Overview of Physics Aspects of

Different Transmutation Concepts

OECD

Nuclear Energy Agency

TABLE OF CONTENTS

1. Introduction

2. R&D Programs on Partitioning and/or Transmutation

2.1 R&D Programs and Their Principal Objectives

2.2 Nuclear Fuel Cycle Consideration on Partitioning and

Transmutation

3. The Activities of NEA on Transmutation

3.1 Activities of the NSC and Data Bank

3.2 Activities of the SSD

4. Proposed Transmutation Concepts

4.1 Thermal Reactors

4 . 2 Fast Reactors

4.3 Accelerator-Driven Transmutation Systems

5. Transmutation Capability of Proposed Concepts

5.1 Thermal Reactors

5 . 2 Fast Reactors

5.3 Accelerator-Driven Transmutation Systems

6. Calculational Methods

6.1 Thermal and Fast Reactors

6.2 Accelerator-Driven Systems

6.3 Used Calculation Methods for proposed concepts

7. Integral Experiments

7.1 Thermal Reactors

7.2 Fast Reactors

7 . 3 Accelerator-Driven Systems

8. Recommendations and Conclusions

References

Appendix

1. INTRODUCTION

Partitioning and transmutation ( P - T ) of actinides and

fission products is currently one of the important subjects of

long term research and development in several OECD countries and

also in Russia,from the viewpoints of its perspectives of poten-

tial utilization of resources and possible advances in waste

management.

In the P-T, undesirable nuclides in spent fuel are first

separated and then incinerated in a nuclear reactor or other

transmutation device. The nuclides that would be incinerated by

the P-T are generally considered to be the transuranium (Sp,Pu,.4m

and Cm) and certain long-lived fission-products,such as 1-129 and

Tc-99. Implementation of the P-T concept would involve intensi-

fied processing (partitioning) to remove long-lived radionuclides

from the waste stream and subsequent use of a transmutation

device to convert transuranium to fission and fission-products to

shorter-lived radionuclides.

Various P-T concepts for high-level radioactive waste (HLW)

are currently being proposed or under investigation at many

organizations in the world. Partitioning technologies for spent

fuel from light-water reactor and fast reactor include aqueous

and pyrochemical processes. Transmutation technologies include

application of light-water reactor,fast reactor and proton accel-

erator-driven transmutor.

Most of the assessment and research performed to date has

focused on the fundamental processes or the system design the

transmutation device or the partitioning process rather than the

integration of a P-T system into nuclear fuel cycle. Under ~ h e s e

circumstances, the OECD/SEA Suclear Development Committee (SDC)

decided to initiate a system study on P-T as international col-

laboration program to identify advantages and disadvantages of

introduction of P-T technologies into nuclear fuel cycle.

With regard to transmutation,various transmutation concepts

are under investigation by using their own data and methods.

However,their physics characteristics are not necessarily pre-

dicted ivel1,because of insufficient accuracy in used data for the

transuranium and long-lived fission-products. In this contesT,the

OECD/NEA Nuclear Science Committee (NSC) decided to form a task

force to review the physics aspects of various proposed transmu-

tation concepts. The task force requested about 30 specialists

working for transmutation to inform the physics aspects of izheir

concepts, including objectives of transmutation, fuel cj-cle con-

cepts with P-T, physics characteristics of proposed transmutors

with used data and methods (see Appendix). The presenT report

overviews the physics aspects of different transmutation concepts

,based on the above mentioned information.

-

2 . R&D PROGRAMS ON PARTITIONING AND/OR TRANSMUTATION

2.1 R&D Programs and Their Principal Objectives

(1) French SPIN program [I]

In order to respond to the public concern about wastes and

in particular the long-lived high level ones,a French law issued

on December 30,1991 identified the major objectives of research

for the next fifteen years,before a new debate and possibly a

decision on final wastes disposal in Parliament. To comply with

the requirements of the management of long-lived high lex-el

wastes, the CEX has launched an important and long term R&D

program. A part of this program called SPIN is devoted to separa-

tion and incineration of these wastes and it includes two sub-

programs :

a) In short and mid term perspectives (1991-2000). PURETEX

aiming primarily at reducing the volume of wastes from

reprocessing from 1.Sm3 to O.Srn3 per ton of reprocessed

heavy metal. This result will be obtained by modifying the

PUREX process in order to be able to eliminate bitumen and

by improving the conditioning of solid wastes.

b) In a long term perspective,ACTINEX devoted to the separation

and transmutation of long-lived elements in view of reducing

wastes toxicity by a factor 100 and then 1000 compared to

direct disposal within 20 and 40 years respectively. With

regard to transmutation,the feasibility of actinide incinera

tion by PWR and fast reactor are investigated. In parallel

the possibilities of advanced systems such as specific nucle-

ar reactors and accelerators are also investigated.

The objective of this program is to limit volume and poten-

tial radiotoxicity of minor actinides in deep storage and also

the quantities of long-lived fission-products for the ouIcorne of

deep storage.

(2) Japanese OMEGA program [ 2 1

The Japan's -Atomic Energy Commission launched in Occober

1988 a comprehensive basic research program (OMEGA) to explore

the feasibility of utilizing HL\V as useful resources and widening

future options for waste management by releasing the report

entitled "Long-Term Program for Research and Development on

Nuclide Partitioning and Transmutation". In the partitioning

technology development, following scopes are to be studied;

- Partitioning from HLW or in main reprocessing process.

- Recovery of useful metals from insoluble residue.

- Technology to make use of separated nuclides.

The development of transmutation technology falls into following

two major R&D categories;

- Application of nuclear fission reactors.

- Application of high-intensity accelerators.

Fission reactors include fast breeder reactor power plants

,and specially designed actinide burner reactors,which would be a

very effective mean to transmute long-lived TRU. High-intensity

accelerator aims at transmuting TRU and other long-lived FP.

The program is planned r o be carried out in two phases.

The first phase (1988-1996) sill be dedicated in general for

basic studies and tesIings TO examine feasibilities. to develop

fundamental technologies, and to conduct overall assessment for

various candidate concepts. In the second phase, engineering

scale tests will be conducted to verify the systems and technolo-

gies evolved by that time. Because of the exploring nature of the

program, several potential-mezhods or concepts are being investi-

gated in parallel in both parritioning and transmutation areas b y

the Japan Atomic Energy Research Institute (JAERI), the Power

Reactor and Nuclear Fuel Development Corporation (PNC), and the

Central Research Institute of Electric P o ~ ~ e r Industry (CRIEPI).

The objective of the program is to significantly reduce the

source term of the long-term potential hazard which arises from

HLW.

( 3 ) ALMR Actinide Recycle Program [3]

The Advanced Liquid ILletal Reactor (ALWR) Actinide Recycle

System is being developed in the United States for application

early in the 21st century. The system is expected to have the

ability to fulfill multiple missions including: (1) the conver-

sion of excess Pu to produce power, (2) utilizing the tremendous

energy potential associated with spent LWR fuel. (3) providing

long-term energy security,and (4) achieving a significant reduc-

tion in the heat load and time constant associated with processed

waste. The ALPlR is a fast reactor design,and its plant design and

development program is a national program involving wide partici-

pation by US industry as well as national laboratories. universi-

ties,and international organizations. The AL?lR utilizes the metal

fuel cycle being developed by Argonne Sational Laboratory ( X S L )

which inherently recycles actinides to the reactor in the refer-

ence breakeven/breeder and burner designs.

The objectives of the program include potential reduction of

major long-lived toxic actinides and removal of heat-producing

radionuclides.

(4) Los Alamos ATW Program [ < I

The Accelerator Transmu~ation of nuclear Waste (AT\Y)is being

developed by the Los Alamos Sational Laboratory (LANL) in the LS.

The ultimate goals of xhe ATG' development effort concern creation

of a system that can desxroy long-lived migratory fission-

products and high toxicity actinides associated with high-lex-el

waste storage, and creation of an advanced energy production

concept that can rectify major obstacles facing nuclear power.

.Application of an ATW system to HLW management programs can

have a significant impact. An ATW system applied to cleanup of

long-lived defense HLW at a DOE site could provide management

options that minimize or eliminate HLW leaving the site. Long-

lived nuclide destruction could also aid creation of more robust

on-site waste storage forms. The impact of an AT\Y system on

geologic repository storage could be significant. It could d e l a ~ .

or potentially avoid altoge~her the need for a second geologic

repository. It could also reduce the performance required of a

geologic repository, e.g. reduce the waste isolation period from

10000 years to a period of around SO0 years.

A T application development involves a staged zpproecn .

Specific technology development allows components to be developed

for a next application stage. A relevant example is thar of

defense waste application. The ATW system components developed

for it share a high degree of commonality with that could be

applicable to destruction of commercial high level waste.

(5) R&D Program at BNL [ 5 , 6 , 7 ]

Brookhaven Sational Laboratory (BgL) in the US has 2roceeded

research programs on transmutation by using large proton linzar

accelerator ( P H O E S I X program), small poiver accelerator 2nd high-

flux thermal reactor. The objective is to substantiallj- reduce

some of the most significant challenges in building a n-aste

repository,by transmuting key elements,such as Pu,minor zctinides

and a few of the long-lived fission products.

(6) R&D Program at ENEA 181

A research program on accelerator-driven transmutation is

being proceeded at ENE.4 under the close collaboration with L.4SL

in the U S . The objective of transmutation is the radiotoxicity

reduction of minor actinides potentially introduced into the

biological sphere with a not very significant increase of the i;\';ti

cost, and in the case of Pu, straight forward reduction of i c s

stockpiles as a solution of Pu problem.

( 7 ) R&D Program a t Tosh iba Corpora t ion [ 9 1

Future contribution to overall energy production by nuclear

power depends on the development of efficient transuranic ele-

ments recovery system for spent fuei reprocessing. The system is

expected to lead to optimum resource utilization and substanrial

reduction in HLW associated with large scale energy production.

The objective of transmutation is:

a) Substantial reduction in volume and long-term radiotosicity of

HLW in fuel recycling sysIem,and

b) Simplification of future TRG-recycling technology in L\VR and

FBR aiming at total cost competitiveness, including reactor,

reprocessing and storage in the zotal nuclear energy system.

( 8 ) R&D Program a t Royal I n s t i t u t e of Technology [lo]

A research program on accelerator-driven transmutation is

going on at the Royal Institute of Technology in Sweden, under

the collaboration with LANL.

The objective of transmutation is:

a) Reduction of radiotoxici~y in the geological storage, drastic

reduction of the duration of radiotoxicity,

b) Combination of transmutation of waste with energy production,

i.e.effective fissioning of transuranic isotopes,and

C) Opening a new "subcritical nuclear option" for energy produc-

tion.

( 9 ) R&D Program at.IPPE [ll]

A research of transmutation with fast reactors is being

conducted at the Institute of Physics and Power Engineerzng

(IPPE) in Russia. The objective is reducti.on of radiotosicitl; in

the deep storage,and finally reduction of health and enl-ironmen-

tal risks. Recycling of Pu in fast reactors could reduce the long

term radiotoxicity of TRL up to 10 Iimes. 2nd recycling of Pu.Sp.

and Am would reduce it up to 100 times. The reduction of h e a l ~ h

and environmental risks is still uncertain.

(10) R&D Program at ITEP [I21

4 research of accelerator-driven transmutation is being

conducted at the Institute of Theoretical and Experimenrel Phj-s-

ics (ITEP) in Russia. The objective is the radiotoxicity and mzss

reduction of HLW before deep (underground) storage and xne energ>-

production in the blanket of an accelerator-driven transmutatian

system.

2.2 Nuclear Fuel Cycle Consideration on Partitioning and Transrnu-

tation

The nuclides to be transmuted have to be recycled many tices

to incinerate them sufficiently. since the considered ~ransrnu~e-

tion cross-sections may be relatively low and the highest achiex--

able neutron flux is limited. This means that the transmutation

technology has a strong correlation with the partitioning rech-

nology and also with nuclear fuel cycle. If the transmutarion

efficiency in one cycle is low, the partitioning efficiency must

be high to achieve the overall required reduction factor during

several cycles. This may give large influence to the overall

nuclear fuel cycle concepts. In this context,the proposed nuclear

fuel cycle concepts with the P-T technology are illustra~ed here,

although it is out of scope.of the present discussion.

(1) SPIN Program

The SPI3 program consists of tn.0 major projects: PLRETEX and

ACTINEX, as mentioned above. The ACTINEX project aims ar separat-

ing Pu, Np, .a, Cm as well as long-lived fission products in an

advanced reprocessing system,and furthermore at incinera~ing them

by using thermal or fast reactor, or accelerator. One of the

nuclear fuel cycle concepts with the P-T technology considering

in the ACTIKEX is illustrated in Fig.2.1 {(P~+Np+~m+Crn) balanced

scenario).

( 2 ) OMEGA Program

The OMEGA program is composed of two major R&D areas: the

transmutation technology development and the partitioning tech-

nology devel-opment. Since several potential methods ot- concepts

in both R & D areas are being investigated in parallel by the JAERI

,PNC and CRIEPI as seen in Fig.2.2, these organizations propose

their own nuclear fuel cycle concepts.

1) J A E R I

The JAERI aims to develop a partitioning process of TRU and

long-lived fission products from HLK and their subsequent trans-

mutation system based on a specially designed fast burner re~ctor

or an intense proton linear accelerator. A double strata nuclear

fuel cycle concept is proposed as shown in Fig.2.3, in which the

first cycle is the conventional fuel cycle and the second is the

P-T cycle. Another concept under investigation is concerned with

TRU recycling in LWR as seen in Fig.2.4.

2) PNC

The PNC's approach is to develop a partitioning process as a

part of advanced reprocessing system where the improved PUREX

process is closely combined with TRUEX-like TRU separation proc-

ess with emphasis on TRU recycling in a FIOX fueled FBR system.

- The nuclear fuel cycle concept is given in F i g . 2 . ~ .

3) CRIEPI

The CRIEPI aims at a concept to separate TRU from HLK by dry

process with pyrometallurgical methods and then to transform them

to the nuclides with shorter half-lives in a metallic fuel FBR.

The nuclear fuel cycle concept is shown in Fig.2.6.

( 3 ) ALMR Actinide Recycle ( I F R ) Program

The IFR is the entire reactor system consisting of reactor,

fuel cycle and waste process. In the IFR pyroprocessing, minor

actinides accompany che plutonium stream and therefore actinide

recycling occurs naturally in a metallic fuel L Y R . The schemaric

picture of the IFR concept is given in Fig.2.7.

( 4 ) LANL ATW Program

The feed to the ATU system is the separated actinides,&p,Pu,

Am and Cm plus the long-lived fission products (LLFP1s).Tc-99 and

1-129. Chemical separation capacity is assumed to allon the L\YR

spent fuel to be partitioned into various streams before :lTR' is

feasible. The ATVi target/blanket converts the LLFP's LO stable

species by neutron capture and the actinides to fission products

by capture and fission. There is a small chemical facility to

remove any LLFP's created by fission for recycle, and to continu-

ously recycle the actinides. XIaterial flow in the .ATW is shonn in

Fig. 2.8.

(5) BNL Program

While not tied to a specific fuel reprocessing/recycling

technology, much of the PHOE5IX concept is based on the clean use

of reactor energy (CURE) approach proposed by the Westinghouse

Hanford [13], which is a waste partitioning process based on the

well-known Purex process and the newer Truex process. The waste

partitioning and transmutation system shown in Fig.2.9 is cornnlon

in all the transmutation concepts including Lhe P H O E Y I S concept

proposed by BNL.

(6) Consideration at ENEA

The present fuel cycle will be completed adding, after the

reprocessing, two new steps : partiIioning (recovery of minor

actinides from H L K ) and transmutation (destruction of TRI;). This

last step can be carried out in different ways: Pu burning sepa-

rated from minor actinide burning or the two together. The first

strategy can be developed -using the IFR for Pu burning and the

ATW system for minor actinide burning or the two tasks can be

performed by means of the latter strategy.

( 7 ) Consideration at Royal Institute of Technology

The transmutation strategy is aimed to propose an energy

producing transmutation system based on thorium fuel cycle. The

Institute would try to design a proliferation safe syscem which

will minimize chemistry requirements, possibly only one to the

end-of-life stage

(8) Program at Toshiba Corporation

An integrated nuclear energy production system should be

developed to minimize impact on the global environment in future.

The recovered minor actinides will be used as fuel together with

Pu in LWRs and FBRs. The considered nuclear fuel cycle is shown

in Fig.2.10.

(9) Program at IPPE

The considered nuclear fuel cycle includes follo~vings:

- Closed fuel cycle w i t h recycling of Pu mainly in f a s ~ reactors

with breeding ratio of nearly unity,

- Partitioning of minor actinides, and recycling of .Am and Np-237

together with Pu in fast reactors with breeding ratio of nearly

unity ,

- Long-term storage of Cm,and recycling of accumulated Pu-240,and

- Burning of excess of Pu and minor actinides in special burner

reactors including one cycle of Pu in thermal reactors.

