MOX fuel production facility Hot labo., PIE facilities Response to the new regulatory regulation Test facilities related to Na, etc. Consolidation of sodium test facility

The “third period medium- and long-term planning" of JAEA reflecting “Strategic Energy Plan“ and the "Monju Research Plan”

Promote the R&D on the following issues. R&D on Fast Rectors

R&D on Monju R&D towards the establishment of proven technology of FR Safety enhancement of FBR /FR Demonstration technology development through

ASTRID cooperation R&D on Nuclear fuel cycle

Reprocessing of SF and fuel fabrication Reducing the volume and potential radio-toxicity

of radioactive wastes

ASTRID

Future Planning of R&Ds

“Monju”

“Joyo”

5

AtheNa facility

Reactor

Fuel cycle plants

Commercialized FBR cycle

R&D for reduction of the waste volume and radiotoxicity

Fuel fabrication Fuel development and irradiation test Reprocessing (separation technology) Reactor characteristics and reactor system The entire system evaluation

R&D on “Monju” Core and fuel technology Equipment and systems design technology Sodium handling technology Plant operation and

maintenance technology MA-bearing MOX fuel irradiation test Demonstrate natural circulation decay heat

removal

6

R&D for safety enhancement of FBR/ FR International development of safety

design requirements National development of structural material

specifications and standards and its reflection in international standards

Core damage mitigation technology development Thermal-hydraulics analysis and evaluation methods development ASTRID cooperation

• Certainty of design • Confirmation of the

safety margin

Next reactor design with international safety design requirements Analysis code, design approach Equipment design

Operation of next reactor Operational standards, operation

and inspection procedure Maintenance and repair method

• Operation and maintenance policy (Inspection frequency, etc.)

•Optimum system concept •Establishment of design technology

•Certainty of effectiveness

Feasibility confirmation of reduction of the waste volume and radiotoxicity Burnup theory demonstrated in

the current Am-containing core Demonstration of MA-bearing fuel

irradiation

Reflection of R&D results into the demonstration and commercialization

Establishment of demonstration technology

7 R&D on “Monju”

Aggregate of the fast breeder reactor technology Core and fuel technology Confirmation of higher isotopes of Pu core

characteristics based on the actual reactor data. Equipment and system design technology Plant system design technology Design technology of large sodium equipment

Sodium handling technology Development of in-service-inspection technology

for the reactor vessel, etc. Plant operation and maintenance technology Establishment of a maintenance program in light of

the characteristics of the FBR power plant, etc.

Irradiation test (X-ray CT image)

Core design approach and core management technology

Examples of specific reflections

Reactor kinetic characterization and shielding evaluation methods

Aging characteristics and Integrity of sodium equipment

R&D for reducing the waste volume and radiotoxicity Evaluate of the MA transmutation and the irradiation behavior

by full-scale irradiation tests with MA-bearing MOX fuel, etc.

R&D of enhanced safety Demonstrate the decay heat removal in the actual plant as a

feature of the sodium-cooled FR with natural circulation

◉Aggregate the outcome of the FBR technology development including the technical feasibility of the FBR plant, and Reflect it in the next reactor design by utilizing "Monju“ of our own design, manufacturing, and construction.

Na management techniques of loop-type FR power plant

Global Standards for SFR SDC and SDG 8

SDC

SDG

Safety Fundamentals

Safety Requirements Targets of global standards

Domestic Codes and Standards

Design guides of the Reactor Coolant System and Associated Systems in Nuclear Power Plants

<Hierarchy of Safety Standards>

◉ Japan leads to build safety design requirements (safety design criteria (SDC) / guidelines (SDG)) toward the safety enhancement of SFR in the world.

◉ FR development countries intend to reflect them in their safety regulations and safety design. De facto global SDC&SDG

Positioning and the purpose of SDC / SDG Global standardization of the safety design concepts toward the

commercialization (mainly on design basis accidents) Safety improvement of FRs in the world by Japanese initiative

Main results: SDC report (approved by GIF in May 2013) Review is in progress among regulatory bodies/technical support organizations

of FR development countries and by international organizations (IAEA, OCED/NEA/CNRA, etc.) Russia, China, India etc. intend to reflect in the safe design

Major progress Design and R&D on Safety enhancement including severe accident measures

Design: Decay heat removal system, Shutdown system, Seismic isolation R&D: fuel, safety, and plant system

Design and related outputs could be directly incorporated into the Japanese SFR development.

