Long-Lived Fission Product Transmutation Studies

W. S. Yang,* Y. Kim,† R. N. Hill, T. A. Taiwo, and H. S. Khalil

Argonne National Laboratory, Nuclear Engineering Division

9700 S. Cass Avenue, Argonne, Illinois 60439

Received March 26, 2003

Accepted June 30, 2003

Abstract –A systematic study on long-lived fission products (LLFPs) transmutation has been performed

with the aim of devising an optimal strategy for their transmutation in critical or subcritical reactor

systems and evaluating impacts on the geologic repository. First, 99Tc and 129I were confirmed to have

highest transmutation priorities in terms of transmutability and long-term radiological risk reduction.

Then, the transmutation potentials of thermal and fast systems for 99Tc and 129I were evaluated by

considering a typical pressurized water reactor (PWR) core and a sodium-cooled accelerator transmutation of waste system. To determine the best transmutation capabilities, various target design and loading

optimization studies were performed. It was found that both 99Tc and 129I can be stabilized (i.e., zero net

production) in the same PWR core under current design constraints by mixing 99Tc with fuel and by

loading CaI2 target pins mixed with ZrH2 in guide tubes, but the PWR option appears to have a limited

applicability as a burner of legacy LLFP. In fast systems, loading of moderated LLFP target assemblies in

the core periphery (reflector region) was found to be preferable from the viewpoint of neutron economy

and safety. By a simultaneous loading of 99Tc and 129I target assemblies in the reflector region, the

self-generated 99Tc and 129I as well as the amount produced by several PWR cores could be consumed at

a cost of ;10% increased fuel inventory. Discharge burnups of ;29 and ;37% are achieved for 99Tc and 129I target assemblies with an ;5-yr irradiation period.

Based on these results, the impacts of 99Tc and 129I transmutation on the Yucca mountain repository

were assessed in terms of the dose rate. The current Yucca Mountain release evaluations do not indicate a

compelling need to transmute 99Tc and 129I because the resulting dose rates fall well below current

regulatory limits. However, elimination of the LLFP inventory could allow significant relaxation of the

waste form and container performance criteria, with associated economic benefits. Therefore, some development of either specialized waste form or transmutation target for the LLFP is prudent, especially

considering the potential accumulation of large LLFP inventory with sustained use of nuclear energy into

the future.

I. INTRODUCTION

As an approach to minimize the nuclear waste sent

to the geologic repository and hence to increase the effective repository capacity, the potential to recycle and

incinerate the radiotoxic transuranics ~TRU! in nuclear

reactor systems has been studied.1,2 If the TRU inventory is reduced significantly by deploying TRU transmuters, the long-lived fission products ~LLFPs! will likely

dominate the long-term dose associated with radionuclide release from the geologic repository. To reduce

the long-term dose, it has been suggested to transmute

LLFPs into stable or short-lived isotopes by neutron capture in reactor systems.

Various studies of the LLFP transmutation in reactor

systems have previously been performed,3–10 ranging from

simple estimations of transmutability based on physical

properties of various LLFPs to actual irradiation tests in

experimental reactors. However, to the best knowledge

of the authors, most of the studies were scoping-type

analyses to estimate the potential transmutation rates in

specific systems with limited consideration of design and

safety constraints. In addition, the overall net benefits of

*E-mail: wyang@rae.anl.gov

†Permanent address: Korea Atomic Energy Research Institute, Yusong, P.O. Box 105, Daejeon 305-600, Korea.

NUCLEAR SCIENCE AND ENGINEERING: 146, 291–318 ~2004!

291
the LLFP transmutation relative to the direct geologic

disposal have not been investigated systematically.

To devise an optimal strategy for LLFP transmutation in critical or subcritical reactor systems and to evaluate its impacts on the geologic repository, a systematic

study on LLFP transmutation has been performed. First,

the open literature information was collected and thoroughly assessed, to decide the order of transmutation

priorities for various LLFPs. Then, the transmutation potentials of thermal and fast systems for 99Tc and 129I

were evaluated for a typical pressurized water reactor

~PWR! core and a sodium-cooled accelerator transmutation of waste ~ATW!system.11 To identify favorable transmutation strategies, various target design and loading

optimization studies were performed under various design constraints. The impacts of LLFP loading on safety

parameters and core performance parameters were also

evaluated. Finally, based on these results, the impacts of

99Tc and 129I transmutation on the Yucca mountain repository12 were assessed in terms of the released dose

rate.

The organization of this paper is as follows: Section II provides a summary on the general characteristics

of important LLFPs and their transmutability in reactor

systems. The LLFP transmutation potentials of a typical

PWR and a sodium-cooled ATW system are discussed in

Secs. III and IV, respectively. In Sec. V, the impact of

LLFP transmutation on repository performance is described. The conclusions and future work recommendations are presented in Sec. VI.

II. TRANSMUTATION CHARACTERISTICS OF

LONG-LIVED FISSION PRODUCTS

Transmutation of toxic LLFPs makes sense only if

the total radiotoxicity of nuclides being sent to a geologic repository is reduced significantly. This requires

that the target FP should interact with neutrons at a much

higher rate than natural decay. In addition, the radiotoxicity of the product isotopes ~including the activation

products of the materials used in the target fabrication!

should be much smaller than the initial radiotoxicity. In

order to identify the candidate LLFPs and appropriate

target forms, basic property data in the open literature

pertaining to LLFP transmutation were collected and

assessed.