- . s l g . 2.1 One of ?u and M A b f i e cyc l i ng Scenarios (CE4)

- (Pu+Np+.Am+Cm) Ba l -anced Scuario -

Fig. 2.3 Fl-ow of Bigh-level Radioactive Waste per Year Throuh Doub1.e Strata F u e l Cycle Conbined with far tit ion in^ and Transmutation ( JLAE1?I )

F i g - 2 . 4 A Concep t of Pu and MA Fuel Cycle in L R R s ( JAERI)

Fig. 2.5 Advanced Reprocessing Systems (PNC)

F i g . 2.6 Scheme of TRU Partitioning and rans smut at ion with Metallic Fuel FBR Recycle ( C R I E P I )

F i g . 2.7 A Schematic P i c t u r e of t h e I F R C o n c e p t , (AfXL)

--

F i g . 2 . 8 General Concept of ATW S y s t e m (Lh\L)

, - , . . . . . - . - - - - 3 ; -,---, - - i . - ; - , - - , - . L . . . - - - , 3 . . - . i . - . 5--, i.! -.

Fig. 2.9 Waste Component Flows through Partitioning and Transmutation System ( B N L )

Fig. 2.10 Fuel Cycle with Partitioning and Transmutation under Investigation (Toshiba Corporation)

3 . TElE ACTIVITIES OF NEA ON TRANSMUTATION

The 5E-A has been involved since the lace 1980s in x-arious

activities on Actinide and Fission Product Partitioning and

Transmutation.

3.1 Activities of the NSC and Data Bank

The program of work of the Nuclear Science Committee iSSC)

covers several items related EO transmutation. Some of ~ h e s e

programs have been performed within the 8SC's own program of

i~ork, whereas others have been carried out using the voluntary

contributions from the Japanese government to the NE.A.

The main emphasis has so far been on different nuclear data

needs for accelerator and reactor transmutation applications. -4

number of projects have been or are being pursued in the field of

intermediate energy nuclear data. Three studies bvere recenclj-

completed: one on the availability of experimental data and

nuclear model codes, one on the requirements for an evaluated

nuclear data file, and one on the international comparison of xhe

performance of compute codes used in intermediate energy data

calculations. A data base of available experimental data will be

set up , as well as a cooperative project on an evaluated nuclear

data file.

(1) Completed Program

1) Review of Fission Product Yields and Delayed Xeutron Data from

the Actinides (July 1990)

-4 critical review of available fission product yields and

delayed neutron data for the actinides of interest to trznsnuxa-

tion, i.e.,Kp-237, Pu-242,-Am-242m,-b-243, Cm-243 and Cm-245 was

performed. The study has identified gaps and inconsistencies in

the existing data base and also identified priority areas for

further experimental.,theoretical and evaluation efforts.

2) Review of High Energy Data and hlodel Codes for .-iccelerator-

Based Transmutation (December 1992)

This srudy investigated the availability of experimental

intermediate energy data. The need for a comprehensive compila-

tion effort of these data was stressed. The most important nucle-

ar theories and some of the associated nuclear model codes were

described, and thei.r applicability to intermediate energy nuclear

data calculations was discussed.

3) Requirements for an Evaluated Nuclear Data File for .-lccelet-a-

tor Based Transmutation (June 1993)

The importance of evaluated intermediate energy nuclear daca

files as part of a global ca1culat;ion scherrre for accelerator-

based transmutation was discussed. The resulting report contains

a p r o p o s a l f o r setting up the f o l l o ~ ~ i n g three data libraries:

- a data library fron 0 TO 100 >lev (first priority).

- a reference data library from 20 to 1300 >Is\-,

- an activation library from 0 to about 100 >lev.

4) Comparison of Codes for Calculation of Intermediate Suclear

Data (February 1994)

An international comparison study was coordinated. The aim

was to assess the predictive ability of compucer codes used in

calculating in~ermediate energy charged particle data. Two cases

were defined: one where thin target data of Zr-90 and Pb-20s

should be calculated and one with the objectil-e to predict the

neutron yield and mass distribution of spallation products fron

800 ?lev proton bombardment of \Y. The results, TO be compared to

experimental data, were collected and analyzed. A specialist

meeting on Intermediate suclear Data will be held on 30 >ray-1

June 1994 to discuss the result of the study.

(2) On-Going Programs

1) International Evaluation Co-operation

The NSC LYorking Party on International Evaluation Co-

operation has started an activity to evaluate incerrnediate

nuclear energy data. The project will partly build on the t-e-

sults from the above study on the Requirements for an Evaluated

Nuclear Data File for Accelerator-Based Transmutation. The

--

\Torking Party has established a subgroup (Subgroup 50.13) on

this subject. The Subgroup has started to inquire data needs

and is planning a start-up meeting on 1 l ~ h $lay, in Gatlingburg.

The Group will discuss items such as the type and forma1 of

the data to be included in the library, before starting f he

actual evaluation effort. .The first data files are expected to be

ready for testing in 1994.

2) A Data Base for Experimental Intermediate Energy Data

.According to a recommendation in the above report on Re\-ien

of High Energy Data and >lode1 Codes for Accelerator-Based Trans-

mutation, the Data Bank has started to set up a special purpose

"transmutation" data base combining relevant material from esist-

ing numerical and bibliographic data bases and supplementing the

information i~ith the data of essential interest to transmutation

needs. This program has been approved by he Steering Committee

as the activity reinforced by the voluntary conLribution of the

Japanese Government. The result of the abol-e program, Comparison

of Codes for Calculation of Intermediate Xuclear Data, nil1 pro-

vide the basis for identifying the needs in improving the comput-

er programs used for the modeling. At the Specialist Meeting on

Intermediate Energy Nuclear Data, it will be discussed on nhich

data compilation and evaluation will be needed in order to corn-

plement the data base.

3 . 2 Activities of the NDC

The Nuclear Development Cornmixtee (SDC) performs "Ac~inide

Information Exchange Program". Under the program.an International

Information Exchange Meeting on Partitioning and Transmutation (P

-T) was held in 1990 and 1992, with the guidance of Liaison Offi-

cers nominated by member countries. This program is supported by

the voluntary contribution-by the Japanese Government. This pro-

gram was started in 1989 as a five year program. It is espec~ed

that the NDC will extend its activity on P-T at the end of i994

for an additiona1,possibly 5 years.period.

(I) Completed Programs

1) 1st International Information Exchange ?Ieeting

(?lit0 City, Japan, 30s-ember 1990)

A number of papers were presented both on policy orienta-

tions and scientific aspects. The discussion ,which concentrated

on wide ranging ideas regarding future technologies, led to the

conclusion that several disparate approaches had already been

taken.

2) Specialist 3leetings

Following the first Information Exchange ?leeting, the SE.4

helped organize two specialist meetings. The first meeting was

held in November 1991on the topics of partitioning technology. A

--

wide range of processes for the separation of actinides and

fission products,both wet and dry,n'ere presented. The second ivas

in March 1992,on the topics of accelerator-based transmutation. .A

wide variety of concepts to discuss were presented together with

presentations on data acquired,data needs and applicable models.

3) 2nd International Information Exchange Fleeting

(ANL,US,Sovember 1992) - .

The presented papers indicated that there was a common

concern about the need of guidance on research needs. -4 number of

emerging important issues were identified during the meeting.

including the legal background, the incentives and the implica-

tions for the whole fuel cycle in different countries.

(2) On-Going Programs

1) 3rd International Information Exchange +leeting

(Cadarache,France,l2-14 December 1994)

At the 2nd International Information Exchange ?leeting,it ivas

concluded that a comparison of systems studies, or proto-system

studies, already in progress should form a significant part oP

the 3rd meeting in December 1994. This is regarded as a first

phase in the approach to a more co-ordinated systems stud). that

~vould seek to identify advantages and disadvantages of introduc-

i .ng the P-T to the nuclear fuel cycle. Other areas of interesL

will be the effects of progress made in various national P - I '

.-

ac~ivities,technical advances,economic a s s e s s m e n t s , objectil-es of

the P-T and environmental impacc considerations. It is currently

interested to invite participants from the Russian Federation to

this meeting.

2) Preliminary Systems Study on Partitioning and Transmutation of

Long-lived Xuclides

In order to improve understanding of the merits and draw-

backs of introducing the P-T technology into the nuclear fuel

cycle, analytic and presentational tools will be developed for

comparing the many different technical approaches in relation to

a 1-arietv of measures of cost and benefit. This activity will be

intended to run for 3 years. In 1998, a state of the art repor;

will be submitted, which will identify advantages and disadx-an-

tages, key technologies, target capabilities of the P-T technolo-

gies.

4 . PROPOSED TRANSMLTTATION CONCEPTS

Although many different types of nuclear particles are

available to transmute waste isotopes, neutron reactions are z o s ~

effective to transmuxe them from the energy balance poinc of

view. Therefore, there have been many studies of IransmutaTion in

neutron fields generated by a variety of devices,including cher-

ma1 reactors, fast r e a c t o r s - . ; a c c e l e r a t o r s , a n d fusion reactors.

4 . 1 Thermal Reactors

Therma.1 neutrons have high reaction cross-sections with rhe

actinide isotopes than fast neutrons. In thermal reactors,trzns-

mutation by neutron capture is the dominant process for the long

lived actinide isotopes, such as Np-237.Am-241 and .AK-243. These

isotopes are incinerated in an indirect n-ay by formation of a

heavier decay product with very high fission cross-seciion, such

as Pu-236,Pu-239,Arn-242m,Cm-243 and Crn-24.5.

Introduction of minor actinides with high neutron capiure

cross-section into reactor core will decease the reactivity and

require higher fissile enrichment. During the life of the reactor

core, the reactivity may increase with the formation of fissile

nuclides. This means that most of the minor actinides play a role

as burnable poison. The presence of non-negligeable rare earths

co-extracted with Am and Cm i.n the separation process reduces r ; he

reactivity. Separation of rare earths Prom Am and Cm should be

needed from the neutron economy point of view.

Two types of minor actinide recycling n~odes can be consid-

ered: homogeneous and heterogeneous recycling. The l a t ~ e r could

rransmute roughly the same amount of minor actinides as cne

homogeneous recycling from reactor physics point of, and nini-

mize the number of fuel pin and assembly containing minor acti-

nides. However, the thermal characteristics of fuel assemblies

loaded with high concentration of minor actinides require the use

of specially designed fuel pins and assemblies. In this case,

strong neutron self-shielding may lead to significant reduction

of transmutation.

The long-lived fission products of interest, such as Tc-99

and I-129,can be transmuted to stable isotopes by neutron caprure

reaction. The thermal neutron cross-sections of these nuclides

are bigger than the fast ones,but are not big enough to transmute

them effectively. This leads to need of high-flus thermal reac~or

for effective transmuxation of Tc-99 and 1-129.

(1) PWR based transmutation concept proposed by the CEA [14]

Two types of transmutation concepts based on PKR are pro-

posed by the CEA. They are 3IOX felled PWRs with homogeneous and

heterogeneous loading of minor actinides. In the latter, volume

ratio of fuels with transmutation materials to all fuels is 30%.

Kuclides to be transmuted in the reactors are Np-237.L4m-241.L-\m-

242 and Am-243.

The principal design considerations are paid for foll.o\vings:

- highest transmutation rate and radiotosicity reduction.

- kse of well known current reactor technology with 1ittl.e

development.and

-

- Same safety and performance levels as current reactors.

The reaczor design and physics parameters Ere gil-en in Tzble

4.1 and 4.2 for homogeneous loading, and Table 4.3 and 4.4 for

heterogeneous loading,respectively.

(2) PWR based transmutation concept proposed by JAERI [I51

JAERI proposes a conc-ept 'io transmute sp-237,whicn is bzsed

on the existing P\YR with much higher enriched uranium (8.2 ;v/o)

fuel. The transmutation target nuclide Np-237 is ho~ggeneouslv

mixed with uranium fuel.

The principal design considerations are paici for follo~vings:

- The transmuration target is limited only to &p-237 \vhich is rhe

most importany nuclide with long half-life and high raaio~osizi~y

generated in uranium-oxide fuel, and

- The Np transmutation should be made by the esisting P\:'F1 n-i~hout

any significant modification.

The reaccor design and physics parameters is given in Tzble

4.4 and 4 . 3 .

( 3 ) PBR based burner concept proposed by BNL [ 6 ]

A high flus particle bed reactor concept (PBR) for: rapid

transmutation of actinides and long-1-ived fission products is

proposed by BSL. This concept is based on the PBR nuclear- rocket;

system currently under development by the Air Fot-ce Space Suclcar

Thermal Propulsion (SSTP) Program and draws on much of he tech-

nology that has been developed by the SNTP prograrn.

The basic building of t h e PBR are the fuel particles ~ h i c h

are similar to the proven HTGR BISO. Two types of particles ;sould

be employed: one containing plutonium for rhe "dril-er" fuel

elements,and one containing minor actinides for the "target". The

fuel elements consist of a bed made up of the appropriate perxi-

cles, and constrained between two porous, co-axial cylindrical

"frits"(see Fig.3.1).

Several potential core designs were e v a l u a ~ e d , falling into

two general categories: heavy water moderated,and solid moderator

systems. The major constraints for the core design are the desire

to minimize the Pu/minor-actinides in\-entory in the reactor, and

the overall reactor size: therefore the total number of elements,

the particle fissile and minor actinide loading ,and the r-dial

reflector are minimized.

The selected characteristics of heal-y waxer moderated PBR

burner concept are g-il-en in Table 4.7.

4.2 Fast Reactors

Fast neutrons n'ill fission all of the actinides, but the

reaction cross-sections are much smalle1- than for thermal neu-

trons. This effect will be compensated by the neutron flus in the

fast reactors, which is 100 to 1.000 times higher than in the

thermal reactors. The build-up of higher actinides by neutron

capture is much smaller with fasL neutrons Khan cherrnal neutrons.

The fission to capture ratio of minor actinides increases ~vith

the mean neutron energy. If a fast reactor concept with very hard

- neutron spectrum is established,the minor actinides Ere addition-

al fissionable resources instead of waste materials as in the

thermal reacror. However,loading amount of minor actinides may b?

seriously limited, because of their unfavorable characteristics

to safety physics parameters, such as coolant-void coefficient.

Doppler coefficient, effective delayed neutron fraction and

prompt neutron life time.

There are two types of minor actinide recycling modes also

for fast reactors: homogeneous and heterogeneous recycling. The

feature is similar to that of the thermal reactors, e x c e p ~ lo\ver

neutron self-shielding in the fast reactors.

The long-lived fission products such as Tc-99 and 1-129 can

also be transmuted in the fast reactors, using neutrons with

appropriate energy.

(1.) LMFBR based transmutation concept proposed by CEA [ I 4 1

Two types of 1430 ?IWe LYFBR based transmutation concepts are

proposed b>- CEA: one with homogeneous loading of minor, and one

with heterogeneous loading of separate Np and .Am targets. The

nuclides to be transmuted are Sp-237 and Am-241 for the first

concept,and Np-237 and Am isotopes.

The Principal design considerations are paid for followings:

- Highest transmutation rate and radiotoxicity reduction,

- Use of well known current LhlFBR technol.ogy with little

development,and

- Same safety and performance levels as conventional L>lFBRs.

The principal design and physics parameters are l i s ~ e d in

T a b l e 4.8 and 4.9 for homogeneous loading of minor actinides, and

Table 4.10 and 4.11 for heterogeneous loading,respectively.

( 2 ) LMFBR based transmutation concept proposed by PNC [ 1 6 ]

PNC has investigated various kinds of LPIFBR based Transnuta-

tion concepts. The present proposal is one of these coEzepts ~vith

homogeneous loading 04 minor acrinides. The nuclides to be trans-

muted are Np-237,Am-241 and ---243.

The design principle is to develop a LXIFBR corz concept

loaded with minor acTinides which brings no serious issxe to core

performances in consideration of fuel cycle techr~ology.

The principal dzsign and physics parameters ar? gir-en in

Table 4.12 and 4.13.

( 3 ) ALMR based transmutation concept proposed by GE and ANL [ I T ]

The ALblR plant utilizes sis reactor modules. The thermal

rating of each module is 840 >l\Vt. Conventional AL?IR core designs

utilize a radially heterogeneous configuration; he inclusion of

internal blanket zones allows fuel cycle operation in a 'break-

even' mode where the fissile material (transuranics, primat-ily

Pu-239 and Pu-241) is consumed and destroyed at rouzhly equal.

rates. The 840 ?l\Vt breakeven core has a total of 192 fueled

assemblies (105 drivers and S4 blankets). The drix.er fuel form is

metal fuel alloy (U-TRL-lO%Zr). PIinor actinides are included in

the source LWR spent fuel (10.7% MA/TRU).

Two burner configurations are presented. A primar:: goal in

- --

developing the burner core configurations is to mainrain conpeci-

bility \vith the breakeven reactor design; design changes 10 the

conventional reactor are to be minimized. Ket consumption of

transuranics in the burner designs is achieved by removing fer-

tile material from the breakeven configuration.