R&D collaborations are on going through experiments and analysis methods.

Status of ASTRID Collaboration

◉ The French President and the Japanese Prime Minster agreed to collaborate on the development of the 4th Generation Reactors.

◉ General Agreement was signed on May 5, 2014. ◉ Implement Agreement was singed on August 7, 2014.

ASTRID technologies

Japanese SFR technologies

Common issues ・Safety design ・Sodium technologies ・Fuel technologies

9

Research projects of FR cycle technologies

Fuel Fabrication

FR

Evaluation of volume reduction, etc. by

utilization of FR cycle Fuel Fabrication: Remote MA-bearing

MOX fuel fabrication technology

Determination of fuel composition range applicable

Reactor Characteristics & Reactor system: Feasibility confirmation of FR plant

technologies Acquisition of characteristics of MA-

containing core

Fuel Development & Irradiation Test: Systematic irradiation tests

of MA-bearing MOX fuel, High Pu-contents MOX fuel

Spent MOX fuel

Monju

Joyo

Pu-3

CPF

◉ Improvement of the flexibility in Pu use, verification of MA partitioning and transmutation technologies, etc. are necessary for obtaining technological perspective on the reduction of the volume and radiotoxicity of radioactive wastes

Reprocessing

Reprocessing: Development of MA partitioning

process and performance evaluation Establishment of feasible process concepts

Comprehensive system evaluation: Integration of information in each area &

narrowing prospective system concepts Verification of effects on reduction of the

volume and radiotoxicity of radioactive wastes

AGF FMF

U, Pu, MA

MA-bearing new MOX fuel

10

11 Status of Technological Development for Partitioning and Transmutation of Long-lived Nuclide Utilizing FRs

Succeeded at absorptive partitioning of more than 99.9 % of MAs in HLW liquid

多孔質シリカ吸着材の 写真(寸法付) とか、、、

Particle diameter about 50μm

SiO2-P Fine cross section

(MA: minor actinide, Am: Americium, Cm: Curium)

Partition MAs from liquid waste

Evaluation of effects of MAs on fuel performance is necessary Effects of re-distribution behavior of MA under irradiation on physical properties (eg. melting point)

Fabricate MA-bearing fuel pellets

Sophistication of fuel fabrication technologies Microstructure and

oxygen content control Acquisition of basic data &

development of technology for fabrication condition optimization

Microstructure

MA-bearing fuel pellet

φ5.4 mm

8 mm

Am co

ncen

tratio

n (re

lative

value

)

High Am concentration at central (Temp. at center: about 2400℃)

1mm

Center

Irradiation tests of-MA bearing fuels at 2000℃ or higher

◉Carry out a series of tests (from partitioning, recovery and conversion to fuel fabrication, irradiation and post irradiation examination) on MAs from spent fuel using existing facilities

LWR FR

Conditioned MA-bearing raw powder

MA fuel for Irradiation Tests

Irradiated MA- bearing fuel

Irradiation Irradiated

MA-bearing fuel

O-arai: AGF

Irradiation Fuel Fabrication

O-arai: FMF

Post Irradiation Experiments

Spent fuel High-radioactive liquid waste (HLLW)

SmART Cycle

Oarai: Joyo

Monju

Tokai: CPF

Partitioning

Pore diameter about 500nm

Outer peripheral part of fuel pellet

12 MA separation technology using extraction chromatography

MA could be separated from genuine HLLW using extraction chromatography method which has been developed at JAEA.

SiO2-P adsorbent Porous silica covered stylene divinyl benzene polymer and impregnated extractant in the pore

Extraction chromatography MA separation is achieved by flowing

the feed with MA and eluent into the packed column of solid adsorbents.

Advantages：Less secondary waste, compact equipment etc.