The radiological properties of major LLFPs and their

production rates in a typical PWR core are summarized

in Table I. The radiotoxicity values were estimated using

the International Commission on Radiation Protection

~ICRP! data, and the production rates were calculated

for a typical three-batch PWR core using the ORIGEN2

code.13 From Table I, it can be seen that 90Sr and 137Cs

have a large radiotoxicity due to their relatively short

half-life. Except for 107Pd, all beta decays accompany

subsequent gamma decays. Palladium-107 has the smallest radiotoxicity ~measured in sievert per gram! since its

beta energy is only 30 keV.

The transmutation rate of an LLFP in a neutron field

can be characterized by the transmutation half-life defined as the time needed to incinerate half of its initial

mass by neutron capture. This transmutation half-life can

be defined as

T102

tr 5 3.171 3 1028 ln 2

sgF @yr# , ~1!

where

sg 5 neutron capture cross section ~b!

F 5 neutron flux ~n0cm2{s!.

For effective LLFP transmutation, the transmutation

half-life T102

tr should be much smaller than the natural

decay half-life T102. To estimate the generic transmutation potential, transmutation half-lives were calculated

for typical fast and thermal systems using the ORIGEN2

code.

Table II compares these transmutation half-lives with

the natural decay half-lives. It can be seen that six nuclides ~94Nb, 99Tc, 107Pd, 129I, 135Cs, and 151Sm! are

transmutable in either thermal or fast reactors. Thus, these

are possible candidate LLFPs for reactor-based transmutation. Other nuclides are very difficult to transmute in

reactor systems because of their very small capture cross

section: Extremely high flux level would be required for

effective transmutation. Even though capture cross

TABLE I

Radiological Properties of Major LLFPs and

Production Rates in PWRs

Decay

Isotope Type

Half-Life

~yr!

Productiona

@kg0GW~thermal!{yr#Radiotoxicityb

~Sv0g!

79Se b2 6.5E14c 0.066 8.259

90Sr b2 29 6.07 1.269E15

93Zr b2 1.5E15 8.04 1.045

94Nb b2 2.0E14 8.1E26 1.410E11 99Tc b2 2.1E15 8.54 6.056E21 107Pd b2 6.5E16 2.34 1.048E23 126Sn b2 1.0E15 0.30 6.306

129I b2 1.6E17 1.96 2.696E21 135Cs b2 2.3E16 2.76 8.532E22 137Cs b2 30 10.65 4.190E14 151Sm b2 89 0.15 1.281E12

a3.2% 235U enrichment, 33 GWd0tonne U, 20-yr cooling. bICRP data.

cRead as 6.5 3 104

.

292 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
sections are higher in thermal systems than in fast systems, thermal systems do not provide a decisive advantage over fast systems because of their lower flux level.

However, in a given system, softening of the local energy spectrum usually provides better transmutation

efficiency.

It is generally recognized that isotopic separation of

LLFPs would be very difficult and costly. Thus, if an

LLFP requires isotopic separation for effective transmutation, it cannot be transmuted efficiently from a practical point of view. For the above six transmutable LLFPs,

the isotopic compositions and the corresponding capture

cross sections for a standard PWR spectrum are summarized in Table III. It can be seen that 135Cs strongly requires isotopic separation for effective transmutation

because of the much higher weight fractions and capture

cross sections of 133Cs and 134Cs; in direct transmutation

of elemental cesium, additional 135Cs would be produced from 133Cs and 134Cs by neutron capture. In addition, very strong gamma rays emitted by 137Cs would

make handling and isotopic separation of cesium very

difficult. Niobium-94 and 107Pd also need isotopic separation for efficient transmutation. On the other hand,

151Sm does not require isotope separation due to its very

large capture cross section, even though the weight fraction of 150Sm is quite large. For 129I, isotopic separation

is not strongly required because the 127I fraction is relatively small, and its capture product 128I quickly decays

TABLE II

Generic Transmutability of LLFPs in Fast and Thermal Neutron Fields

Capture Cross Sectiona Transmutation Half-Lifeb

Isotope

Fast

Neutron

Thermal

Neutron

Half-Life

~yr!

Fast

Neutron

Thermal

Neutron

Pure Isotope

Transmutability

79Se 0.002 0.33 6.5E14c 1.1E14 666 Nontransmutable

90Sr 0.01 0.08 29 2.2E13 2.7E13 Nontransmutable

93Zr 0.09 1.03 1.5E15 244 213 Questionable

94Nb 0.22 4.22 2.0E14 100 52 Transmutable 99Tc 0.45 9.32 2.1E15 49 24 Transmutable 107Pd 0.53 2.79 6.5E16 42 79 Transmutable 126Sn 0.007 0.03 1.0E15 3.1E13 7.3E13 Nontransmutable

129I 0.35 3.12 1.6E17 63 70 Transmutable 135Cs 0.07 2.48 2.3E16 314 89 Transmutable 137Cs 0.01 0.03 30 2.2E13 7.3E13 Nontransmutable 151Sm 2.09 660 89 11 0.33 Transmutable

aORIGEN2 library ~fast: oxide fuel liquid-metal fast breeder reactor; thermal: standard PWR!. bThermal flux 5 1.0 3 1014, fast flux 5 1.0 3 1015 ~n0cm2{s!. cRead as 6.5 3 104

.