The Core layouts of the burner designs are shon-n in Fig.4.2

and the neutronics parameters are given in Table 4.14.

(4) LMFBR based transmutation concept proposed by CRIEPI [ I S ]

CRIEPI proposes a 1000 .'I'lX;e L N F B R based transmutation concept

of minor actinides. The fuel is metallic alloy type with minor

actinides (Np,.lm,Cm) of 5 w/o. The design principle is to de\-sop

600-1000 ?]\Ye commercial F B R with U-Pu-$LA-Zr fuel. which is pro-

duced from dry process n,ith pprochemical partitioning. The pro-

posed 1000 $l\Ve F B R is expected to transmute minor actinides

generated from 6 plants of 1000 >l\Ye L K R .

The principal design and physics parameters are lisced in

Table 4.13 and 4.16.

( 5 ) LMFBR based transmutation concept proposed by Toshiba Corpo-

ration 191

The transmutation concept based on 600 WC'e L3IFBR ivith flat

core is proposed by Toshiba Corporation. The fuel. is metallic

alloy type of 3 w/o minor actinides. In Lhe core, the minor

actinides mixed fuels are homogeneously arranged, and in the

blanket,the fuels are heterogeneously arranged. This arrangement

. -.

i s e f f e c t i v e TO reduce 4'a vo id reac i ix - ixy e f f e c t .

The principal design considera~ions arz paid for folloivings:

- Homogeneous TRU-recycling called "actinide recycling" ~vithout

separation of minor actinides from Pu in LWR and FBR spent fuel

- Safety consideration,and

- Excess neutron utilization for balancing breeding capacity,

storaged minor actinides transmutation and long-lived fission

products incineration. ..

The principal design and physics parameters are listed in

Table 4.17 and 4.18.

( 6 ) Fast burner reactors proposed by JAERI

Three types of fast burner reactors with nitride fuel for

TRU transmutation are proposed by J A E R I .

The first is a helium-cooled fast reactor wish coated parti-

cle fuel (P-.lBR:see Fig.4.3) and the second is a lead-cooled

modular type fast reactor ivith pin type fuel (L- .ABR) [19].

The both reactors ha\-e the core n.ith very hard neutron energy

spectrum,in which much of the minor actinide transmutation occurs

by fission reaction, not by neutron capture.

The principal design considerations for these two t-eactors

are paid for followings:

- Minor actinides burner reactor specially designed for efficient

fissioning of minor actinides,

- Primary fuel material of minor actinides.and enriched uranium

or plutonium,

- .A fast reactor with very hard neutron spectrun~ and high neutron

flux.

Lse of enriched uranium instead of plutonium increases effecsil-e

delayed neutron fraction of the reactor.

The principal design and physics parameters are gix-en in

Table 4.19 and 4.20 for the P--ABR,and Table 4.21 and 4.22 for the

L--ABR.

The third concept is a lead-cooled fast reactor nicn ni~ride

fuel of Th-Pu-lOw/o;LLA [IS]. In this reactor, minor actinides can

be incinerated simultaneously by burning excess pluroniua in

a closed fast reactor fuel cycle.

The principal design and physics parameters are g l T . - e n in

Table 1.23 and 4.24.

4.3 Accelerator Driven Transmutation Systems

These systems use spallation reaction to produce.high ensrgj-

particles (e.g.1 GeV protons), a large amount of neutrons chzs in

a second moment are introduced in a multiplying medium.

Spallation, a reaction in which a high-energy primary parti-

cle interacts with a target nucleus, is thought to take place in

two stages. In the first stage (the intranuclear cascade phase).

the incident proton creates a high energy particle cascade inside

the nucleus. During the intranuclear cascade,high-energ~ (>20>ieV)

"secondary" particles and loiv-energy ( <20 >lev) "cascade" par.r;i.cLes

escape the nucleus; at the end the nucleus is typically left in a

highly excited state. In the second stage (the evaporation phase)

, the esci ted nucleus relases , primarily by emitting loiv-energy

. -

(<20 &lev) " e v a p o r a t i o n " n e u t r o n s . I t i s d e f i n e d l o w - e n e r g -

" s p a l l a t i o n " n e u t r o n t h e sum o f r h e l ow-ene rgy c a s c a d e 2nd ex-apo-

r a t i o n n e u t r o n s .

F o r t h i c k t a r g e t s , t h e h i g h - e n e r g y s e c o n d a r y p a r t i c l e s ( p l u s

t h e i r p r o g e n y ) c a n unde rgo f u r t h e r s p a l l a t i o n r e a c t i o n s . F o r some

t a r g e t m a t e r i a l s , lorn-energy s p a l l a t i o n n e u t r o n s c a n e n h z 2 c e

n e u t r o n p r o d u c t i o n t h r o u g h low-energy ( n , x n ) r e a c t i o n s . The t o z a l

l ow-ene rgy n e u t r o n p r o d u c t i o n f rom a t a r g e t i s t h e sum o f l o x -

e n e r g y s p a l l a t i o n n e u t r o n p r o d u c t i o n p l u s t h e n e t p r o d u c ~ i o n f r o m

low-ene rgy ( n , x n ) r e a c t i o n s .

U s i n g a c a l c u l a t i o n code a s a model wh ich i s a p p r o x i m a t e l y

c o r r e c t , a b o u t 90% o f t h e n e u t r o n s coming f rom a c y l i n d r i c a l

t a r g e t bombarded b y 1 GeV p r o t o n s , have l e s s t h a n 20 ?Iel- wi t 'h a n

a v e r a g e e n e r g y o f o n l y 4 . 8 PIei'; r e m a i n d e r , 1 0 % o f t h e t o r a l , hrx-e

e n e r g i e s below 400 Me\; w i t h a n a v e r a g e of 1 0 5 MeV.

(1) A c c e l e r a t o r - d r i v e n s y s t e m p r o p o s e d by J A E R I [ 2 0 ]

T h r e e t y p e s o f a c c e l e r a t o r - d r i v e n s y s t e m a r e p r o p o s e d b y

J A E R I , e a c h o f which u s e s a l a r g e l i n e a r p r o t o n a c c e l e r a t o r r o

d r i v e and c o n t r o l i t s s p e c i f i c s u b c r i t i c a l c o r e c o n t a i n i n g mino r

a c t i n i d e s and o t h e r l o n g - l i v e d n u c l i d e s .

The f i r s t c o n c e p t c a l l e d ".Alloy f u e l e d c o r e s y s t e m " i s shown

i n F i g . 4 . 4 . I n t h i s c o n c e p t . t h e a c c e l e r a t o r i n j e c t s 1 . 5 GeV

p r o t o n beam o f 3 9 rnA i n t o t h e t u n g s t e n t a r g e t l o c a t e d a t i h e

c e n t e r o f l h e sod ium-coo led f a s t r e a c t o r c o r e , {vhictl i s l o a d e d

w i t h a l l o y f u e l c o n t a i n i n g minor a c t i n i d e s . The p r i n c i p z l d e s i g n

p a r a m e t e r s a r e g i v e n i n T a b l e 4 . 2 5 .

-7-

The second is called "?Iolten salt core" concept. The accel-

erator injects 1.5 GeV proton beam of 25 mA in the core wirh hard

neutron spectrum, which is loaded with chloride molten salt fuel

containing minor actinides. This concept would be a continuous

processing system,in which the reaction products are removed from

the fuel on line. The principal design parameters are lisxed in

Table 4.26.

The third is called "Eutectic target-core" concept. The

accelerator injects 1.3 GeV proton beam of 30 m\ into the eutec-

tic alloy (Np-Pu-Co-Ce-Tc) target-core with graphite blanket,

which is cooled by molten fluoride salt. This system intends

simultaneous transmutation of 3 p - 2 3 7 and Tc-99. The design parrm-

eters are given in Table 4.27.

(2) Accelerator-driven systems proposed by BNL

Three types of transmutation concepts are proposed by BSL:

one with a large linear proton accelerator [5], and two with a

small power accelerator [21].

The first is the PHOENIX concept using a large linear procon

accelerator which can produce a 104 m.A beam of 1.6 GeV protons. .A

modular concept is developed for the PHOENIX subcritical lattice.

Each module resembles the core of the Fast Flux Test Facility (FF

TF), with the minor actinides,formed into oxide fuel rods replac-

ing the uranium and plutonium in the FFTF fuel. The concept is

shown in Fig.4.S.

The second is the "MOS fueled core" concept,in which a multi

segmented cyclotron in.jects 2.5-3 mA beam of 1.5 GeV protons into

the lead target located at the center of >lox fueled fzst reactor

core which operates at sligh~ly subcritical condition. .'Iinor

actinides are incinerated by fast neutron fission reaction, and

long-lived fission products such as Tc-99 and 1-129 are t.ransmut-

ed in yttrium-hydride moderator surrounding the core. The design

parameters are given in Table 4.28.

The third is the "Particle fueled core" concepts. This uses

a multi-segmented cyclotron to inject 4-8 rn-4 beam of 1.5 GeV

protons into the lead spallation target located at he center

of the core nith fast neutron spectrum, which is loaded nith

nitride coated particle fuel. The core operates 21 slightly

subcritical condition. The long-lived fission products can also

be transmuted as in the second concept. The principal design

parameters are given in Table 4.29.

(3) Accelerator-driven systems proposed by CEA [14]

Three types of transmutation systems are under investigation

at the CE.4 to mainly incinerate Tc-99. Each system use the

accelerator \vith the proton beam of 1.3 GeV and 70-100 m4 to

strike the lead chloride molten salt target containing Th, Th-Pu

or Pu. The design parameters and mass balances of the systems are

given in Table 4.30.

(4 ) Los Alamos ATW concepts

Contrary to the above mentioned accelerator-driven transmu-

tation concepts in which the transmutation occurs mainly by fast

neutron, the AT\V concepts uses the thermal neutrons to transmute

minor actinides and long-lived fission products.

In the ATW concept,the linear proton accelerator operaces aT

1.6 GeV at a continuous--wave current of 250 mA. The primary

proton beam is then split into four beams, each having a current

of 62.3 mA. Each of the four beams directed into four seperated

target/blanket modules. The high-energy proton beam strikes a

centrally located spallation target to produce an intense source

of neutrons.

The base-case design is comprised of heavy-water-cooled

tungsten rods, and its blanket region and balance-of-plant design

is based on existing heavy-water reactor technology employed in

the CANDLL reactor system [22]. Another option is the use of a

flowing lead target. The use of such a target adds complexit>-

to the design but has the potential to increase the neutron

utilization efficiency. The layout for the target/blanket of the

ATlY is shown in Fig.4.6, and the key design parameters are

compared with those of the CANDU in Table 4.31.

An advanced AT\Y concept is also proposed,~hich has a target-

blanket with sloi~ly circulating higher actinide liquid fuel and

heat removal by a larger, thorium-bearing rrlolten salt loop [ 2 3 1 .

This concept is shown in Fig. 4.7.

( 4 ) Other p r o p o s e d systems

There are some XTW type concepts proposed by the ESE-A [ S ] .

the Royal Institute of Technology [lo] and the ITEP [I?]. In che

ENE.4 concept, a subcritical core with lead spallation carget is

driven by a proton accelerator with the beam of 1.6 Gei- and 200

A . The core is loaded with minor actinides and long-1i.i-ed

fission products. In the- concept of the Royal Ins~irute of

Technology, a subcritical core with lithium fluoride salr

(liquid Pb, solid Th 1 spallation target is driven b3- a proron

(deutron) linear accelerator or a cyclotron with the beam of 1

GeV and 5-100 mA. The core is cooled by helium and is loaded

with minor actinide molten salt fuel or graphite pebble bed

particle fuel.

The design parameters of the concepts proposed by 3 organi-

zations are given in Table 4 . 3 2 , 4 . 3 3 and 4.34, respecti\-ely.

Fig. 4-1 A Fuel Element for Pressure Vessel Configuration of P B R

SmdI Burner Large Eiurner

F i g . 4 . 2 c o r e C o n f i w r a t i o n of ALMf: f o r A c t i n d e s B u r n i n g

. . . ? . .? Cross s?ciion

' U S ! t l e a e n t

F i g . 4 . 3 Fuel Concept of H e l i u m - C o o l e d Actinides Burner (L'-mt:)

F i g - 4 . 4 Schematic Figure of Acce le ra tor - l ) r i .veI1 ' l ' r ansmuta t ion Concept - A l 1 . o ~ F u e l Core Sys tem -

Table 4.1 Principal Design Parameters of PR Based Transmutation Concept Proposed by CEA - 3oaogeneous Loading of Minor Act inides - [I41

Items

Thermal power (MWt) Electric power (MWe) Equivalent core diameter (cm) Equivalent core height (cm) Averaged core composition (v/o) fuel/clad+structure/coolant

Type of fuel Core averaged fresh fuel composition (w/o) U/Pu/Np, A m and Cm Isotopic composition of Pu in fresh fuel (w/o)

23. 2/7. 45/69. 35 oxide

Table 4 . 2 Generic and Safety-Belated Physics Parametes of the Proposed Concept Given in Table 4.1

2 iapU/l;spU/2<vu/2< ~ ~ ~ , , 2 . : 2 ~ ~

Isotopic composision of minor actinides (w/o) 2 3 7 hp/24 . l ~ ~ / ' & ' ~ ~ / ' < ' ~ ~ / 2 ~ 3 ~ ~ / 2 ~ < ~ ~

- Power density: average in core (w/cm3) maximum in core

- Linear heat rate: average (w/cn) nay imum

- ~ e u t r o n flux averaged in core (n/cm2 sec) - Neutron energy averaged in core (KeV) - Fuel dwelling time in core (EFPD) " - Burnup reactivity swing (%6k/k/365EFPD) - Coolant void reactivity effect (%8 k/k)

All reactor region voided Core and axial blanket voided Only core voided

- Doppler reactivity coefficient (Tdk/dt)

- Effective delayed neutron fraction - Control rod material - Central control rod worth ( % 6 k/k/kg)

1. 8/58. 0/22. 5/12. 0/5. 7

55. 7/27. 8/0 07/16. 4/0/0 ~

1) Begining of Life. 2) Begining of Equilibrium Cycle. 3) Equivalent full power days

Table 5.2 Material inventory and mass balance of Np Target in PWR Based

Transmutation Concept Proposed by CEA - Heterogeneous Loading -

Table 4 . 3 Principal Design Parameters of PWB Based Transmutation Concept Proposed by CEA - Heterogeneous Loading of Np and Am Target - [I41

Items I Thermal power Electric power Equivalent core diameter Equivalent core height Averaged core composition fuel/cladtstructure/cooiant

Type of fuel ?ip target composition u/Pu/"'x~

.Am target compos i t ion U/Pu/Am Isotopic composision of .An 2 2 lAn/2'2Am/243AAm

30. 0/10. 8/59. 2 oxide

Table 4.4 Generic and Safety-Belated Physics Parametes of the Proposed Concept Given in Table 4 . 3

1) Begining of Life. 2) Begining of EquiIibrium Cycle. 3) Equivalent full power days

i tems

- Power density: average in core (w/cm3) maximum in core

- Linear heat rate: average (w/cm) maximun

- Neutron flux averaged in core (n/cn2 sec) - Neutron energy averaged in core (Key) - Fuel dwelling time in core (EFPD) " - Burnup reactivity swing (%6 k/k/365EFPD) - C o o l a n t v o i d r e a c t i v i t y e f f e c t (%6k/k)

All reactor region voided Core and axial blanket voided Only core voided

- Doppler reactivity coefficient (Tdk/dt) - Effective delayed neutron fraction - Control rod material - Central control rod worth (% 6 k/k/kg)

BOP 1 BOEC':

105 245 178 419

3. 0 x 10'' . -

1120 11

-

-

-

-

Ag, In. Cd

-

-

-

-

-

-

-

-

-

Table 4.5 ' Principal Design Parameters of PWE Based Transautation Concept Proposed b y JAEBI 1151

Thermal power (MWt) Electric ~ o w e r (ifwe) Equivalent core diamet?: (cm) Equivalent cors height (cm) Averaged core conposi t izn ( li/o) fuel/clzd+strccture/czc!znt

Type of f u e l Core averaged fresh Eui: conposition (w/o) U/Pu/Xp. 3.n azd Cn Isotopic composition of ?u in fresh fuel (w/o)

2 53pu/2?3pu/1;3 DU/2< 12. . / 2 < 2 . - Pu Isotopic composision of ?inor actinides (w/o)

2 i T ~ ~ , / 2 ; 1 . ~ ~ / 2 4 2 ~ ~ / 2 . ' ? ~ .?,. '2.' 3cm/2.'; cm

3 4 1 0 1146

3 3 7 3 6 6

3 1/11/58 oxide

Table 4.6 Generic and Safety-Related Physics Parametes of the Proposed Concept Given in Table 4 . 5

I tens - I

- Power density: average in core (w/cni) - - , . t a x inum i ~ . core

- Linear hsat rate: averaze (w/cn) iilZY i i i ~ 2

- Neutron flux ayeraged i ; core (n/cns sec) - Neutron energy averzzed in core (KeV) - Fuel dwelling time in core (EFPD) "

- Burnup reactivity swing ("mk/k/365EFPD) - Coolant void reactivit:; effect (%6 k/k)

All reactor region voided Core 2nd axial blznket voided Only core 5;oided

- Doppler react ivi ty co2ificien t (Tdk/dt) - Effective delayed neuirDn fraction

1) Begining of Life. 2) S23ining of Equilibrium Cycle. 3 ) Equivalent full powzr cays

- Control rod material - Central control rod worth (% 6 k/k/kg)

i\g-In-Cd 1 - ~ -

Table 4.7 Selected Characteristics of Heavy Water Moderated PBR Burner

! -- , . . -.-,. . .,-- .... -->--q ' -..- h "' -...-.,-+,.; .... \ .. . ! 2..- _YC I... ;, -- --., Y . ; - - ..> I . ...... !