MA separation results from HLLW （2-step separation）

SiO2-P adsorbent

Organic polymer

Extractant impregnation

Grain size 40-60 μm, Pore size 500-600nm

Column

Feed

MA+RE

1st stage MA+RE separation

（RE：Rare earth elements）

2nd stage

Waste

Eluent

MA

MA+RE

Adsorption and elution

SiO2

Schematic of surface layer on adsorbent

0 2 4 6 8 10 12 14 16 18 20 22 24 26 280

0.5

1 Sr Pd Ru-106 Sb-125 Cs-137 Ce-144 Eu-155 Am-241 Cm-242 pH

C/C

0

Through Bed Volume

DV Feed 3M HNO3 H2O 50mM DTPA (pH=3)

3

1.5

pH

Chromatographic separation of genuine HLLW by the column with CMPO/SiO2-P adsorbents

Chromatographic separation of real HLLW by the column with BTP/SiO2-P adsorbents

MA+RE recovery

MA recovery

0.0

0.2

0.4

0.6

0.8

1.0

0 1 2 3 4 5 6 7 8 9 10 11 12

C/C

0

Through bed volume

Sr YZr MoBa LaCe NdSm EuGd AmCm

DeadVol.

Feed 1 mol/dm3 HNO3 H2O

Waste

Eluent CMPO BTP

PO O

NN

N N N N

NN

Examples of extractant

MA separation

Adsorption and elution

HLLW : high level liquid waste

13 Unified physical property model for fuel technologies

Basic properties and correlations have been studied, and a fundamental technology to control oxygen content in MA-bearing MOX was obtained.

Development of fuel behavior models in fuel fabrication process and operation conditions based on the property model

Oxygen control technology Sintered MOX pellets

MOX granules

Calculation examples by the correlations

Calculation of oxygen content change in pellets during heat treatment

Apply to sintering condition

Sintering condition for low oxygen content pellets

U/Pu ratio adjustment in nitric acid solution

Products

Microwave heating de-nitration and granulation

Pressing

Sintering and O/M adjustment

Development of an unified physical property model through property measurements, database construction and mechanistic relational equation derivation

Chemical stability Oxygen diffusion

1.96

1.98

2.00

2.02

2.04

500

1000

1500

2000

0 2 4 6 8

FJK

Tem

pera

ture

(o C)

Oxy

gen

cont

ent (

O/M

ratio

)

Heat treatment time(h)

O/M ratio Temperature

ExperimentAverage O/M ratio

CalculationsEquibrium O/M ratioAverage O/M ratio

14 Irradiation performance of MA-bearing MOX fuel

Slight increase in MA (Am) concentration near the center of the fuel pellet due to irradiation was confirmed from the result of the MA-bearing fuel irradiation experiment in Joyo. A computer analysis model was developed which could reproduce this phenomenon.

・Irradiation condition (LHR / period) ～450W/cm / ～24hr

・MA content of fuel pellet Am：～5wt% / Np：～2wt%

Radial profile of Am concentration

Am measurement

Central void

Center Edge

Center Edge

Short-term and high linear heat rate (LHR) irradiation of MA-bearing MOX fuel pins

1mm

Ceramography of irradiated MA-bearing MOX fuel pellet

Change of Am concentration due to irradiation was calculated by using vapor phase and solid phase diffusion analysis models.

Irradiation in Joyo

15 Minor Actinide Transmutation in FRs

Items JSFR (Breeder) Pu-MA burner

Electric output [MWe] 1500 750

Breeding ratio (Conversion ratio) 1.1 0.58

Core height [cm] 100 75

Fuel pin diameter [mm] 10.4 6.5

Fuel lifetime [EFPD] 3200 1295

Fuel burnups [GWd/t] 146 151

MA content in heavy metal [wt%] 3.0 4.6

Basic characteristics of FRs for MA transmutation

① Transmutation of initial MA load ② MA generation from core fuel

Fast reactor core concepts for MA transmutation were developed. When the high-burnup fuel is realized, 50-60% of initial MA load in fuel fabrication can be transmuted before fuel exchange.