TABLE III

Isotopic Composition of Transmutable LLFPs

Isotope

Isotopic Composition

~wt%; capture cross section!a

Nb 93Nb ~90.0; 0.41!, 94Nb ~10.0; 4.22!

Tc 98Tc ~8E24, not available!,

b 99Tc ~99.9992; 9.32!

Pd 104Pd ~18.75; 0.66!, 105Pd ~32.84; 3.79!, 106Pd ~15.54; 0.28!, 107Pd ~19.47; 2.79!, 108Pd ~13.38; 7.08!, 109Pd ~0.02 ; 0.28!

I 127I ~22.98; 4.89!, 129I ~77.02; 3.12!

Cs 133Cs ~76.41; 10.6!, 134Cs ~0.292; 11.3!, 135Cs ~16.83; 2.48!, 137Cs ~6.47; 0.03!

Sm 150Sm ~63.90; 14.8!, 151Sm ~2.55; 660!, 152Sm ~26.27; 74.5!, 154Sm ~7.28; 1.51!

aORIGEN2 capture cross section for a standard PWR spectrum.

bRead as 8 3 1024

.

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 293

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
to 128Xe with a half-life of 25 min. Technetium-99 requires no isotopic separation.

Another important factor to be considered in LLFP

transmutation is the impact on the geologic repository. This

impact of LLFP transmutation on the repository can be

measured by the reduction of the leakage-dose risks to biosphere. It is well known that 129I and 99Tc are major contributors to the biosphere release dose rates because of their

high geochemical mobility. Technetium-99 and 129I are

well soluble in the groundwater, and they are hardly sorbed

by the geological rocks. The LLFP leakage rate from a

repository highly depends on its design of engineering

barriers and the geological properties.

Table IV summarizes the transmutation needs of the

six transmutable LLFPs in terms of radiotoxicity, halflife, mobility, and inventory and the isotope separation

requirement. Note that the dose risk is based on the current evaluation for direct disposal of spent fuel. In terms

of reduction of the collective leakage-dose risks ~products of toxicity and mobility! in the repository, the transmutation priority can be ranked in the order of 129I, 99Tc,

135Cs, and 94Nb. The transmutation need is not high for

151Sm due to its relatively short half-life and for 107Pd

due to its low toxicity and low repository dose rate. However, for the effective transmutation of 135Cs, isotopic

separation is strongly required, and a very high neutron

flux ~.1.0 3 1016 n0cm2{s! is necessary even in a thermal spectrum. Consequently, without isotopic separation, the practical candidates for LLFPs transmutation

would be 129I and 99Tc.

The most problematic isotopes, 99Tc and 129I, can

be transmuted to short-lived or stable isotopes via a

single neutron capture reaction. A neutron capture of

99Tc forms 100Tc, which shortly decays to stable 100Ru

with a half-life of 15.8 s. The daughter nuclide 100Ru is

transmuted only to stable or short-lived nuclides by further neutron capture. The capture cross sections of 99Tc

and 100Ru are inversely proportional to the neutron velocity in the thermal energy range. Technetium-99 has a

large resonance at 5.6 eV and a series of smaller resonances at higher energies: The resonance integral is

;340 b. This indicates that 99Tc might be effectively

transmuted in the epithermal energy range. Since 100Ru

has a much smaller capture cross section than 99Tc, 99Tc

would be mainly transmuted to 100Ru with slow conversion rates into 101Ru or 102Ru. This also implies that 99Tc

transmutation results in a reactivity increase, although it

might be quite small.

Since the half-lives of the unstable nuclides 128I,

130I, 131I, and 132I are very short, 127I and 129I are primarily converted by neutron capture to 128Xe and 130Xe,

respectively. The conversion of 127I to 129I would be

very small since 128I quickly decays to 128Xe. As a result, both 127I and 129I are transmuted into stable xenon

isotopes through neutron capture reactions. In the thermal energy range, the capture cross sections of 127I, 129I,

and 130Xe are inversely proportional to the neutron velocity, but 129I has a much larger capture cross section

than the other two isotopes. Iodine-129 has a series of

resonances, but they are much smaller than those of 99Tc;

the resonance integral is ;36 b. Xenon-130 has a significantly smaller capture cross section than that of 129I.

Thus, 129I target burnup in a thermal spectrum would

increase the reactivity.

With respect to the transmutation of 129I and 99Tc in

nuclear reactors, several target material forms for 99Tc

and 129I transmutation have been studied. Typically, metallic targets have been proposed for 99Tc transmutation,

and iodide forms ~NaI, CaI2, CeI3, PbI2, CeI3, YI3, etc.!

have been considered for 129I. Technetium can also be

loaded homogeneously by comingling it with the fuel,

unlike iodine, which is difficult to mix with the fuel.

Irradiation tests performed in the High Flux Reactor at

Petten concluded that there is no technical limitation to

the use of metallic technetium and that sodium iodide

seems to be the best candidate among three iodine target

forms considered ~NaI, CeI3, and PbI2! from the point of

view of comparability with 15-15 titanium stainless steel

cladding and with reprocessing technology.6–10 It has also

been reported that sodium iodide ~NaI! and calcium iodide ~CaI2! are desirable target forms for 129I in terms of

their chemical characteristics.10 Also, CeI3 and YI3 have

favorable properties such as good thermodynamic stability and high iodine density, but fabrication is difficult

because of their instability toward oxygen and moisture

~i.e., easy formation of oxyiodide!. For the NaI form, a

potential concern is that the sodium may melt when it is

liberated from the target as the iodine is transmuted to

xenon gas.