Table 4 . 8 Principal Design Paraaeters of MOX-LMFBB Based Transmutation Concept Proposed by CEA - Homogeneous Loading of Minor Actinides - 1141

I terns

Thermal power (\iwt) Electric p o ~ e r (hiwe) Equivalent core diameter (cm) Equivalent core height (crn Averaged core composition (v/o) f uel/clad+structure/coolant (Na)

Type of fuel Core averaged fresh fuel composition (w/o) U/Pu/Np. Am and Cm Isotopic composition of Pu in fresh fuel (q/o) 23apU/233pU/240pU/2.:1Pu~2<2pU

Isotopic composision of minor actinides (w/o) 2 3 7 ~ ~ / 2 4 1 ~ ~ / 2 4 2 . ~ ~ / 2 4 5Am/2~3Cm/2<< cm

3!/25/38 oxide

- Power density: average in core (w/cni) maxinum in core

- Linear heat rate: average (w/ciil ) maximum

- Neutron flux averaged in core (n/cm\ec) - Neutron energy averaged in core (Ke'i) - Fuel dwelling time in core (EFPD) " - Burnup react ivi ty swing ( X 6 k/k/365EFPD) - Coolant void reactivity effect ( X 6 k/k)

All reactor region voided Core and axial blanket voided Only core voided

- Doppler reactivity coefficient (Tdk/dt) - Effective delayed niutron fraction - Control rod material - Central control rod worth ( % 6 k/k/kg) - Prompt neutron life time (sec)

Table 4.9 Generic and Safety-Belated Physics Parametes of the Proposed Concept Given in Table 4. 8

1) Begining of Life. 2) Begining of Equilibriun Cycle. 3) Equivalent full povier days

Items i j

- %

$31" BOEC='

Table 4.10 Principal Design Paraaeters of MOX-LMFBB Based Transautat ion Concept Proposed by CEA - Beterogeneous Loading of Np and Am Targets - [I41

It ems

I Thermal power (MWt) Electric power (\!We) Equivalent core dizaeter (cm) Equivalent core heizht (cm) Averaged core composition (v/o) f uel/clad+structur~/coolant (Na)

Type of fuel Core averaged fresh fuel composition (Y/o) U/Pu/Np. Am and Cm isotopic compositio- of Pu in fresh fuel (vi/o) 25€?pU/233pU/"3pU/2'L PU/'"PU

Isotopic composision of K p and Am targets (w/o) "iNp and 211~m/'<'Jm/24'Am

Table 4.11 Generic and Safety-Related Physics Parametes of the Proposed Concept Given in Table 4 . 1 0

- Power density: average in core (w/cm5 ) mauiaua in core

- Linear heat rate: zverage (w/cm) mzsinum

- Neutron flux averzsed in core (n/cm2 sec) - Neutron energy averaged in core (KeV) - Fuel dwel ling tine in core (EFPD) " - Burnup reactivity swing (% 6 k/k/365EFPD) - Coolant void reactivity effect ( % 6 k/k)

All reactor region voided Core and axial blanket voided Only core voided

- Doppler reactivity coefficient (Tdk/dt) - Effective delayed neutron fraction - Control rod materizl - Central controi rod worth (% 6 k/k/kg) - Prompt neutron lif? time (sec)

1) Begining of Life, 7 ) Begining of Equilibrium Cycle. 3) Equivalent full poqer days

Table 4.12 Principal Design Parameters of MOX-LMFBB Based Transmutation Concept Proposed by PNC [I61

Thermal power (hIWt) Electric power (hiwe) Equivalent core diameter (cm) Equivalent core height (cm) Averaged core composition ( V/O) f uel/clad+structure/coolant(8a)

Type of fuel Core averaged fresh fuel composition (w/o) U/Pu/Np, Am and Cm Isotopic composition of Pu in fresh fuel (w/o)

2 ~ a p u ~ 2 i s p u ~ 2 ~ ~ u ~ 2 4 1 T U ~ 2 ~ 2 P u

lsotopic composision of minor actinides (w/o) 2 3 7 y . I D /2< L ~ m / 2 C 2 A m / 2 4 3 A m / 3 - . 1 3 C m / 2 4 r . cm/245Cm

41 . 6/20. 9/37. 5 oxide

Table 4 . 1 3 Gene~ic and Safety-Belated Physics Parametes of the Proposed Concept Given in Table 4.12

- Povier density: average in core (w/cni ) naximum in core

- Linear heat rate: average (w/cm) maximum

- Neutron flux averaged in core (n/cm2 sec) - Neutron energy averaged in core (XeV) - Fuel dwelling time in core (EFPD) - Burnup reactivity swing (%6 k/k/365EFPD) - Coolant void reactivity effect (Z6 k / k )

A.11 reactor region voided Core and axial blanket voided Only core voided

- Doppler reactivity coefficient (Tdk/dt) - Effective delayed neutron fraction - Control rod material - Central control rod worth (% 6 k/k) - Prompt neutron life time (sec)

1) Begining of Life, 2) Begining of Equilibrium Cycle. 3) Equivalent full power days

Table 4.14 Neutronics Results of ALNR Actinide Recycling

C~ie Hzi$i (i~. ) 12 2 6 Core Diziierz (m) 14 1 14 !

= o f f ucl As? 108 192 = oiEI!d<et ?.ssy 84 Conversion R a i o 1.06 0.72.

Cycle Lengih jrnon+s) 2 3 i 2

Burnu? 2eac;ivi;ir). SLLln? I S ) 0.57 8.99 ?--< L m a - Power 9.5 10.4

S d i u i Void %'orin. 6.2 -2.5 ( ~ 7 % L7 G-mU-zr) - 7 1 1 ?/? 3

i K U Livzn?orl.. (k9'~072) 268 1 Iii; -- - lit5 Cons~x? i ion 2.2~2

kg'>.-.- -2s.z. b.i -,- - - O/o :1;.=n:o-;'\.e-- - -

. . - i . l 3 . 2

Table 4.15 Principal Design Parameters of Metal-Fuel-LMFBB Based Transmutation Concept Proposed by CBIEPI 1181

i I Thermal power (&!Wt) 2632 1 Electric power Equivalent core diameter Equiva!ent core height Averaged core com?osition -, (v/o) f uel/cladtstructure/coolant (Xa)

Type of fuel Core averaged fresh fuei composition (a/o)

U/Pu/?ip, .Am and Cm/rare earths isotopic composition of Pu in fresh fuel (w/o)

2 ;apU/2;spu/2<:ip,,,z< I Pu/z-*.2PU

Isotopic composision of minor act inides (vi/o) 2 3ihp/2-^. 1 Am/X-'Am/2< 3Am/24 SCm/254 cm

Table 4.16 Generic and Safety-Belated Physics Parametes of the Proposed Concept Given in Table 4. 15

i terns 1 B O L ~ '

- Power density: average in core (w/cnS ) mzxinum in core

- Linear heat rate: average (w/cn) max imuiil

- aeutron flux zveraged in core (n/cn2 sec) - Xeutron energy averaged in core (KeV) - Fuel dwelling time in core (EFPD) ' )

- Burnup reactivity swing (% 6 k/k/365EFPD) - Coolant void reactivity effect (%6 k/k)

All reactor region voided Core and axial blanket voided Only core voided

- Doppler reactivity coefficient (Tdk/dt) - Effective delayed neutron fraction - Control rod material - Central control rod worth ( % a k/k/lO~ kg) - Prompt neutron life time (sec) - Axial thermal expansion (6 k/k)/(6 L/L)

1) Begining of Life. 2) Begining of Equilibrium Cycle, 3 ) Equivalent full power days

Table 4.17 Principal Design Parameters of Metal-Fuel-LMFBR Based Transmutation Concept Proposed by Toshi ba Corporation [9]

Thermal power Electric power Equivalent core dimeter Equivalent core height Averaged core composition -,.

f uel/clad+structur~/coolant (Na) Type of fuel Core averaged fresh fuel composition U/Pu/?ip. Am and Cm Isotopic composition of Pu in fresh fuel 238pU/'39pU/243pU/'~LpU/242pu

Isotopic composision of minor actinides 2 3 7 ~ p / 2 4 ldm/2Q2Am/2LRAm/243Cm/24Lcm

38. 5/35. 9/25. 6 metal 1 i c

Table 4.18 Generic and Safety-Related Physics Parametes of the Proposed Concept Given in Table 4.17

1) Begining o f Life, '2) Equilibrium Cycle, 3) Equivalent full power days

items

- Power density: average in core (w/cm3 ) aaximurn in core

- Linear heat rate: zlierage (w/cm) ma:< imum

- Neutron flux averzged in core (n/cm2 sec) - Xeutron energy averaged in core (Kell) - Fuel dwelling time in core (EFPD) " ' - Burnup reactivity swing (% 6 k/k/355EFPD) - Coolant void reactivity effect (X6 k/k)

All reactor region voided Core and axial blanket voided Only core voided

-Dopplerreac:ivitycoefficient (Tdk/dt) - Effective delayed neutron fraction - Control rod materizl - Central control rod worth (% 6 k/k/kg) - Prompt neutron life time (sec)

BOL" I 3 ~ 3

350

4 9 5

1 0 9 5 4 . d

-0. 3 5 - 0. 00 0.25 - 0.60

-1.7 :i lo-; 0. 0035

3; C -

2.4 .u lo-'

Table 4.19 Principal Reactor Design Parameters of Helium-Cooled Actinide Burner Concept (P-ABR) Proposed by JAEBI [19]

Items

Thermal power Electric power Equivalent core diameter Equivalent core height Averaged core composition -,

f uel/clad+structure/coolant (He) Type of fuel Core averaged fresh fuel composition ti/Pu/Xp, Am and Cm

Isotopic composition of U in fresh fuel 2 35L/2i5ti/23d ti

Isotopic composision of minor actinides 2 3 TNp/2< 1 Am,/" ZAm,,/2< "Am,/' "cm,/2&& Cm

22. 1/25. 9/5 2. 0 nitride

Table 4.20 Generic and Safety-Belated Physics Parametes of the Proposed Concept Given in Table 4.19

- Power density: average in core (w/cm; )

maximum in core - Linear heat rate: average (w/c~ )

mzsimum - Neutron flux averaged in core (n/cm%ecc) - Neutron energy averaged in core (KeV) - Fuel dwelling time in core (EFPD) " - Burnup reactivity swing (f 6 k/k/365EFPD) - Coolant void reactivity effect ("m k/k)

A11 reactor region voided Core and axial blanket voided Only core voided

- Doppler reactivity coefficient (Tdk/dt) - Effective delayed neutron fraction - Control rod materizl - Central control rod worth ( X 6 k/k) - Prompt neutron life time (sec)

YO! 1090

5.9 x !0" 'i 0 0 3 0 0 7 . 96

-

1) Begining of Life. 2 ) Begining of Equilibrium Cycle, 3) Equivalent full gower days

Table 4.21 Principal Design Parameters of Lead-Cooled Actinide Burner Concept Proposed by JAERI 1191

Items 1 Thermal power (MWt) Electric power (MWe) Equivalent Cora diaaeter (cm) Equivalent core height (cm) Averaged core coaposition ... (v/o) fuel/cl~d+structure/coolant(Pb)

Type of fuel Core zveraged f r z s h fuel composition (w/o) U/Pu/Np. Am 2nd Cn isotopic composition of L in fresh fuel (w/o) 2 i5u/25Su/23.?L

isotopic conposisioa of minor actinides (w/o) 2 >F!ip/'< LAm/ '< 'AAL3/ '<?Am/2<3cm/2<< Cm

28. 9/10. 3/60. 8 nitride

Table 4 . 2 2 Generic and Safety-Related Physics Parametes of the Proposed Concept Given in Table 4 . 21

1 teas 1 B O L ~ ' I 6 0 E C 2 : I

- Power density: sverage in core (w/cmS) flzxi~um in core

- Linear heat rate: average (w/cm nzx imum

- Neutron flux zber~ged in core (n/cm2 sec) - Neutron energy averaged in core (KeV) - Fuel dwelling tine in core (EFPD) " - Burnup react ivi ty swing (%ti k/k/365EFPD) - Coolant void rezciivity effect (%6 k/k)

All reactor region voided Core and axial blanket voided Only core voided

- Doppler reactivity coefficient (Tdk/dt) - Effective delayed neutron fraction - Control rod aaterizl - Central control rod worth (% 6 k/k/kg) - Prompt neutron !if? time (sec)

1) Segining of Life. 2 ) Begining of Equilibrium Cycle. 3) Equivalent full power days

Table 4.23 Principal Design Parameters of Th-Loaded Lead-Cooled Fast Reactor Based Transmutat ion Concept Proposed by JAERI [I51

I t ems I Thermal poqer (MWt) Electric power (MWe) Equivalent core diameter (cm) Equivalent core height (cm) Averaged core composition (v/o), fuel/cladtstructure/coolant (Pb) ,

Type of fuel Core averaged fresh fuel composition (w/o) Th/Pu/&p, An and Cm Isotopic composition of Pu in fresh fuel (w/o)

30/10/60 nitride

Table 4 . 2 4 Generic and Safety-Related Physics Parametes of the Proposed Concept Given in Table 4. 23

2 38pu/213puj243 "/"'pU/'"'pu Isotopic com?osision of minor actinides (w/o)

2 3ihp/3; L A m , / 2 : 2 ~ 4 n / 2 4 i ~ m / 2 . : 3 C m / 2 - " . c crn~2-15crn

- Power density: average in core (w/cm3) azsinum in core

- Linezr hezi rzte: averzge (w/cm) naximum

- Xeutron flus averaged in core (n/ca2 sec) - Neutron energy averaged in core (KeV) - Fuel dwel ling tine in core (EFPD) " - Burnup rezct ivi t y swing ( % 6 k/k/365EFPD) - Coolant void reactivity effect ( Z 6 k / k )

All rezctor region voided Core and zxia! blanket voided Only core voidad

- Doppler reactivity coefficient (Tdk/dt) - Effective delayed neutron fraction - Control rod aaterial - Centrzl control rod worth (% 6 k/k/kg) - Prompt neutron life time (sec)

I

0/58/2?/14/4 I

59/28/0. 1/10/0. 2/2. 6/0. 1 1 1 I

1) Begining of Life. ? ) Eegining of Equilibrium Cycle. 3) Equivalent ful! power days

Table 4.25 System Design Parameters of Accelerator-Driven Transmutation Concept

Proposed by JAEBI - Alloy Fuel Core System - [20]

I terns

Accelerator

-Type -Particle

Target -Equivalent diaaetsr -Equivarent height

-Target material -Cooling material

Subcriticzl core -Equivalent diameter

(MeV) (rn!,)

-Equivalent height (tin) -Material coaposition (v/o)

Fuel/Target/C!ad+Structure/Coolant -Cheaical for7 oi fuel -h!atsrials of coo!;nt and moderator

-Averaged fresh ftisl composition (w/o) G/Pu/Np, .Am. Cin/LL?P

-Isotopic composition of Pu (%/o) 23dpU/23YpU/2::1pU/2<lpU/242pu

-Isotopic composition of !TI.\ (w/o) 2 3 7 & p / ' " A n / 2 " > ~ z / 2 " ~ m / 2 . z ' ~ m / 2 ' 5 ~ i i l

-Averaged conpos i t ion of Long-1 i ved FP (w/o)

"TC/'" I System Characteristics -Effective multiplication factor: k , r t

-Thermal Power in core (MW t )

-Power density: average (W/cm3) mas imun (W/cm5)

-Linear heat rate: average (W/cm) mas imum (W/ciil)

-Neutron flus averaged in core (n/cm2 s )

-Neutron energy averaged in core ( K e V) -Fuel dwelling ti;? (EFPD) " -Target dwelling tiiile (EFPD) -Eurnup reaciivit? swing ( % B k/k/"sSEFPD)

Proton L I X A C Proton

3. 5/1. 7/80. 7/14. 0/0 metal 1 ic >!a / -

1) Equivalent Full Tower Day

Table 4.26 Systea Design Parameters of Accelerator-Driven Transmutation Concept

Proposed by JAEBI - Molten Salt Core System - 1203

I tems

Accelerator

-Type -Particle

Target -Equivalent diameter -Equivarent height

-Target material -Cooling material

Subcriiical core -Equivalent diameter -Equivalent height

(MeV) (mA)