Core layout of a Pu-MA burner

Core height 75 cm

Core equivalent diameter 3.3 m

Fuel sub-assembly Primary control rod Backup control rod

Linear heat rate [W/cm]

16 Development of ODS steel cladding tube

ODS steel cladding tube

• Increasing burnup => High MA transmutation ratio, Reducing nuclear fuel costs

•High power generation efficiency produced by high outlet Na temperature

•Increasing fuel lifetime, Safety enhancement

⇒ Fuel performance improvement in terms of environmental, safety and economical aspects

Evaluated MA transmutation ratio vs neutron dose, burnup

Recent progress:

Substantial swelling resistance

Modified type 316 steel

Ferritic steel， ODS steel

0 50 100 150 Average discharge burnup (GWd/t)

twice

0 10 20 30 40 50 60 Neutron dose (x1026 n/m2, E>0.1MeV)

MA tr

ansm

utatio

n rati

o (%

)

0

20

40

60

80

100

Modified type 316SS Ferritic steel ODS steel

Swell

ing (

vol.%

) 0

1.0 2.0 3.0 4.0 5.0

Fabrication technology development, which can reliably and consistently produce high strengthen ODS steel cladding tube

Excellent mechanical strength at high temperature

JAEA has developed the long lifetime oxide dispersion strengthened (ODS) steel cladding tube, which can double MA transmutation ratio compared with existing technology. The fabrication process was successfully developed to reliably produce high performance ODS steel tubes.

Creep strength

17

JAEA promotes SmART cycle project which includes MA separation from irradiated fuels and irradiation of the MA-MOX products in Joyo for the first time. Preparation of HLLW was completed to get MA more than 1g.

SmART cycle (Small Amount of Reuse Fuel Test Cycle)

Fuel fabrication

Irradiation PIE

Separation

LWR, FBR

Irradiated fuel

AGF (Oarai)

Joyo (Oarai)

CPF(Tokai)

SmART cycle

【Objectives】 • Separation and transmutation data of MA etc. • Feasibility of small scale of partitioning and transmutation cycle

Irradiation behavior of MA isotopes in the MOX fuel will be evaluated for the first time in the world.

• Effect of FPs on fuel fabrication and irradiation

• Effect of MA isotopes on transmutation • Material balance of MA through the cycle

【Significance】

【Current status】 Shearing, dissolution and extraction of

irradiated Joyo fuel pins were completed.

MA separation from the raffinate of extraction will be implemented hereafter.

【MA amounts】 • MA more than 1g is separated from

4 pins of irradiated Joyo fuel

MA yields recovered from irradiated fuel are top level in the world

Denitration, conversion

For Restart of Irradiation Test Activities in Experimental FR Joyo

Door valve

Temporary pit cover

Cask

Wire jack

Pull-out work of UCS

R&D for inspection and repair the interior of the reactor vessel at high dose (max. 300 Gy/h) and high temperature (~200 ℃ )

18

Completion of the restoration work on the damaged experimental device

May-Nov. 2014: Completed replacement of the UCS and retrieval of the experimental device

June 2015: Completed the re-installment of the retrieved device; brought Joyo back into normal state

• High radiation resistant fiber-scope, etc. • Replacement of the large in-vessel component:

Upper Core Structure (UCS) (~16.5 ton) • Retrieval by using a remote device

◉ Completed the restoration work on the damaged experimental device that had been in trouble and brought Joyo back into recovered and normal state

◉ Plan to submit an application for permission of changes in reactor installation in FY2016

◉ Plan to conduct irradiation tests, etc. related to R&D for reduction of the volume and radio-toxicity of radioactive wastes and for ASTRID project after the restart

Tem

pora

ry c

radl

e

19 Direction of the fast reactor cycle R&Ds should be headed

At the time the policy towards the commercialization of FR cycle is embodied, the R&D should be steadily promoted in order to present the following achievements. Development of research infrastructure Early improvement and re-start of the test facilities such as Monju," "Joyo” Development and accumulation of human resources and technology platform to

support the FR cycle technology using "Monju," "Joyo,” fuel cycle facilities, etc. R&D Confirmation of the technical feasibility of innovative technology that reflects the

safety enhancement measures in light of the Fukushima Daiichi nuclear power station accident Establishment of FR reference plant concept incorporating international safety

design requirements (SDC, SDG) Prospect of technical feasibility on reducing waste volume and its radiotoxicity Drawing the path to commercialization (technology roadmap), etc.

In the implementation of the policy, promoting dialogue and information sharing with stakeholders, reflected in the direction of R&D

To that end, along with promoting human resources development and technology transfer specifically, operate test facilities in the highest priority on safety by implementing the response to the new regulatory regulation ASAP.