III. TRANSMUTATION OF 99Tc AND 129I IN A

PRESSURIZED WATER REACTOR

Since capture cross sections are high in thermal

systems compared to fast systems, it is intuitively

TABLE IV

Transmutation Needs and Isotope Separation Requirement

of Transmutable LLFPs

Transmutation Needs

Isotope Toxicity Half-Life

Repository

Impact Inventory

Isotope

Separation

Requirement

94Nb High High Low Very low Weak

99Tc Medium High High High No 107Pd Low High Low Medium Strong 129I Medium High Very high Medium Weak 135Cs Medium High Medium Medium Strong 151Sm High Low Low Low Weak

294 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
considered that LLFPs could be efficiently incinerated

in a PWR core. However, it seems that there is no general agreement on the LLFP transmutation capability of

PWRs. Some previous studies reported that the transmutation of 99Tc and 129I in a PWR core is not promising

due to a low transmutation rate and a large required inventory, in spite of relatively large capture cross sections

of the LLFPs ~Ref. 3!. This is mainly due to the low flux

level in the PWR core. On the other hand, a recent related study indicated that PWRs have some potential to

transmute the LLFPs effectively.6 However, these previous studies employed relatively simple models and

methods, and focused on limited aspects of the LLFP

transmutation. Therefore, the full LLFP transmutation

potential of PWRs is reevaluated in a detailed and consistent manner in this study.

Current PWR cores are well optimized in terms of the

neutronic and thermal-hydraulic design point of view.

Therefore, it is desirable to select LLFP loading options

in such a way that the fuel assembly and core design

changes due to LLFP loading be minimized. In particular,

the maximum loading feasible should be searched for

under the design criteria on the feed enrichment, power

peaking, fuel burnup, and the available space limit for

LLFP loading. For example, the amount of LLFP loading

in the conventional UO2 fuel PWR is constrained by the

uranium enrichment limit of 5.0 wt% 235U typically imposed by enrichment plants. More importantly, the negative impacts of LLFPs on the core performance and safety

parameters ~e.g., cycle length, reactivity swing, reactivity coefficients, rod worths, etc.! should be minimized.

Under these considerations, various loading options were

investigated for 99Tc and 129I, which were identified to

have the highest transmutation priorities in Sec. II.

Metal pins were selected as the 99Tc target form based

on previous studies. As discussed in Sec. II, NaI and

CaI2 are reasonable iodine target forms in terms of their

chemical characteristics. However, in the case of NaI,

the sodium is liberated from iodide when iodine isotopes

are transmuted to xenon, and hence its explosive reaction with water coolant might be a problem at abnormal

conditions. Therefore, CaI2 was selected as the 129I target form. To avoid the expense of isotopic separation,

the iodide target is directly formed with the elemental

iodine extracted from the spent nuclear fuel, which includes both 129I ~77%! and 127I ~23%! FPs.

In this section, the maximum LLFP transmutation

capability of PWRs under typical design constraints is

investigated. The analysis is mainly focused on the standard UO2 fuel assembly, and UO2 and PuO2 mixedoxide ~MOX! fuel assembly cases are included for

comparison purposes. In Sec. III.A, the calculation models and methods are described. LLFP target pin designs

and PWR-based LLFP stabilizer and burner designs are

discussed in Secs. III.B and III.C, respectively, and the

impacts of LLFP loading on major safety parameters are

discussed in Sec. III.D.

III.A. Computational Models and Methods

Typically, PWR core analyses are done in two steps:

Assembly calculations are first performed to generate

the few-group homogenized group constants, and then

full core neutronics and depletion calculations using these

few-group data follow. Based on previous work that demonstrated good agreement between assembly-level and

full-core analyses for calculating the heavy metal mass

flows in a homogeneous MOX assembly,14 assembly calculations with reflective boundary conditions were utilized to estimate the LLFP transmutation performance of

PWRs. Since the average behavior of a PWR core can be

accurately analyzed by assembly-level analyses when the

core is loaded with the same type of fuel assemblies, it is

considered that the core-averaged LLFP transmutation

rates are estimated with reasonable accuracy by these

assembly calculations.

For a standard 17 317 Westinghouse-type PWR assembly with 24 control rod guide tubes and a central

instrument tube, unit assembly calculations were performed using the WIMS8 assembly analysis code15 with

a 172-group master library. The assembly configuration

is shown in Fig. 1, and the assembly design parameters

are provided in Table V. The reference PWR conditions

for these calculations are also summarized in Table VI. It

is assumed that all the fuel pins have the same fuel enrichment, as is the usual case in the Westinghouse fuel

assembly design. For MOX fuel assembly analyses, a

Fig. 1. Configuration of 17 3 17 Westinghouse-type fuel

assembly.

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 295

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
full-MOX core was assumed with the same fuel management scheme as the UO2 fuel core. A typical light water

reactor ~LWR! discharge isotopic composition was used

for the PuO2 component of the MOX fuel.