-Material conposition (v/o) Fuel/Clad+Structure/Coolant/~loderator

-Cheaical forn of fuel -kfaterials of coolant and moderator

-Averzged fresh fuel composition (w/o) ti/Pu/Np, Am. Cm/LLFP

-Isotopic composition of Pu (W/O) 2 3 3 p U / 2 3 9 p U / 2 ~ O p U / 2 i l p U / 2 ~ 2 p u

-1 sotopic con~osi t ion of hi4 (w/o) 237,ip/2i 1 ~ ~ / 2 4 3 ~ ~ / 2 4 5Cm/24<cm/24 Scm

-Averaged composition of Long-lived FP (w/o)

99rc/"9 I Systen Characteristics

-Effective multiplication fzctor: k,tr -Theraal Power in core (MW t ) -Power density: average (w/cm5)

max inum (W/cmS) -Linear heat rate: average (W/cm)

nax inuiii (W/cm) -Xeution flus averaged in core (n/cn2 s) -Neutron energy averaged in core (KeV) -Fuel dwelling time (EFPD) "

-Tarset dwelling time (EFPD) -Burnup reactivity swing (3 k/k/36 5EFPD)

Proton LINAC Proton 1500

2 5

14. 5/85. 3/0/0 no!ten salt

- I

1) Equivalent Full Power Day

Table 4.27 System Design Parameters of Accelerator-Driven Transmutation Concept Proposed by JAEBI - Eutectic Metal Target-Core System - [201

I tens

Accelerator

-Type

Target -Equivalent dieaetei -Equivarent height

-Cooling material

Subcritical core -Equivalent diaaeter (cm)

-hiaterial coz?osition (c./o) Fuel/Clad+S:ructure/Coolant/~loderaior

-Ch?nical form of fuel -!,laterials of coolant and moderator

-Averaged fresh izel composition (W/O)

ti,/?u/Yp, An. Cn/LiFP

-[sotopic congosirion of Pu ( V/O ) z>dpU/z3UpU/2;3?U/2LL PU/'~'PU

-[sotopic composition of MA (w/o) 3 ; ~ ~ j z - L~~/'; :.An/2.z 394Til/2.:5Cm/244Cn

-Av?raged congosi tion of Long-1 ived FP (w/o)

9 U ~ ~ / L ' Y I System Characteristics -Effective multiplication factor: k , r :

-Thermal Power in core (hlW t) -Power density: average (W/cm3)

nax imun (W/cm3) -Lifiear heat rzte: average (W/cn)

max imun (W/cm) -Neutron flux averaged in core (n/cm2 s) -Neutron energ:; averaged in core ( K eV) -Fuel dwelling tize (EFPD) "

-Target dwelling time (EFPD) -Burnup reactivity swing (%6 k/k/365EFPD)

Proton LLNAC Proton 1500

3 0

4 0 10 0

(Np-Pu-Co-Ce-Tc) liquid fuei -

io/o/s/a 5 (Np-Pu-Co-Ce-Tc) 1 iquid fuel

LiF-BeFz / Gr~ghiie

1) Equivalent Full Power Day

Table 4.28 System Design Parameters of Accelerator-Driven Transmutation Concept PProposed by BNL - MOX Fuel Core System - [211

I I terns

Acce!erator

-Type -Particle

-Energy (MeV) -Current (mi\)

Target -Equivalent diameter (cm) -Equivarent height (cm)

-Target material -Cooling material

Subcritical core -Equivalent diameter (cm) -Equivalent ?.eight (cm)

-!,laterial coa?osition (v/o) Fuel/Clad-S~ructure/Coolant/~loderator

-Chemical for3 of fuel -ktater ials of coolant and moderator -Avoraged fresh fuel coaposition (w/o)

G/Pu/)Jp. An. Ciil/LLFP

-Isotopic co~position of ?u (w/o) 2 jdpU/2j9D,, ;113 L?u/2<2pu

1 LI

- I sotogic con?osi t ion of MA (w/o) 2 5 7;;?/2< L p r / 2 4 . 11 2 ~ ~ / 2 < 3: / 2 < 1 ~ ~ / 2 3 . 4 ~ ~ 7 m

-Averaged conpos i t ion of Long-1 ived FP (w/o)

y y ~ ~ / " y I Systen Characteristics

-Effective multiplication factor: k L c t

-Thermal Power in core (MW t )

-Power density: average (~/cm') mas inum (w/cm3)

-Linear heat rate: average (W/cm) max iiiium (W/cm)

-Neutron f lu:i averaged in core (n/cm2 s) -Neutron enersy averaged in core (KeV) -Fuel dwelling time (EFPD) "

-Target dwelling time (EFPD) -Burnup reactivity swing ( % 8 k/k/365EFPD)

hiulti-Segmented Cyclotro~ Proton

1500 2.0 - 5.0

35/24/41/0 oxide 3.2 / -

1 ) Equivalent Full Power Day

Table 4.29 System Design Parameters of Accelerator-Driven Transmutation Concept

Proposed by BNL - Particle Fuel Core System - C21l

I tems

Accelerator

-Type -Particle

-Energy (MeV) -Current (mA)

Target -Equivalent dianeter (cm) -Equivarent hai3ht (cm) -Target material -Cooling material

Subcritical cors -Equivalent diaaeter (cm) -Equivalent height (cm) -hiaterial conposition (v/o)

Fuel/Clad+Structure/Coolant/Moderator -Chemical form of fuel -Illaterials of coolant and moderator

-Averaged fresh fuel composition (w/o) U/Pu/Np. .Am, CdSLFP

-Isotopic coixposition of Pu (w/o) 238Pu/239Pu/2<3Pu/2< L ~ u /2"~u

-Isotopic conposition of IliA (w/o) 2 3 7 ~ ~ / 2 l l ~ ~ / 2 ; 2 ~ ~ / 2 l 3 ~ ~ / 2 4 ' . ~ ~ / 2 i 4 ~ ~

-Averaged conpos i t ion of Long-1 ived FP (w/o)

9 9 ~ ~ / L " 1 System Characteristics -Effective multiplication factor: kc:: -Thermal Poker in core (MW t)

-Power density: average (w/cmi) nax imum (w/cmi)

-Linear heat rate: average (W/cm) maximum (W/cm)

-Neutron flus averaged in core (n/cm2 s) -Neutron energy averaged in corz (KeV) -Fuel dwelling time (EFPD) " -Target dwelling time (EFPD) -Burnup react ivi ty swing ("m k/k/365EFPD)

htulti-Segmented Cyclotron Proton 1500 4 - 8

nitride coated particle He / Y H L :

1) Equivalent Full Power Day

7 1

- lable 4.30 Systea D e s i m ?rrrmeters o f Actolerator-Driven Concer~s froposed 3~ ZEA 11~2 -

Table 4.31 Key Design Parameter Comparison Between ATW and C - D U

E l a n k e t Arraneernent

haw wafer

2 . Fuel

Xurnber of Fuel ."isscrn'DLizs

Fuel Assembly h . k ~ r i z !

Toid i l las of FEE!

F o i x I I fuel bundle assemb!y; I flowiig fuei so!oSc:

232

Zirconium-Niobi tm

53174 kg

f ut!

250 I

I m . e

l55Okg - - I (10-d PL=T 1 ~ 2 ) 1

Same

I compacted/sinlered I aqueous ix&:idz I

Bunde Ouici Dirieiei

B~,~dics /Fu~, l Assezbly

Number of Hear-Tiznq~rt

P u n ~ s

n z u d U& pellets

1 suction. double dischxec

S O ! U ~ O ~ I ( ~ ~ ~ / I ) ; 1 (Pu. ND. Am. C n ) 1

0.1024 m

12

0.10 in i

3. Heat Transport systerr-/

3 N u m k r of S t e m ~ene i r to rd A

i w. e

Number of IK/Y Pumas 2 I

- - b

Tqble 4 . 3 1 (continuation)

13. Power I I i 1 Toid fission I iez; 1 lC0 .3 IMLV, I 1542 hDVt 1

Table 4.32 System Design Parameters of Accelerator-Driven Transmutation Concept Proposed by ENEA - ATW Type - [8]

Accelerator

-Type -Particle

-Energy -Current

Target -Equivalent diameter -Equivarent height

-Target material -Cooling material

Subcritical core -Equivalent diameter -Equivalent haight --Mat er i a 1 coa?os i t ion

Fuel/C!ad+Stru~ture/Ccoiant/hiodeiator -Chemical Cora of fuel -Materials of coolant and moderator

-Averaged fresh fuel cogposition U/Pu/Np. An. Cm/LLFP

I -Isotopic conposition of Pu 25dpU/23UpU/2'3PU/2<LPU/2.:2PU

- I sotopic co??osi t ion of !,!A 2 37>jp/21 1.AE/2< 2 ~ ~ / 2 < 3 ~ ~ 1 2 .: Icm/?<< cm

-Averaged co~position of Long-livsd FP

V 9 ~ ~ / L 2 y I System Characteristics

-Effective multi?lication factor: k,f: -Thermal Power in core (hlift)

-Power dens i ty: average (W/cm') max inun (W/cm3)

-Linear heat rate: average (W/cn) max izun (W/cn)

-Neutron flus averzged in core (n/cm2 s) -Neutron energy averaged in core (KeV) -Fuel dwelling time (EFPD) ' -Target dwel 1 ing time (EFPD) -Burnup reactivity swing ( X 6 !l/k/365EFPD)

Proton LINAC Proton I !

1600 2 0 0

2 7 200

Liquid Lead

.'~!oli.en salt or slurry

!

1) Equivalent Full Power D2y

Table 4.33 System Design Parameters of Accelerator-Driven Transmutation Concept Proposed by Royal Institute of Technology -ATW Type - [lo]

I tens

-Energy -Current

Target -Equivalent diameter -Equivarent height (cm)

-Target material -Cooling material

Subcritical c o r - -Equivalent diameter -Equivalent height -Material compositior,

Fuel/Clad+Structur?/Coolani/itoderator -Chemiczl form of fuel -hlaterials of coolant 2nd nodera tor

-.Averaged fresh fuel coinposition (w/o) C/Pu/&p. Am. Cm/LLFP

-Isotopic composition of Pu (w/o) 23dpUj,239pU/213pU/2< Lpu/'-Spu

-Isotopic conpos i t ion of !,!.A ( W/O) 2 3 ~ ~ i p / l ~ ~ n / 2 1 2 A n / 2 4 3 ~ m / 2 1 ~ c m ~ 2 ~ . : C m

-Averaged compos i t ion of Long- 1 ived FP (w/o)

I System Characteristics -Effective multiplication factor: k , : :

-Thermal Power in core (h1W t ) -Power density: average (H'/cm')

max imun (w/cm3) -Linear heat rate: average (W/cm)

maximum (W/cm) -Neutron flux averaged in core (n/cm2 s) -Neutron energy averaged in core (KeV) -Fuel dwelling time ( E F P D ) " -Target dwelling time (EFPD) -Burnup reactivity swing (%A k/k/36SEFPD)

Proton L1XI.J.C or Cyclotron Proton or Dectron

--I000 5 - 100

4 0 400

Solid Th or Liquid P b or ' L i - - I

Molten salt or Slurry

1) Equivalent Full Power Day

Table 4 . 3 4 System Design Parameters of Accelerator-Driven Transmutation Concept

Proposed by Institute of Theoretical and Experiuental Physics (ITEP) [I21

1) Equivalent Full Power Day

I tens

4ccelerator

-Type -Particle -Energy (MeV) -Current (mA)

Target -Equivalent diameter (cm) -Equivarent height (cm) -Target material -Cooling material

Subcritical core -Equivalent diameter (cm) -Equivalent height (cm)

-hiaterial coii~posi t ion (v/o) Fue!/Clad+Struc ture/Coolant/?Jodera tor

-Che~ical form of fuel -Xiateiials of coolant and moderator -Averaged fresh fuel composition ( W/O)

ti/Pu/?p. Am. Cm/LLFP

-1sotcpic composition of Pu ( W/O) 2 3 3 ~ U / 2 3 U P u / 2 ~ : l P U / 2 ~ 1 P U / 2 ~ 2 P u / 2 4 1 A m

-Isoto?ic conposi tion of !,!A (w/o) 2 37>ip/2L LAm/2L 2.Am/2?. 3 ~ ~ / 2 ? . i c ~ / 2 < L c ~

-Averaged compos i t ion of Long- 1 ived F? (w/o)

"TC/"' I System Characteristics

-Effective multiplication factor: k , r :

-Thermal Power in core (hiW t)

-Power dens i t y : average (W/cm3) max imum (w/cm3)

-Linear heat rate: average (W/cm) maximum (W/cm)

-Neutron flux averaged in core (n/cm2 s )

-Neutron energy averaged in core ( S P Y )

-Fuel dwelling time (EFPD) "

-Target dwelling time (EFPD) -Burnup reactivity swing (X 8 k/k/365EFPD)

Proton L1XIAC Proton 1OOC) 10 0

5 0 40 0

W , Tb-Bi eutectic alloy H 2 0

SO0 6 0 0

4.7/4. 5/3. 5iS.i. 5 no1 ten sa!:

2/60/2 2/10/1/2

0. 97 1100 ( 3 8 0 W e )

5. 3 10. 6 al. 4 162. 8

' 4

5 . TRANSMUTATION CAPABILITY OF PROPOSED CONCEPTS

The transmutation capability has been usually discussed

using the transmutation raIe defined as a ratio of weight of

minor actinides which is transmuted by fission and capture 10

that of 'initial loading of minor actinides per unit time. There

is another definition, so-called burnup rate, which is a ratio of

weight of minor actinides fncinerated by fission reaction to thaT

of initial loading of minor actinides per unit time [15]. This is

because the aim of transmutarion is the conversion of long-lived

nuclides to shorter-lived or stable nuclides and fission, not

capture,is a real transmutation reaction for minor actinides.

Since it is difficult to separate transmutation by fis-

sion from by capture in the burnup calculation, the Transmu-

tation capability of the proposed concepts is discussed based

on the classical definition.

;LLA(BOEC) - FU(E0EC)

Transmutation capabilitj- = . . . . . . . . . . . . . . . . . . . . . . . . . . . .

(cycle time)~( thermal power)

FIA ( BOEC ) - $IA (EOEC )

Transmutation rate = . . . . . . . . . . . . . . . . . . . . . . . . . .

>lA(BOEC)x(cycle time)

where. hIA(B0EC) and FlA(E0EC): minor actinides quantities at the

beginning and end of equilibrium cycle (kg), cycle time: equiva-

lent full power year ( E F P Y ) between BOEC and EOEC, and thermal

power : ( GWt ) .

5-1 Thermal Reactors

?laterials inventory and mass balance of PIYR based transmuta-

tion concepts proposed by the CEA is given in Table 5.1 for the

homogeneous arrangement of minor actinides, and in Table 3.2 and

5.3 for the heterogeneous arrangement with Np and Am targets. In

the first concept with the homogeneous arrangement,the transmuta-

tion capability is 11.3 kg/G\Yt/EFPY and then the corresponding

transmutation rate is S.O%/EFPY. This substantially low transmu-

tation capability is due to significant buildup of Cm-244. In the

second concepts, the transmutation capability is 37.2 kg/G\'it/EFPY

(11.1%/EFPY) and 17.3 kg/G\Vt/EFPY (16.S%/EFPY) for the S p and .a

targets themselves,respectively.

Materials inventory and mass balance of U02 fueled P\YR based

transmutation concept proposed by JAERI is given in Table 3 . 4 .

The transmutation capability is 11.3 kg/GWt/EFPY (G.S%/G\Yt/EFPY).

The J.AERI concept has the equivalent transmutation rate of Kp-237

with that of the CEA first concept (3.7%/EFPY).

5 . 2 Fast Reactors

Materials inventory and mass balance are given in Table 5.5

to 5.12 for fast reactor based transmutation concepts proposed by

the organizations.

The transmutation capabilities are 57.9 kg/GWt/EFPY for the

CEA first icIOX-LJIFBR, 58.2 for the PNC-FIOX-LMFBR. 53.6 for the

CRIEPI-metal fuel LMFBR, 256.4 for the JAERI P-ABR. 290.3 for the

JAERI L-ABR and 1 1 2 . 6 for the JAERI-Th loaded LMR, respectively.

The cor responding transmuration rates are S.O%/EFPY. 7.0, 1 2 . 4 ,

20.6, 10.4 and 5.5,respectively. -1s for the LYFBR based concepts

loaded heterogeneously with minor actinides targets, the transmu-

tation capability is 48.2 kg/GWt/EFPY (4.9%/EFPY) for Xp-targer

and 39.0 kg/GWt/EFPY (3.9%/EFPY),respectively

The fast reactor based transmutation concepts has higher

transmutation capability than LWR based ones, because of their

higher minor actinides inventory and higher neutron flux than LRR

concepts.

5 . 3 Accelerator Based Transmutation Concepts

Exact burnup calculation is very much troublesome for the

accelerator-driven systems, since various kinds of high energy

particle reactions include in the calculation. Therefore, there

is a few calculation results related to the material mass balance

as shown above for reactor based transmutation systems.

Table 5.13 and 3.14 show simple transmutation rates of

transmutation nuclides, without treating their burnup and decay

chains, for the "PIOX fueled core" and "Particle fueled core"

concepts proposed by the BNL. In these tables,transmutation rate

of Tc-99 and 1-129 is straight forward. The transmutation capa-

bility of the first concept is 84 kg/G\Yt/EFPY (10.7%/EFPY)

for Tc-99 and 13 kg/G\Vt/EFPY (23.1%/EFPY) for 1-129, and the

transmutation capability is 44.3 kg/GWt/EFPY (G.l%/EFPY) for

Tc-99 and 6.4 kg/GWt/EFPY (13.6%/EFPY) for 1-129.

Table 5.15 shows materials mass balance of the "Al.loy fueled

core" concept proposed by the JAERI, using preliminary but more

- - b

rigorous approach as shoivn in Fig.S.1. In this concept,the trans-

mutation capability of minor aczinides is 305 kg/GKz/EFPY (10.7%

/EFPY) .

Table 6 summarizes the data existing in literature

concerning the transmutation capability. In the case of the

PHOENIX concept, the comparison was performed between the Fvaste

inventory and the release limits for 10000 years after disposal

(see Fig.5.2).

' - - . . a \ = + ; - - - I

..-.... ' J . . . : : 5 , . I 3 I !

Fig. 5.1 Flow Chart for B u r n u p Calculation in Accelerator Transmutation System at J A E R I

(Cihonne of Heavy Meta i )

Table 5 . 1 Material inventory and Pass balance of PWR Based Transmutation Concept

Proposed by CEA - floaogeneous Loading of Minor Actinides -

1) Equilibrium C y c l e . 2) Begining of Equilibrium C y c l e .

3 ) End of Equilibrium C y c l e . T ( E 0 E c ) - T ( B O E C ) = A T = 0 . 8 EFPY

Table 5.3 Material inventory and mass balance of Am Target in PWJ3 Based

Transmutation Concept Proposed by CEA - Heterogeneous Loading -

1 ) E q u i l i b r i u m C y c l e . 2 ) B e g i n i n g of E q u i l i b r i u m C y c l e ,

3 ) E n d of E q u i l i b r i u m Cyc!? .

Table 5.4 Material inventory and mass balance of PFE Based Transmutation Concept Proposed by JAEBI

1) Equi l ibriun Cycle, 2) Begining of Equi l ibriun Cycle.

3 ) End of Equilibrium Cycle. T (EOEC) -T (BOEC) = A T= 4 EFPY

Table 5.5 Material inventory and mass balance of MOX-LWBB Based Transmutation

Concept Proposed by CEA - Ho~ogeneous Loading of MA -

Nuc 1 ides

Initial

inventory

(kg)

Fuel

Uranium

U-235

U-236 U-238

Plutonium

Pu-238

Pu-239

Pu-240

~ Pu-241 Pu-242

Minor act inides

Nep t un i um ?ID-2 3 T

Americium

Am-241

Am-24211 Am-243

Cur i urn

Cm-242

Cm-243 Cm-244

Cm-245

Other actinides

( Mass balance in EOC" (unit :kg) I

Transmutation capability 57. 9

1) Equilibrium Cycle. 2) Begining of Equilibrium Cycle. 3) End of Equi!ibrium Cycle. T(EOEC)-T(EoEC) = A T= 4 EFPY

Table 5 . 6 Material inventory and Mass balance of Np target in MOX-FBR Based

Transmutation Concept Proposed by CEA - Heterogeneous Loading -

-1) Equilibrium Cycle. 2) Begining of Equilibrium Cycle.

3) End of Equilibrium Cycle. T(E0EC)-T(BOEC)=A T= 12. 3 EFPY

Xuc 1 ides

Fuel

Uranium U-235

U-2 3 6

U-238

Plutonium

Pu-238

Pu-239

PU-240

PU-241 Pu-242

Minor actinides

Nep t un i uin

\ip-2 37

Americium Am-241

Am-242m

Am-243

Cur i um

Cm-242 Cm-243

Cm-244 Cn-245

Other actinides

Transmutation

Initial

inventory

( k g ) .

capability

, I

(unit : k g )

EOEC-BOEC

+11OS

-i 151 I

Mass

BOEC"

balance in E0cLJ

EOEC"

1108

151 + 21: i I + ! i + 0.: 1

i -2134

I

I

2 (Np)

3574

(kg/CWt EFPY)

2 o 1 0. 1

1440

4 8 .

Table 5.7 Material inventory and mass balance of Am target in MOX-LWBB Based

Transmutation Concept Proposed by CEA - Heterogeneous Loading -

1) Equilibrium Cycle, 2) Begining of Equilibrium ~ycl-e.

3) End of Equilibrium Cycle. T(E0EC)-T(B0EC) = A T= 12. 3 EFPY

Table 5.8 Material inventory and mass balance of MOX-LMFBR Based Transmutation

Concept Proposed by PNC

Transmutation capability (kg/CWt E F P V ) ~ 5 8 . 2

1) Equi 1 ibrium Cycle. 2 ) Begining of Equi 1 ibrium Cycle,

3 ) End of Equilibrium Cycle. T(E0EC)-T(BOEC)=A T= 3 EFPY

Table 5.9 Material inventory and mass balance of Metal Fuel LMFBR Based Transmutation Concept Proposed by CRIEPI

1) Equilibrium Cycle, 2) Begining of Equilibrium Cycle. 3 ) End of Equilibrium Cycle.

Xucl ides

Fuel

Uraniuiil U-2 3 5

U-236

U-238

Plutoniua

Pu-238

Pu-239

Pu-210

Pu-211 Pu-242

Minor actinides

Nep t un i uiil 3p-2 3 7

Americium

Am-241

Am-242m

Am-243

Cur i urn

Cm-242

Cm-243

cm-244 Cm-245

Other actinides

Transnutation

lni t i a l inventory

(!id

capability

Mass

BOEC':

5. 7

1. 1 5635

338

593 330

4 6 8 5

1 7 5

6 6

2.0

6 2

3. 5

0.4

5 9 12

(kg/GW t EFPY)

balance in EOC"

EOEC':

2. 5

1. 7 4978

204

631 3 29

49 ,.?

1 3

9 0

3 2

1. 7 4 1

4. 3 0. 3

5 5 13

53.

(unit:kg)

EOEC-BOEC

- 3. 2

+ 0. i

- 658

t 36 - I - 1. 5

+ 2. 5 - 9.8

- a 5

- 34

- 0. 2

- 21

+ 0. 8

+ 0.0 - 3. a + 1. 6

6

Table 5.10 Material inventory and .ass balance of Helium-Cooled Actinide Burner

(P-,433) Proposed by JAEBI

1) Equilibrium Cycle. 2) Begining of Equilibrium Cycle. 3 ) End of Equi!ibrium Cycle.

Xuc 1 id?s

Fuel

Uranium

1'-235

ti-2 3 6 ti-238

Plutonium

P U-2 3 8 Pu-239

Pu-240

Pu-241 Pu-242

hiinor actinides

Xeptuniun Np-237

.her icium Am-241

Am-242n

Am-213

Cur i urn

Cm-242 Cm-243 Cm-244

Cm-24 5

Other actinides

Transmutation

Initial

inventory

( kg )

-,

904

0

10 1

-+ 1049

$9 3

0

224

0

0. 53 9 5

5 . 3

capability

Mass balance in E O C ~ ' (unit : kg) 1

BOEC"

575

12 5

131

3 7 9

6 2

4 9

2. 1 3 2

808

356

6 . 8

18 0

0. 53

1. 6 12 2

16

2 7

( k z / ~ ~ t <FPY)

I

EOEC"

425

14 2

I

EOEC-BOEC I i 1

- 5 I + 1;

I

125 1 - ; 7 ".

I

1

391

7 9

4 9

3. 7 3 'i

645

277

6. 8

147

2 1

1. 6 123 16

2 7

+ 12 i

+ 15 I + c . 0 I

+ 1. j 1 i t 2.7 i

I I

I - 1 5 2

- 3 - f b I + 0.0 1 - 33 1

t 21) + 0.0 + 0. 53

+ 0. 53

+ 0 . 3 3 1 256. 4

Table 5 . 1 1 Material inventory and mass balance of LLead-Cooled Actinide Burner

(L-ABB) Proposed by JAEBI

Transmutation capability 290. 3

1) Equilibrium Cycle. 2) Bkgining of Equilibrim C y c l e .

3) End of Equilibrium Cycle.

Table 5.12 Material inventory and Mass balance of Th-Loaded Lead-Cooled Fast

Reactor Based Transmutation Concept Proposed by JAEBI

1 ) Equilibrium Cyc le . 2) Beginin3 of Equi iibrium C y c l e .

3) End of Equilibrium Cycle.

Table 5.13 Material Inventory and Transmutation fate of 'MOX Fuel Core- Concept Proposed by BNL

Table 5.14 Material inventory and transautation rate of -Particle Fuel Core- Concept Proposed by BNL

Nucl ides

Fuel Uranium

U-235 U-2 3 6 U-238

Plutonium Pu-238 Pu-239 Pu-240 Pu-241 Pu-242

hlinor actinides Neptunium

Xp-2 3 7

Amer i c i um Am-24 1 Am-24 2ii1 Am-243

Curium Cm-242 Cm-24 3 Cm-244 Cm-245

Other actinides

Long-lived FP Tc-99 1-129

(1-127)

1 TransnutaLion capabi

Initial inventory

(kg)

-,

108 338 243

5 7 3 1

7 6 4

3 6 0 -

16 3

-

7 4 - -

5 5 0 3 9 13

1 i i y

I

(kg/year)

total

i

Transmutation .

by capture

rzte

by fission

I i

i I 206 i

i I

107 I

31 I 17

I

i i 59 i !

3 1 1 i I --- I

12 1

6 9

Id

6. 8

5 9 9

(kg/GW t EFPY)

5 5

3 7

13

10

Table 5.15 Materials Inventory and Mass Balance of Accelerator-Driven Transmutation Cncept Proposed by JAEBI - A1 loy Fuel Core System -

1 Transmutationcapability (kg/GWt EFPY) I 3 0 7 I 1) Equilibrium Cycle. 2) Beginins of Equilibrium Cycle. 3) End o f Equilibrium Cycle.

T (EOEC) -T (BOEC) =AT= 6 EFPY

6 . CALCULATIONAL METEIODS

6.1 Thermal and Fast Reactors

In the case of thermal and fast reactors, the presence of

actinides does not change the calculational methods presently

applied provided the nuclear data, in particular cross-sections

are available in the required group structure. Self shielding

effect and Doppler broadening are calculated in the usual manner.

Multigroup cross-section libraries are generated by retriev-

al and processing codes (such as NJOY) which depart directly from

ENDFB/4, ENDFB/S, ENDFB/G, JEF or JEF2 and JESDL3. In the case of

continuous cross-section libraries as used by some hIonte Carlo

codes,in particular YCSP,the ACER module of SJOY is applied.

Reactivity and flux calculations are carried OUT for thermal

reactors by code systems like WIPIS or APOLLO-ICAFK-A. Fast reactors

are calculated by similar systems based on nodal methods or

different S\ codes. Burnup and fission-product decaj- is deter-

mined by ORIGEN,ORIGISS,FISBI3 or PEPIN.

6 . 2 Accelerator Driven Systems

In accelerator-driven systems, the neutron energy extends to

several hundred MeV. In classical reactor codes the upper energy

limit lies between 15 and 20 PIeV. In accelerator driven systems,

between 10% and 20% of the spallation neutrons are above this

limit and therefore require to be considered in the calculations.

Usually a classical >lonte Carlo code like NCNP or ?IORSE is

- G

extended to higher energies by codes like HETEC (High Enere-

Transport Code,OR_\;L) or WITC (Neutron Meson Transport Code).

Several research institutes upgraded two of the above men-

tioned codes and tailored the combination to their own require-

ments. Brookhaven National Laboratory is for example using BXLF.2

modified combination of NNTC and MORSE. A similar combination is

used by J-\ERI and the KFA JuJich. In the Rutherford Laboratory

Atchison used experimentaEly adjusted parameters for the fission

process and put this fission model into HETC. -At L-ISL Prael

developed a code system by combining an upgraded HETC ivith 'ICS?

and called it LAHET. In LAHET the geometry transport capability

is that of the Monte Carlo Neutron and Photon transport code

?lC6P. LAHET includes two models for fission induced by high

energy interactions : the ORNL model and Rutherford Apple~on

Laboratory model. HETC treats all interactions by protons, pions

and muons, but neutron interactions only above the cut-off energy

of 20 >lev.

F. Atchison reviewed in detail "Data and Jlethods for t h e

Design of .Accelerator Based Transmutation Systems" [241.

6.3 Used Calculational Methods for Proposed Concepts

At the CEA, in the case of LWRs, nuclear data used is JEF.

Neutronic calculation is made by two dimensional transport code

APOLLO with 99 group cross-sections generated from JEF. The code

is a modular one which solves the multigroup transport equation

by the collYsion probability method,and a multicell approximation

is available for 2D geometries. Burnup calculation is made by the

- - b

three dimensional diffusion code CROXOS which allows pin by pin

core calculations.

In the case of fast reactor,standard isotopes cross secrions

are provided from CARSXVAL. Minor actinides data are provided

from JEF and added to the standard isotopes cross-sections.

Neutronic calculations are made by two dimensional diffusion code

ER-4x0s with 25 group cross sections mentioned above. Burnup is

calculated also by ERAKOS. .

AT the JAERI, nuclear data library JENDL-3 is commonly used

for transmutation studies. In the case of LWR, neutronic calcula-

tion is made by the two dimensional diffusion option of the

modular code system SRAC with 107 group cross-sections generated

from JENDL-3. However ,resonance absorptions are calculated accu-

rately by the ultra-fine group method. Burnup calculation is

carried out by zero-dimensional burnup option of SRAC, where FP

chain treats explicitly important 63 fission product nuclides.

In the case of fast reactors including burner reactors,

neutronic calculations are carried out by the two dimensional

diffusion option of the code system ABC-S or SRAC/COREBN, with 70

group cross-sections generated from JENDL-3. Burnup is calculated

with burnup option of the above mentioned code systems.

In the case of accelerator-driven system,nuclear data source

comes from JESDL-3 or ESDFB/5. High and medium energy reaction

calculations are performed by SMTC/JAERI. Neutronic calculations

are made by TWOTRAN-2 with 30 group cross-sections. Burnup is

calculated by COPIRAD of which flow chart is given in Fig.3.l.

The three Japanese organizations described below propose

their L?IFBR based transmutation systems.

At the PNC,neutronic and burnup calculations are carried OUT

by the two dimensional diffusion code CITATION-FBR (a modified

version of original CITATION) with 18 group cross sections gener-

ated from JENDL-2.

.*

At the CRIEPI, standard isotopes cross-sections are provided

from JFS 70 group cross-sections set and minor actinides cross

sections are generated from JENDL-2 or ENDFB/S. Neutronic and

burnup calculations are carried out by two dimensional diffusion

code CIT-ATION-TRU with 70 or 18 group cross-sections. The code is

a modified version of CITXTION which can easily calculate complex

minor actinide nuclides burnup/decay chain.

At the Toshiba Corporation, JFS 70 group cross-sections set

is used for standard isotopes and minor actinides cross-sections

are generated from JEYDL-3. Neutronics are calculated with 70

group two dimensional or 7 group three dimensional diffusion

code. Burnup calculations are based on two dimensional code.

The research organizations described below propose their

accelerator-driven transmutation systems.

At the BXL, Suclear data source is ENDFB/4. High and medium

energy reaction calculations are carried out by L - M E T and N>lTC/

BSL. In the latter, a fission model based on the Fong's statisti-

- i

cal model is adopted into the ?;?TTC code. Seutronic calculations

are performed by- the continuous energy Monte Carlo code YCSPH.

Burnup calculations are based on the ORIGES-2 code with associat-

ed cross-sections.

At the EYEA, nuclear data source is JEF or ENDFB/5. High and

medium energy reactions are performed by NMTC/JAERI and HETC/KF-12

codes. Neutronics are calculated with the bIonte Carlo code ;LICXP-4

and burnup is calculated by the ORIGEN-1 and 2 codes.

At the Royal Institute of Technology, nuclear data source is

ENDFB/S or 6. High and medium energy reaction calculations are

made by the LAHET code. Neutronic and burnup calculations are

carried out by ?ICNP-4 and ORIGEN-2,respectively.

At the ITEP, nuclear data is ENDFB/6. Seutronic and burnup

calculations are performed by a two dimensional transport code

with 26 group cross-sections and the three dimensional TRIFOB

code,respectively.