In this study, the once-through fuel cycle and a typical three-batch fuel management scheme was assumed

with an 18-month cycle length and an 80% capacity factor. The power density was set to 37.5 W0g HM, which is

a typical value of conventional PWRs. The fuel management scheme results in a fuel assembly residence time of

1314 days and an average discharge burnup of ;49 GWd0

tonne HM. In assembly depletion calculations, the pinwise depletion option was used, and hence the material

density and fluxes are evaluated for individual pins at each

depletion time step over the entire irradiation period.

Thirty-one depletion time steps were used with variable

time intervals; small time steps were used for the early

depletion stage to accurately account for the xenon buildup.

In WIMS8 calculations, a solution option based on

the method of characteristics was used to solve the twodimensional neutron transport problem. Although the

master library is given for a 172-group structure, the

two-dimensional transport calculation is performed with

28-group data, which are condensed from a detailed spectral calculation with the 172-group library. It was confirmed that the 28-group results agree well with those of

a direct 172-group calculation. The fuel enrichment required for the envisioned fuel management scheme was

determined using the linear reactivity model. Assuming

that the core leakage effect is 3.5% Dk in a PWR core,

the kinf value should be 1.035 at 876 effective full-power

days ~EFPD! ~equivalent critical burnup state! for the

three-batch, 18-month fuel cycle.

In this work, the transmutation rates of LLFPs were

evaluated via assembly depletion analysis without accounting for the soluble boron concentration, and the burnable absorbers. In general, the PWR core is maintained

critical by adjusting the soluble boron concentration, and

a significant amount of excess reactivity is usually controlled by the burnable absorbers. Therefore, for accurate

fuel assembly depletion, both the critical soluble boron

concentration and the burnable absorbers need to be considered during the depletion calculations. Such thermal

neutron absorbers have two opposite impacts on LLFP

transmutation. Since they make the neutron spectrum

slightly harder, the effective capture cross sections ~and

hence transmutation rates! of LLFPs are slightly reduced. At the same time, for a fixed power density, they

increase the flux level relative to the no-absorber case,

thereby enhancing the LLFP transmutation rates. Preliminary analyses showed that the net effects of these absorbers on the LLFP transmutation rate are slightly

positive and were neglected in this study for simplicity.

III.B. Long-Lived Fission Product

Target Pin Design Study

Aiming at maximizing the 99Tc and 129I transmutation rates in PWRs under the current design criteria, LLFP

target design studies have been performed. In order to

minimize the fuel assembly and core design changes,

these studies were focused on the LLFP target pin design, with no assembly geometry modification. The option of employing special LLFP target assemblies~distinct

from fuel assemblies! was not considered for the PWR

study; this option is not favorable since the interior target pins would be exposed to relatively low neutron fluxes

because of large shadowing effects of periphery pins.

In order to characterize LLFP transmutation in PWRs,

various loading options of 99Tc and 129I that do not require the fuel assembly geometry change were considered using metallic 99Tc and CaI2 ~density of 4.52 g0cm3 !

target forms; for the 129I target, elemental iodine with an

isotope vector of 77% 129I and 23% 127I was assumed.

Since both 99Tc and 129I have relatively large capture

cross sections in the thermal and epithermal energy range,

a solid cylindrical target form reduces the effective capture cross sections significantly because of spatial selfshielding effects. This is especially true for 99Tc since it

has a large capture resonance at 5.6 eV. In order to enhance the capture reactions by reducing the spatial selfshielding effects, the LLFP target materials need to be

loaded as dilutely as possible. In addition, to take full

TABLE V

Design Data of Reference Fuel Assembly

Parameter Value

Pin pitch ~cm! 1.26

Pin diameter ~cm! 0.95

Cladding thickness ~cm! 0.0572

Fuel pellet radius ~cm! 0.4095

Guide-tube outer radius ~cm! 0.6120

Guide-tube thickness ~cm! 0.0405

Interassembly gap ~cm! 0.08

Active assembly height ~cm! 365.76

TABLE VI

Reference Computational Conditions

for PWR Fuel Assembly

Parameter Material

Density

~g0cm3 !

Temperature

~K!

Fuel pellet UO2 ~MOX! —a 900

Cladding Zirconium 6.50 630

Coolant Water 0.7053 583

Guide tube Type 303 stainless steel 7.95 600

aComposition-dependent fuel density.

296 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
advantage of large thermal capture cross sections of 129I

and 99Tc, further softening of the neutron spectrum by

extra moderation might be desirable. Accordingly, thin

annular target forms with homogeneous mixing with fuel

or moderator were also considered. In order to investigate the effects of further spectrum softening, ZrH2 ~density of 5.61 g0cm3 ! was chosen as extra moderator since

it has a superior moderating power6 and a good irradiation experience in TRIGA reactors.

For 99Tc, five types of loading options were

considered:

1. homogeneous mixing with fuel

2. 99Tc coating of fuel pellet

3. steel-clad annular target loaded in the guide tube

4. annular target material enclosing ZrH2 moderator placed in the guide tube

5. cylindrical target mixed with ZrH2 loaded in the

guide tube.

For the 129I target, only options 3, 4, and 5 were considered; the options loading 129I inside the fuel cladding

were excluded because of the production of gaseous xenon. Figure 2 shows the schematic views of these LLFP

loading options.