7 . INTEGRAL EXPERIMENTS

7.1 Thermal Reactors

Extensive integral experiments on minor actinides have been

carried out and are planned especially in France. As for LWR,as a

part of the SHERWOOD program,irradiation experiments were carried

out by irradiating GO2 pins doped with minor actinides at center

the material testing reactor YELUSINE. The capture reaction rates

were measured for Am-241,Am-243, and Cm-244. During R&D on high

conversion light water reactor, the ICARE program and the ERASJIE

program were performed. In the ICARE program,two experimental LO2

pins doped with minor actinides were irradiated at the center of

261 pin lattice of MOW fuel in the NELUCINE. The ERASME program

was performed in the critical facility EOLE at Cadarache. '

In the framework of the SPIN program, the ACINEAU experimenr

is planned to take place in the OSIRIS at Saclay at the end of

1994. The purpose is to assess the feasibility of transmutation

of Np and .Am in water reactors, to study the metallurgical beha-

viour of fuels and targets and the therrnomechanical behaviour of

the rods with a view of their subsequent optimization, and to

compare the efficiency of actinide burning in a thermal flux with

that in a fast flux.

7 . 2 F a s t Reactors

Many integral experiments at fast reactors on minor acti-

nides were carried out in several countries.

- - b

In the united s t a t e of .America, i r r a d i a t i o n experiment U-as

performed at the experimental fast reactor EBR-I1 to measure che

capture and fission rates of Xp-237 together with major acti-

nides. In the Fast Flux Test Facility (FFTF), transmutaxion of

Tc-99 was measured using moderated assemblies at the periphery of

the FFTF. Ender the collaboration with the UKAEA, extensive irra-

diation experiments were carried out at the UK prototype fast

reactor PFR. Irradiated minor actinides are Np-237,Pu-238,Pu-242,

Pu-244,Am-241,:k11-243,Cm-243,Cm-244 and higher Cm.

In France,PROFIL,TRAPU and SUPERFACT programs were performed

in the French prototype fast reactor PHENIX. In the PROFIL pro-

gram, separate isotope samples were irradiated,which included Sp-

237,-Am-241,Am-243, Cm-244. In the TRAPU experiment, special fuel

pins were irradiated and analyzed to identify the detailed iso-

tope composition in the irradiated pins. The SUPERFACT program

has allowed to conclude on the feasibility of transmutation of 1-p

homogeneous fuels (in the range of 2% to 4 3 % ) and its extension

to higher burnup ( > 10 a%) is under way. In the fast critical

assembly ?LASURUC-4,physics experiments have been carried out, such

as fission chamber measurement and reactivity measurement. The

BALZAC experiment measured the reactivity/atom of minor Pu and

Am-231. Basing on the experiences of the SUPERFACT.a irradiation

experiment of many pins with 2 % Np is planned to start in 1994

in the SUPERPHENIX.

In Japan,at the JAERI, a series of integral experiments were

performed at the critical assembly FCA, to measure fission rate

and sample reactivity of minor actinides, such as Np-237, Pu-238,

Am-241 and .Am-243. The measurements were made in the wide range

of neutron spectrum fields. Under the collaboration with rhe ORSL

in the USA, the minor actinide samples irradiated at he PFR

mentioned above were sent to the J-AERI and are being analysed to

verify the JENDL cross sections. At the P Y C , reaction rate meas-

urement of Np has been made at the pulse reactor YXYO1,and trans-

mutation rate measurement and irradiation test of minor actinide

containing pins are planned at the experimental fast reactor

JOYO. The CRIEPI has proceeded nuclear data verification test at

the KNK-I1 reactor under the collaboration with the TUI and KfK.

7.3 Accelerator-Driven Systems

S.Cierjacks overviewed the present status of integral vali-

dation experiments related to neutron and proton produc~ion ,and

spallation and fission products [ 2 3 ] .

(1) Experiments planned in the United State of .America

An apparatus is built to generate the nuclides under condi-

tions simulating a spallation neutron production,which is used at

one of the beam lines of the Brookhaven Sational Laboratory's

Alternate Gradient Synchrotron. Target material used is lead or

tungsten and proton fluence is determined by an aluminum foil.

The lead or tungsten foils are irradiated to energies in the

range of interest,about 1 GeV, and assayed for their gamma spec-

tra, using germanium diode gamma detectors. The spectra are then

analyzed for the quantities presence of the respective nuclides

as a function of time after the end of irradiation.

- L

The af~erheat develops in target material due to formarion

of radionuclides during proton-bombardment spallation process.

It is experimentally determined by directly evaluating rhe hear

produced by the spallation-induced radioactivity using a novel

type of calorimeter. The first experiments made use of the BXL

Linac Isotope Producer with 0.2 GeV proton energy impinging on

lead target.

An experimental study has been conducted at Texas -4 & > I

University to compare measured and calculated spallation product

yields from a lead target. A special spallation product, decay

gamma 1ibrary.was constructed and incorporated into the standard

GETIE gamma-ray peak and nuclide identification software. -4 rota1

of 11 different nuclides in nine mass chains were identified from

the multichannel counting spectra. Additional experiments employ-

ing copper,gold and thorium targets are planned.

In order to resolve the difference between the code predic-

tion and existing experimental data relating to the n/p parameter

, a related experiment is made by creating a configuration that is

simple enough from a material and geometric standpoint.

The experiment uses the BiYL's Alternate Gradient Syncrotron.

ItIomentum-analyzed protons in the giga-electron-volt range will

pass through two "paddle" detectors.further defining their energy

by time of flight and arranged to produce a coincidence for each

proton of the correct energy passing through to a lead or tung-

sten target located within a neutron detector. This detector is a

cylindrical ( C H 2 ) moderator pierced by He-3 proportional coun-

ters arranged concentrically with the cylinder axis.

- ., ( 2 ) Experiments planned in Japan

Spallation integral experiments are underway in order to

obtain data on nuclide production ,to estimate the yield of neu-

trons and spallation products, and to investigate the validity of

the spallation simulation code NMTC/JAERI. The 500 MeV booster

proton syncrotron facility at the National Laboratory of High

Energy Physics (KEK) is used for the experiments. Target materi-

als used so far are lead and tungsten. Experiment with a targec

of depleted uranium is also planned. The number of induced reac-

tions in the activation samples were measured. The measured

results agree fairly well with the prediction by R>ITC/J.L\ERI.

(3) Experiments planned in Switzerland

The experiment ATHEY,\ is in progress at the Paul Scherrel-

Institute, which is aimed at solving some specific data and meth-

ods problems relating to the accelerator-based transmutation of

actinides. In a first phase,thin samples of actinides are irradi-

ated with 590 $lev protons from the PSI ring accelerator. The

generated spallation and fission products are analyzed and are

compared with theoretical predictions. The principal motivation

is to confirm the high potential of the high-energj- fission

reactions for transmutation. In a second phase, it is proposed

to study relating to multiplying target-blanket assemblies.

8 . CONCLUSIONS AND RECOMMENDATIONS

The main purpose of phase-I of this program is to have

common understandings in the member countries for the announced

transmutation devices,their physics performances and their relat-

ed fuel cycle consideration. To this end the task force has

examine more than 20 different concepts of transmutation and has

collected the results of the calculation of transmutation rate of

each concept.

The transmutation concepts which the task force has examined

are listed below;

1) Thermal Reactor based concepts

- PWR based systems with homogeneous and heterogeneous

arrangements of minor actinides,contributed from the CE-4,

- PWR based system,contributed from the J-4ER1,and

- High flux PBR based system proposed by the BNL,taken from

the open literature.

2) Fast Reactor based concepts

- MOX fueled systems with homogeneous and heterogeneous

arrangements of minor actinides,contributed from the CEA,

- MOX fueled system,contributed from the PNC,

- Metal fueled system,contributed from the CRIEPI,

- Metal fueled system with flat core,contributed from

the Toshiba Corporation,

- Two types of minor actinides burner system,contributed

from the JAERI,

- Th loaded system,contributed from the JAER1,and

- ALMR actinide recycling system proposed by the GE and ANL,

taken from the open literature.

3) Accelerator based concepts

- Two types of system consisting of fast subcritical core

and small accelerator,contributed from the BNL,

- Three types of system consisting of fast subcritical core

and relatively large accelerator,contributed from J . A E R I ,

- Los Alamos ATW system,taken from the open literature,

- PHOENIX system proposed by the BNL,taken from the open

literature,

- Three types of Tc-99 incineration system,contributed from

the CEA,and

- ATW-type systems,contributed from the ENEA,Royal Institute

of Technology and ITEP.

Based on the comparison of the results of the calculation of

transmutation rate of each concept, there seem to be significant

discrepancies between the concepts. As for reactor based system,

for example, there is a significant difference of Cm-244 burnup

characteristics among fast reactor based systems, although it

might depends not only cross-section itself,but also neutron flux

and initial isotopic vector of minor actinides. And there is only

one calculated result of mass balance of minor actinides between

BOEC and EOEC of accelerator based systems, this might be partly

due to some difficulties of taking account of high energy neutron

above 20 MeV into burnup calculation. Therefore, it is felt that

further investigation by benchmarks for a set of common system

on calculation methods of transmutation rate should be beneficial

in order to understand and assess their physics performances pre-

cisely.

In addition to that, the task force members have suggested

that the following fundamental issues should be resolved in terms

of scientific aspects of transmutation concepts,such as:

- radiotoxicity after transmutation

- safety features of transmutation system

- nuclear data of transuranium nuclides

A further discussion should be held to analyze discrepancies

and uncertainties lying in the above scientific issues. Also it

has been suggested that the above issues are too broad to be

reviewed thoroughly within the present small task force. There-

fore,as a starting point of phase I1 of the program, a specialist

meeting is proposed with a view to identifying subjects to be

handled by the NSC with respect to the above issues.

PREFERENCES

[I] J.Y.Barre,J.Bouchard: French R&D Strategy for he Bach End

of the Fuel Cycle, Proceedings of International Conference

and Technology Exhibition on Future Nuclear System;Emerging

Fuel Cycles & Waste Disposal Options,GLOBXL'93,P27-32,

Seattle,\Yashington, September 12-17,1993.

[2] Japan's Atomic Energy-Commission; Long-term Research and

Development on Partitioning and Transmutation Technology in

Japan,October,l988.

[3] M.L.Thompson,J.E.Quinn: The ALMR System's Flissions for

Transmuting Waste into Energy,Proceedings of International

Conference and Technology Exhibition on Future Suclear

System;Emerging Fuel Cycles & Waste Disposal Options

GLOB~~L'93,P110-117.Seattle,\Yashington,September 12-17.1993.

[4] E.D.Arthur,H.J.Dewey: The Los Alamos Accelerator Transmuta-

tion of Muclear Waste (ATW) Concept,LAXL-USA,Presen~ed at

the Symposium on Separation Technology and Transmutation

Systems (STATS),\ii'ashington D.C.,January 13-14,1992.

[51 G.J.Van Tuyle et al.: Accelerator-Driven Subcritical Target

Concept for Transmutation of Nuclear Wastes,Nuclear Tech

nology,Vol.lOl,Pl-17,January 1993.

[61 J.Ponel1 et al.: High Flux Particle Bed Reactor Systems for

Rapid Transmutation of Actinides and Long Lived Fission

Products,Proceedings of International Conference and Tech-

nology Exhibition on Future Nuclear System;Emerging Fuel

Cycles & Waste Disposal Options,GLOBAL'93,P1300-1305,Seat-

tle,Washington,September12-17,1993.

- IL.

[71 H.Takahashi: The Use of Minor Actinides and a Small Proton

Accelerator for Fast Reactor with a High Breeding Gain.

Proceedings of the Specialists' Meeting on Xccelera~or-Based

Transmutation,P63-82,PSI Villigen,March 24-26,1992.

[ 8 ] P.A.Landeyro: Contribution from the ENEA in Italy to the

Task Force Request.

[9] ?l.Ka~vashima: Contribution from the Toshiba Corporation in

Japan to the Task Force Request.

[lo] W.Gudowski: Contribution from the Royal Institute of Tech-

nology in Sweden to the Task Force Request.

[Ill V.S.Kagramanian: Contribution from the Institute of Physics

and Power Engineering (IPPE) in Russia to the Task Force

Request.

[12] G.V.Kiselev: Contribution from the Institute of Theoreticzl

and Experimental Physics (ITEP) to the Task Force Reques~.

[I31 J.A.Rawlins et al.: CURE: Clean Use of Reactor Energy,

WHC-EP-0268 ,$lay 1990.

[14] A.Zaetta: Contribution from the CEA in France to the Task

Force Request.

[I51 H.Takano: Contribution from the JAERI in Japan to the Task

Force Request.

[I61 T.Wakabayashi: Contribution from the PNC in Japan to the

Task Force Request.

[17] C.L.Cockey et al.: Higher Actinide Transmutation in the ALJlR

Proceedings of International Conference and Technology

Exhibition on Future Nuclear Systems;Emerging Fuel Cycles &

Waste Disposal Option,GLOBAL'93,P123-135,Seattle,\Yashington,

September 12-17,1993.

- i

[ I 8 1 T.?latsumura: Contribution from the CRIEPI in Japan TO the

Task Force Request.

[19] T.?Iukaiyama: Contribution from the JXERI in Japan to the

Task Force Request.

[ 2 0 ] T.Takizuka,T,Nishida,T.Sasa: Contribution from the J-4ERI in

Japan to the Task Force Request.

[21] H.Takahashi: Contribution from the BNL in the US to the Task

Force Request. ..

[22] i\.I.Cappiello et al.: Target/Blanket Conceptual Design for the

Los Xlamos ATW Concept,Proceedings of the Specialists' $leer-

ing on Accelerator-Based Transmutation,PSI Villigen,>larch

24-26,1992.

[23] C.Bomnan et al.: Suclear Energy Generation and Kaste Trans-

mutation Using an Accelerator-Driven Intense Thermal Seutron

Source,LA-UR-91-91-2601,LAhiL,(1991).

[241 F.rltchison: Data and Nethods for the Design of Accelerator-

Based Transmutation Systems,Proceedings of the Specialists'

>Ieeting on Accelerator-Based Transmutation,P440-498,PSI

Villigen ,March 24-26,1992.

[251 S.Cierjacks: Cross Section I \ - l easurements , In tegra l Validation

Experiments and Nuclear-Model Code Developments of Impor-

tance to Accelerator-Based Transmutation,Proceedings of the

Specialists' Meeting on Accelerator-Based Transmutation,

P356-567,Villigen,Plarch 24-26.1992.

APPENDIX

F O R M - A I INFORMATION ON T R A N S M U T A T I O N DEVICES

AND THEIR PHYSICS P A R A M E T E R S

Name :

A f f i l i a t i o n :

Address :

Tel : Fa. : :

I . General

(1) Selec ted nuc l i de s t o be transmuted i n your t r a n s m u t a t i o n dev i ce :

( 2 ) P r i n c i p a l design cons ide ra t i on i n your de s ign work, such as t r a n s n u t a t i o n c a p a b i l i t y , conserva t iv? d e s i g n , small change from o r i g i n a l core performanc?, s a f e t y cons idera t ion and so on:

1 ( 3 ) Type of t ransmuta t ion d e v i c e s , r e a c t o r o r a c c e l e r a t o r - d r i v e n d e v i c e :

O r e a c t o r 0 a c c e l e r a t o r - d r i v e n dev i ce

I f you give x f o r r e a c t o r , g o t o Sec t ion I 1 i n page 2 , and i f you g i v e x f o r a c c e l e r a t o r dr iven device ,go t o Sect ion I11 i n page 8 .

11. Nuclear Reactor used f o r t r ansmuta t i on

We provide t h e i n fo rna t i on form f o r ura im-plu tonium c y c l e k s e d nuc l ea r r s r c t o r h t r e . Then , i f you work f o r thorium-urmium based r e a c t o r , p l e a s e k i n d l y provide s i n i l a r i o ~ .

(1) Desc r ip t i on of nuc l ea r r e a c t o r :

1 ) Type of r e a c t o r :

PWR BWR Gas-cooled r e a c t o r Heavy wzter-moderated r e a c t o r LMFBR Liquid metal-cooled r e a c t o r O the r s

-,.

I f you s e l e c t Others o r t h e r e i s f a i r l y laqge d i f f e r e n c e between your propos ing r e a c t o r and i t s o r i g i n a l r e a c t o r l i s t e d above ,p lease g i v e i t s b r i e f d e s c r i p t i o n a- i l l u s t r a t i o n here or another whi te s h e e t .

2 ) Thergal and e12c l r i c powers:

Thernal power : M Wt E l e c t r i c power: MWe

3 ) Core dimension

Equivalent core d iameter : cm Equivalent core h igh t : cm

4 ) Fuel arrangement regard ing t r ansmuta t i on ~ a t e r i a l s :

Honogeneous arrangement Heterogeneous arrangement

In t h e case of Heterogeneous arrangement ,

Volume r a t i o of f u e l s w i t h t rannsmuta t ion m a t e r i a l s t o a l l f u e l s : w ,TI

5 ) Type of f u e l s :

Oxide f u e l N i t r i d e f u e l Carbide f u e l N e t a l l i c f u e l

6 ) Core averaged f r e sh f u e l composition (w t%) :

U : P u : Np,An,Cm : long- l ived FP =

7 ) I s o t o p i c ~ o n p o s i i i o n ( w t z ) o f P u , H i n o r A c t i n i d e ( N ~ , A m . C n ) z n d l o n g - i i v l d F? : . Pu

H i n o r A c t i n i d e (Np,An, Cm)

L o n g - l i v e d f i s s i o n p r o d u c t s ( e x . T c , I e t c . )

P l e a s e a d d i s o t o p e s o f l o n g - l i v e d F P , i f n e c e s s a r y .

8 ) A v e r a g e d c o r e c o r n g o s i t i o n ( ~ 0 1 % ) :

i I

I F u e l 1 C l a d i S t r u c t u r e C o o l a n t , ? ! o d ? r a ~ ~ ; r < I ! I I

* : f o r g a s - c o o l e d r e a c t o r , h e a v y ~ a t e r - m o d ~ r a t e d r e a c t o r , e i c .

( 2 ) G e n e r a l p h y s i c s p a r a m e t e r s a n d t r a n s m u t a t i o n c a p a b i l i t y :

I ) P r i n c i p a l c o r e c h a r a c t e r i s t i c s :

I t e n s

T h e r m a l power i n c o r e ( n w t ) Power d e n s i t y : a v e r a g e d i n c o r e (w/cm3)

maximum i n c o r e (w/cm3) L i n e a r h e a t r a t e : a v a r a g e d (w/cm)

maximum (wlcrn) N e u t r o n f l u x a v e r a g e d i n c o r e (n /cm2 .set) N e u t r o n e n e r g y a v e r a g e d i n c o r e ( K e v ) F u e l d w e l l i n g t i m e i n c o r e (EFPD)'" Burnup r e a c t i v i t y s w i n g (%6 k/k/365EFPD)

* : E q u i v a l e n t F u l l Power D a y s , I ) B e g i n n i n g o f l i f e , 2 ) B e g i n n i n g o f E q u i l i o r i u a cyc!?

2) Inventory 2nd trmsnutation czgability:

- Fuel end transnutation meterial inventory and their aess S2lence

1) :3eginning Of Equilibrium Cycle, 2): End Of Equilibrium Cycle. Please add isotopes of long-l ived FP for transmutation, i f necessary.

- Transmutation c a p a b i l i t y

Difference of minor a c t i n i d e s betwzen E O E C t?!d BOEC Transmutat ion c a p a b i l i t y =

(EFPY"between EOEC and B O E C ) x thermal poazr(YKt)

*:Equivalent F u l l ?over Year.

(3 ) Sa fe ty - r e l a t ed phys ics parameters:

P l ea s? f i l l t h e t a b l e with t h e values a t BOL.

Parameters I

I Values I I

1 Coolant void r e a c t i v i t y e f f e c t ( % 6 k / k ) I

\ I

- A11 r e a c t o r region voided

- Core 2nd a x i a l blanket voiti~ti"

- O n l y core voided"

Doggier r e a c t i v i t y c o e f f i c i e n t ' (Idk/dT)

E f f e c t i v e delayed neutron f r a c t i o n *

Proapt neutron l i f e time" ( s s c )

Contol rod mz te r i a l

Cen t r a l con t ro l rod worth (%6 k/k /kg)

*:only f o r f a s t r e a c t o r s .

( 4 ) Data and method used in r eac to r design work:

1) Nuclaz: d a t a :

Nuclear da t a source l i b r a r y o r group c ros s - sec t ion l i b r a r y :

ESDF/B- D JEF 17 JEXDL- O Others

Cl ORIGEN- WI?lS JFS E l C , - \ R N A V A L - 17 Others

If y o u s e l e c t O t h e r s o r t h e r e are a n y s i g n i f i c a n t m o d e f i c a t i o n s f o r t h? o r i ~ i n a l l i b r a r y on m i n o r a c t i n i d e s a n d l o n g - l i v e d FP i s o t o p e s , 2 1 2 ~ ~ 2

g i v e t h e i r b r i e f d e s c r i p t i o n .

I i

2 ) C a l c u l a t i o n m e t h o d :

N e u t r o n i c c a l c u l a t i o n c o d e :

D i f f u s i o n T r a n s p o r t 2 D C1 3D

S u n b ~ r Of e n e r g y g r o u p u s z d f o r c s l c u l a t i o n :

If t h e r e a r e a n y s i g n i f i c a n t d i f f e r e n c e s b e t w e e n y o u r c o d e a n d u s u s l l y u s 2 6 c o d e , p l e a s e g i v e t h e i r b r i e f d e s c r i p t i o n .

B u r n u p c a l c u l a t i o n c o d e :

I f t h e r e a r e a n y s i g n i f i c a n t d i f f e r e n c e s i n c a l c u l a t i o n m o d e l a n d / o r b u r n u ? c h a i n model o f m i n o r a c t i n i d e a n d l o n g - l i v e d FP i s o t o p e s , p l e d s e g i v e t h e i r b r i e f d e s c r i p t i o n .

( 5 ) I n t e g r a l experiment r e l a t e d t o t ransmuta t ion :

1) l r r c d i a t i o n experiments i n power r eac to r :

Nvne of f a c i l i t y : Fas t r e a c t o r Thermal r e a c t o r

- I r r a d i a t e d minor a c t i n i d e and long-l ived FP i s o t o p e s :

Purpose of experiment: a*. Nuclear d a t a v e r i f i c a t i o n Transmutation r a t e measurement

O Others

I f you s e l e c t O the r s ,p l ea se give a b r i e f d e s c r i p t i o n .

Experiment i n c r i t i c a l assembly and research r e a c t o r :

- Nme of f a c i l i t y : Fast neut ron f i e l d Thernal neutroa f i e l d

Yeasured minor a c t i n i d e and long-lived FP i s o t o p e s :

Kind of measurements: Reaction r a t e r a t i o 0 Sample r e a c t i v i t y worth 0 Others

If Y O U s e l e c t O the r s ,p l ea se give a br ie f d e s c r i p t i o n .

3 ) Future programme, i f you have :

Thank you f o r your hear ty coopera t ion ! !

El. Accelera tor -dr iven Devices used f o r t r m s s u t a t ion

We provide t h e information form f o r urwium-plutoniun c y c l e oesed r ,uclear r e ? t o r s h e r e . .

Tnen, i f you work f o r thorium-uranium b ~ s e d r e ~ c t o r s , p l e a s e k ind ly provide s l a l l a r f o r a .

(1) Desc r ip t i on of Acce l e r a to r d r i v e n devices :

1) Type of Acce le ra tor :

Proton LINAC l e c o n I D Others

I f you s e l e c t O the r s , p l ee se g i v e i t s b r i e f d e s c r i p t i o n o r i l l u s t r a t i o n h2re o r another white s n e e t . -,

2) P a r t i c l e ' s energy and c u r r e n t :

P a r t i c l e : . .

Energy : Me'i Current : m.4

3 ) Type of device :

@ S u b c r i t i c a l core 12 S u c r i t i c a l core wtih t a r g e t " O Others * : s p e c i f i c s p a l l a t i o n t a r g e t , such 2s Pb 2nd W.

I f you s e l e c t O the r s , p l ea s2 g ive i t s b r i e f d e s c r i p t i o n o r i l u s t r a t i o n her? o r another white s h e e t

I

4 ) Xeutron s ~ e c t r u m i n d e v i c e :

Fast Thermal Others

5 ) Thermal and e l e c t r i c D O , , * ~ I - s :

Thermal power : ?t#t E l e c t r i c power : ?!We

6 ) Target and/or s u b c r i t i c a l core dimens i o z , 2nd subc r i t i c= ! i t y

Equivalent t ~ r g e t d i m e i e r : cn and hz ight : cm

- Equivalent s u b c r i t i c e l core diameter : cn and ngight : cm

- Ef fec t ive neutron m u l t i p l i c a t i o n f a c t o r :

7 ) Fuel arrangement in s u b c r i t i c a l co re :

CJ Homogeneous s r rangesne t - i? Heterogeneous arrangement

In t h e case of i i e t~rogeneous a r r a g e m e n t ,

Volume r a t i o of f u e l s with t r m s m u t a t i o n m a t e r i a l s t o a l l f u e l s : V "I

8 ) T ~ D ? of f u e l s :

@ ? i e t a l l i c fu?! Xolten s a l t o r s l u r r y f u e l Oxide f u e l N i t r i de f u e l O Czrbide 4221 C! Others

I f you s e l e c t O the r s ,p l ea se give b r i e f d e s c r i p t i o n .

9 ) Core avearged f r s s h f u ? i c o n ~ o s i t i o n (wi%) :

U : P u : Xp,An,Cm : long-l ived FP =

1 0 ) I s o t o p i c conposi t ion (xi%) of Pu, minor zctinide(No,An,Cm) and lonz-l ived FP:

. P u

?tinor a c t i n i d e (Np.An,Cn)

- h

long-lived f i s s i o n products (ex .Tc, I e t c . )

12) Type of t a r g e t

Tc-99

4

Target mater ia ls and i t s cooling mater ia ls I

i I I i

1-129 I Sr-90 I Cs-137

( 2 ) General physics parameters and transmutation capab i l i ty

1 ) ?rincipa! core & t a r g e t c h a r a c t e r i s t i c s :

Mater ia ls and Avarged composition i n s u b c r i t i c a l core:

H ~ t e r i a l s

Ave . composition ( ~ 0 1 % )

*:Equivalent Full Power Day, 1 ) Beginning of l i f e , 2 ) Beginning of equil ibrium cycle

Items

Thermal power i n core (MWt) Power densi ty : averaged i n core (w/cm3)

maximum i n core (w/cm5) Linear heat r a t e : averaged (w/cn)

maximum (wlcm) Neutron f l u x averaged in core (n/cm2-sec) Neutron energy averaged i n core (keV) Fuel dwelling time i n core ( E F P D ) ' Target dwelling time i n core (EFPD )' Surnup r e a c t i v i t y swing (%6 k/k/365EFPD)

Fuel 1 CladtSt ructure

BOLL'

Coolant I Moderator

---

B O X C 2 '

'

-- -

2 ) I n v g n t o r y a n d t r a n s m u t a t i o n c a u ~ b i l i t y :

- F u e l a n d t r a n s m u t a t i o n m a t e r i a l i n v e n t o r y a n d t h g i r m a s s b ~ l a n c z

1 ) B e g i n n i n g O f E q u i l i b r i u m C y c l e , 2 ) End Of E q u i l i b r i u m Cyc le . P l e a s e a d d i s o t o p e s o f l o n g - l i v e d F P f o r t r a n s m u t a t i o n . i f n e c e s s a r y .

Dif f e r ence of minor a c t i n i d e s between E O E C 2nd 3032 n ~ r z n s n u t t t i o n c a p a b i l i t y =

(EFPY*betweon E O E C and B O E C ) x thermal power(X'H;t)

- - lig1MIi-l- EFPY

*:Equiva len t F u l l Power Year.

( 3 ) Safe ty - r e l a t ed phys i c s parameters :

P lease f i l l t h e t a b l e wi th t h e - v e l u e s a t BOL

( 4 ) Data and method used i n t h e device des ign work:

1) Nuclear d a t a :

Nuclear d a t a source l i b r a r y o r group c ros s - s ec t i on l l b r a r y :

ENDFIB-- JET JE;iDL- Cl HILO Others

Ci ORIGEN-p WI!;S JFS C A R N A V A L - El Others

Parameters

Coolant void r e a c t i v i t y 2 f f e c t ( % 6 k / k )

- A l l device region voided

- Core and t a r g e t voi<?d

- Only core voided

Doppler r e a c t i v i t y c o e f f i c i e n t ( %idk/dT )

E f f e c t i v e delayed neutron f r a c t i o n

Bean swi tch ing of f t ime ( sec )

Bern equ iva l en t worth ( % B klkl?iW)

P a r t i c l e beam elignernent

Values

1: y o u s e l e c t Others o r t k r e zr2 L ~ Y s i g n i f i c m t modi f ica t ions f o r t h2 o r i g i i l r l l i b r a r y on g ino r e c t i n i a e s a d long-l ived ?P i s o t o p e s , p l ea se g i v e t k i r b r i e f d e s c r i p t i o n .

I 2 ) Calcula t ion method:

Neutronic c a l c u l a t i oz code:

N a e of code: 0 Diffusion 3 Transpor t 0 2D 3D

Nunber of energy group ~ s e d f o r c a l c u l a t i o n :

High and medium energy r e a c t i o n s c s l c u l s t i o n code:

C! N?ITC/J:\ERI i -AHBT O H E T C l K F A 2 0 Others

I f t h e r e a r e any s i g n i f i c ~ n t d i f f e r z n c e s between your code and usua l ly 2326 ~ ~ 6 2 , p l ea se g ive t h e i r b r i e f d e s c r i p t i o n .

Burnup c a l c u l a t i o n code :

Nme of code: C] O D O 1D Cl 2D 3D

I f t h e r e a r e any s i g n i f i c a n t d i f f e r e n c e s i n c a l c u l a t i o n a l model and/or burnup chain model of minor a c t i n i d e s and long-l ived F P i so topes , p l ea se give c i e i r b r i e f d e s c r i p t i o n .

I I

I I I

1 I

,

1

( 5 ) I n t e g r a l experiments r e l a t e d t o t r m s m u t a t i o n :

1) S p a l l a t i o n i n t e g r a l experiment:

Name of a c c e l e r a t o r f a c i l i t y :

I r r a d i a t e d n ino r a c t i n i d e 2nd i z r g e t n a t e r i a l i s o t o p e s :

- Purpose of exper imsnts : Measurement of nuc l ea r d a t 2 i n t h e nigh energy r eg ion V e r i f i c a t i o n of s p a l l a t i o n codes React ion r a t e , r e a c t i o n ~ r o d u c t s , p a r t i c l e ensroby s p z c t r ~ a , energy depos i t i on Others

I f you s e l e c t Others , p l e a s e g ive t h e i r b r i e f d e s c r i p t i o n .

2 ) Experiment us ing s u b c r i t i c a l assembly :

Name of f a c i l i t y : Cl Fas t neutron f i e l d El Thermal neut ron f i e l d

. Measured minor a c t i n i d e and long- l ived F P i s o t o p e s :

I

i Kind of measurements: C1 Rezct ion r a t e r a t i o

O Oihs r s

I f you s e l e c t Others , p l e a s e g ive t h e i r b r i s f d e s c r i p t i o n

3 ) Future programme, i f you have:

Thank you f o r your hea r ty c o o ? ? r a t i o ~ I I

If ORM-B / I N F O R M A T I O N O N T R A N S M U T A T I O N S T R A T E G I E S

Name :

A f f i l i a t i o n : -,

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T e l : Fax :

I . O b j e c t i v e s of Your t r a n s m u t a t i o n s i r a t e g i e s , s u c h a s r a d i o t o x i c i t y r e d u c t i o n i r ,

t h e d e e p s t o r a g e ( s o u r c e - t e r m r e d u c t i o n ) , r a d i o t o x i c i t y r e d u c t i o n a t t h e outcome i n t h e b i o l o g i c a l s p h e r e ( l o n g - t e r m r i s k r e d u c t i o n ) , mass r e d u c t i o n i n a s t r a i g h t f o r w a r d way , a d o t h e r s .

i n you r n a t i o n a l programme I3 i n y o u r o r g a n i s a t i o n p r o g r a m ? your own o p i n i o n

11. Your f u e l c y c l e c o n c e p t w i t h t r ~ n s a u t a t i o n and c s s o c i a t e d p e r t i t i o a i n g , i n c l u d i n g an i l l a s t r a t i o n i f nossib!?.

111. Your o p i n i o n on o v e r a l l e f f e c t o f i n t r o d u c i n g p a r t i t i o n i n g and t r a n s m u t a t i o n i n t o t h e n u c l e a r f u e l c y c l e :

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Thank you f o r your h e a r t y c o o p e r a t i o n ! I

C. The proposed nuclear reaptor-based and accelerator-based transmu-

tation concepts are summarized in this report, based on reply to

the request of information on proposed concept to about 30 spe-

cialist working for transmutation in the United State of America,

European countries, Russia and Japan. The task force reviews the

concepts,and recommends followings;

(1) Benchmarks for a set of common transmutation system

model on calculation methods,

(2) Effect of trasmutation rate on radiotoxicity reduction.

(3) Safety features of transmutation systems, and

(4) Nuclear data of transmutation nuclides.

Members of Task Force

S. Matsuura (JAERI)

P.A. Landyro (ENEA)

K. Abrahams (ECN)

H . Gruppelaar (ECN)

D . H . Rief (CEC)

J.P. Grouiller (CEA)

Argiori (ENEA)

H . Yoshida ( J A E R I )