The self-shielding effects could be minimized by mixing 99Tc with fuel homogeneously since 99Tc could be

loaded very dilutely. In addition, the high thermal conductivity of 99Tc may enhance the thermal conductivity

of fuel. On the other hand, formation of TcO2 might be a

potential problem since it sublimes at 1000 K, even though

no serious problem is anticipated when a small amount

of 99Tc is mixed with fuel. Thin 99Tc film coating of fuel

would reduce the self-shielding effects, but fuel fabrication might be costly. Based on the previous 99Tc irradiation tests, it is expected that there would be no material

compatibility problem between 99Tc and cladding. When

annular target pins are loaded in guide tubes, the available space for LLFP loading is limited because a significant fraction of the locations must be reserved for control

rod insertion. In addition, these options cannot reduce

the self-shielding effects as effectively as the first two

dilute loading options. It is also not clear whether LLFP

target materials are compatible with ZrH2 moderator. On

the other hand, LLFP target pins could be fabricated easily, and they can be managed independently of the fuel.

For the above five LLFP loading schemes, WIMS8

calculations were done for both UO2 ~4.5 wt% 235U! and

full-MOX ~10.9 wt% PuO2! fuel assemblies. Figures 3

and 4 show the effective one-group capture cross sections of 99Tc and 129I, respectively, for the various loading options. The neutron capture reaction rates of 99Tc

and 129I are presented in Figs. 5 and 6, respectively.

The results in Figs. 3 and 4 show that the capture

cross sections of 99Tc and 129I increase monotonically as

the LLFP loading decreases; the capture cross section of

99Tc increases almost exponentially with loading decrease, while the 129I cross section increases linearly.

The increase of the 99Tc cross section is mainly due to

the reduced self-shielding of the large resonance peak at

5.6 eV. When the 99Tc loading is significant, the loading

options of homogeneous mixing with fuel and 99Tc coating of the fuel pellet give slightly larger cross sections

than all others. As the loading decreases, the resonance

self-shielding effect decreases, but the spectral softening

effect due to extra moderation increases. As a result, for

a relatively small 99Tc loading, the cross sections of the

cases using additional moderator become larger than the

others due to increased spectral softening effect.

In the case of 129I, spectral softening is the main

cause of cross-section increase, and the resonance selfshielding effect is relatively insignificant. For the same

loading of 129I, the effective cross sections can be very

different depending on the spectrum as shown in Fig. 4.

However, the cross-section variation is relatively small

compared with that of 99Tc since the neutron spectrum

of a typical PWR assembly is already very soft. The use

of ZrH2 moderator increases the capture cross section

due to the higher moderating power of ZrH2 relative to

water; a homogeneous mixture of target and ZrH2 provides a slightly larger value than the annular target enclosing the moderator because of smaller self-shielding

effect. Comparing UO2 and MOX fuel assemblies, the

harder spectrum in a MOX assembly makes the capture

cross sections of both 99Tc and 129I significantly smaller

than those of the UO2 assembly, by ;15 and ;50%,

respectively.

In general, these results indicate that the target isotope loading needs to be minimized in order to maximize

the effective capture cross section. Because the fractional

consumption is directly proportional to the effective onegroup cross section, it is desirable to minimize the target

material loading within limits imposed by transmutation

goals. However, as shown in Figs. 5 and 6, the capture

rates decrease monotonically as the target material loading decreases, even though the effective cross sections

increase. As a result, it is necessary to maximize the

target material loading under various design constraints

if the goal is to maximize the absolute consumption rate.

III.C. Pressurized Water Reactor–Based

Long-Lived Fission Product

Stabilizer and Burner

The effective capture cross sections and capture reaction rates discussed in Sec. III.B indicate that in thermal systems, the fractional consumption and absolute

consumption rate are somewhat contradictory objectives

that cannot be optimized at the same time. Therefore, the

targeted mission of LLFP transmutation needs to be defined first, and the performance objective needs to be

specified accordingly. In this work, two approaches for

LLFP transmutation were postulated. In an LLFP burner

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 297

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
aimed at burning the accumulated LLFP as quickly ~or

much! as possible, the maximum consumption rate is of

interest. Here one uses the maximum loading of target

material that would not degrade core performance and

safety parameters. Alternatively, the reactor core could

be used only for preventing the buildup of ~stabilizing!

the LLFP such that only reprocessing losses are sent to

the repository. In this LLFP stabilizer, the minimal LLFP

loading is determined such that LLFP mass produced is

equal to that consumed in a given recycle stage. At assembly discharge, the fuel and LLFP target pins are reprocessed, and the recovered LLFPs are fabricated into

target pins ~or mixed with fuel! and reintroduced.

The feasibility of employing a PWR as an LLFP

stabilizer or burner was evaluated by performing depletion calculations for the 99Tc and 129I transmutation rates.

The initial loading in the stabilizer case was determined

such that the LLFPs produced in the core by fission are

balanced by the amount destroyed by neutron capture. In

the burner case, since the amount of LLFP destroyed

Fig. 2. Target designs for LLFP loading in PWRs.

298 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
increases with the initial loading, the LLFP loading was

maximized under the uranium enrichment limit of 5.0

wt% 235U. A specific power density of 37.5 W0g HM

was assumed, and an 18-month cycle length ~80% capacity factor! was used. In each case, the fuel enrichment or

the LLFP loading was adjusted such that kinf is 1.035 at

876 EFPD ~equivalent critical burnup state! for a threebatch fuel management scheme. For the cases in which

LLFPs are loaded in the guide tube, it was assumed that

40% of the locations were reserved for control rods.

Therefore, a stabilizer assembly should destroy the LLFPs

produced in the assemblies where LLFP targets are not

loaded, in addition to the self-generated LLFPs. For a

UO2 fuel assembly, these excess quantities amount to

;0.35 kg0assembly for 99Tc and ;0.083 kg0assembly

for 129I, respectively. In the case of a MOX fuel assembly, ;0.35 kg of 99Tc and ;0.11 kg of 129I need to be

additionally destroyed per assembly.

The resulting transmutation performances are summarized in Table VII. For the cases of homogeneous

mixing of 99Tc and fuel, the production rate was approximated by the production rate of the corresponding Tccoated fuel pellet case for the UO2 fuel and by that of the

base case for the MOX fuel, since the WIMS8 code

does not distinguish the 99Tc nuclides produced by fission from the initial target nuclides. Table VII shows that

Fig. 3. Effective one-group capture cross section of 99Tc

in PWR target assemblies.

Fig. 4. Effective one-group capture cross section of 129I

in PWR target assemblies.

Fig. 5. Capture reaction rate of 99Tc in PWR target

assemblies.

Fig. 6. Capture reaction rate of 129I in PWR target

assemblies.

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 299

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
in a UO2 core, ;20 and ;12% of the initial loading 99Tc

and 129I, respectively, are consumed at discharge. Additionally, the results indicate that a PWR LLFP stabilizer

for 99Tc and0or 129I would be feasible with an increase in

uranium enrichment, relative to the base case with no

LLFP loading. The uranium enrichment is generally less

than the 5% limit, except for the case using 99Tc target

pins mixed with ZrH2. In this case, the required uranium

enrichment is slightly higher than 5.0 wt% because of the

larger amount of LLFP loading required to compensate

for the assemblies not loaded with target pins. However,

this enrichment is artificial because of the assembly model

used in this study, and it would be slightly reduced in a

full-core calculation. This is similarly valid for the 129I

results since 129I targets are loaded in guide tubes only.

The potential of using a PWR as an LLFP burner is

limited because the initial LLFP loading is constrained

by the uranium enrichment. Nonetheless, results indicate

that 99Tc and 129I can be burned with support ratios @defined as the number of conventional light water reactors

~LWRs! that a burner supports, i.e., the ratio of megawatt~thermal! from LWRs to megawatt~thermal!from a

burner# of ;0.5 and ;3, respectively, but the low discharge burnup of 129I would require a large number of

reprocessing stages for complete burning. The amount

of LLFPs incinerated could be increased by maximizing

the LLFP loading within the imposed enrichment limit,

but this would reduce the cycle length and the LLFP

discharge burnup.

In the MOX core, the discharge burnups of 99Tc and

129I are smaller by ;19 and ;50%, respectively, than

those for the UO2 core, due to the hardened neutron spectrum ~see Figs. 7 and 8!. A more efficient LLFP transmutation could be achieved in the MOX-fueled core by

increasing the neutron moderation. For example, the highmoderator-ratio16 ~HMR! reactor concept for a fullMOX core would be helpful in enhancing the LLFP

transmutation performance. Nevertheless, it is expected

TABLE VII

Transmutation Performance of 99Tc and 129I in PWR Assemblies*

Case

235U or Pu

Fraction of

Heavy Metal

~wt%! LLFP

Initial

Loading

~kg!

Production

~kg!

Net

Consumption

~kg!

LLFP

Discharge

Burnup

~at.%! Remarks

UO2 Fuel Assembly, LLFP Stabilizer

Base case 4.38 99Tc 0.0 0.532 20.532 — No initial LLFP

129I 0.0 0.123 20.123 —

Tc mixed with fuel 4.79 99Tc 2.568 0.540a 10.000 21.0 0.49 wt% Tc

Tc target mixed with ZrH2 5.17 99Tc 4.588 0.540 10.356 19.5 95.4 vol% ZrH2

Tc-coated fuel pellet 4.79 99Tc 2.511 0.540 20.004 21.7 8.8-mm Tc

CaI2 target mixed with ZrH2 4.58 129I 1.714 0.121 10.088 12.2 93.4 vol% ZrH2

Tc mixed with fuel and CaI2 4.99 99Tc 2.655 0.540a 20.001 20.3 0.50 wt% Tc

target mixed with ZrH2 129I 1.777 0.120 10.086 11.6 93.2 vol% ZrH2

UO2 Fuel Assembly, LLFP Burner

Tc mixed with fuel 5.00 99Tc 3.968 0.540a 10.246 19.8 0.75 wt% Tc

Tc-coated fuel pellet 5.00 99Tc 3.882 0.544 10.249 20.4 13.6-mm Tc

CaI2 target mixed with ZrH2 5.00 129I 5.515 0.120 10.471 10.7 78.9 vol% ZrH2

MOX Fuel Assembly, LLFP Stabilizer

Base case 10.8 99Tc 0.0 0.527 20.527 — No initial LLFP

129I 0.0 0.164 20.164 —

Tc mixed with fuel 12.8 99Tc 3.180 0.527b 10.013 17.0 0.60 wt% Tc

Tc mixed with fuel and CaI2 14.5 99Tc 3.256 0.527b 20.004 16.1 0.60 wt% Tc

target mixed with ZrH2 129I 4.835 0.165 10.125 6.0 81.5 vol% ZrH2

*Masses are per unit assembly, and the active assembly height is 365.76 cm.

aSelf-generated 99Tc from the Tc-coated fuel case. bSelf-generated 99Tc from the base case.

300 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
that the 99Tc and 129I transmutation performance in the

HMR MOX core would not be significantly higher. In

the same context, a TRU-loaded PWR core is expected

to have LLFP transmutation performance similar to that

of the typical PWR loaded with MOX.

Figures 7 and 8, respectively, compare the average

neutron spectra ~172-group! of UO2 and MOX assemblies with and without LLFP loading. For both UO2 and

MOX fuel assemblies, the neutron spectra of LLFPloaded assemblies are only slightly harder than that of

the base case because of thermal neutron captures in

LLFP targets. The additional moderator in the guide tube

does not affect the assembly average neutron spectrum

noticeably since it is confined only to the guide-tube

locations. Therefore, it is expected that the neutronic properties of the LLFP-loaded fuel assemblies are not very

different from those of the base case without LLFPs. In

Figs. 9 and 10, the kinf value is plotted over the irradiation period for various target configurations. It is observed that 99Tc and 129I transmutation in a PWR reduces

the reactivity swing noticeably. This is because 99Tc and

129I are respectively transmuted into Ru and Xe elements, which have significantly smaller capture cross

sections than their parent isotopes. The reduced reactivity swing indicates that 99Tc and 129I could be used as

burnable absorbers although they have relatively large

residual reactivity penalties due to limited discharge

burnups. The reduced initial excess reactivity in an LLFPloaded PWR may result in a reduced maximum soluble

boron concentration.

Fig. 7. Neutron spectrum in UO2 fuel target assemblies.

Fig. 8. Neutron spectrum in MOX fuel target assemblies.

Fig. 9. Infinite multiplication factor of UO2 fuel target

assemblies versus irradiation time.

Fig. 10. Infinite multiplication factor of MOX fuel target

assemblies versus irradiation time.

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NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
Both 127I and 129I are transmuted mainly into stable

isotopes 128Xe and 130Xe, respectively. Since these xenon isotopes are in gaseous state, the xenon gas pressure

buildup in an iodine target would be one of the limiting

factors for the iodine transmutation, if an iodine target is

placed in a sealed pin. Therefore, a preliminary evaluation of the xenon gas pressure buildup was performed

for a CaI2 target pin using the ideal gas law. The xenon

pressure buildup depends on the gas release rate from

the iodine target, but no data is available at the moment.

Therefore, for a conservative pressure evaluation, it was

assumed that all the xenon gas produced in the target is

released into the space between the target and cladding.

The results are presented in Fig. 11 for various burnup

levels of the iodine isotopes as a function of the volume

ratio of gas plenum to CaI2 target.

Figure 11 shows that a relatively large gas plenum is

required for a high iodine burnup. For example, consider

a case of homogeneous mixture of CaI2 and ZrH2 moderator loaded in the guide-tube location, for which the

iodine discharge burnup was shown to be ;10 to 12

at.% depending on the volume fraction of the moderator.

Assuming a 15.25-cm gas plenum as in the typical fuel

rod in PWRs, the volume fraction occupied by the target

mixture is ;0.93. If the moderator fraction is 0.8 in the

target mixture, the volume ratio of gas plenum to CaI2

target becomes ;0.38. In this case, the internal xenon

pressure is ;540 atm for a 12 at.% burnup of iodine. If

we take the allowable internal gas pressure of a typical

PWR fuel pin, which is ;157 atm similar to the system

pressure, as the maximum allowable xenon pressure, the

CaI2 volume should be smaller than 77% of the gas plenum volume for a 12 at.% iodine discharge burnup. For

the iodine target with an inner moderator, it would be

difficult to obtain the required gas plenum space for a

12 at.% discharge burnup. On the other hand, if an annular iodine target with inner void region is loaded in the

guide tube, a sufficient gas plenum space would be available for the 12 at.% iodine discharge burnup.

III.D. Impacts of Long-Lived Fission Product

Loading on Safety Parameters

In order for LLFPs to be routinely transmuted in

PWRs, their loading should not significantly degrade the

safety potential of the core. Major safety-related parameters were evaluated for several LLFP loading options

and compared with those of the standard UO2 fuel assembly. Specifically, the fuel temperature coefficient

~FTC!, moderator temperature coefficient ~MTC!, soluble boron worth, and control rod worth were estimated.

Generally, the reactivity coefficients of a PWR core are

evaluated for the critical boron concentration. With assembly calculation methodology, however, the critical

conditions are not typically searched. Thus, a constant

soluble boron concentration of 500 ppm was assumed

for this analysis. The reactivity coefficients were calculated for the base case without LLFP loading and for two

LLFP-loaded assemblies: the 99Tc burner case and the

99Tc0129I stabilizer discussed in Sec. III.C. The 99Tc burner

was loaded with Tc by homogeneous mixing of Tc with

fuel, and Tc loading was maximized within the fuel enrichment limit of 5 wt% 235U. The 99Tc0129I stabilizer

was simultaneously loaded with Tc and I by additionally

placing the CaI2 targets mixed ZrH2 in the guide-tube

locations.

Figure 12 compares the burnup-dependent FTCs of

the LLFP-loaded assemblies with that of the standard

assembly. It is observed that FTCs of the LLFP-loaded

assemblies are slightly less negative. This is mainly due

Fig. 11. Xenon gas pressure in CaI2 target pin for various

iodine burnup levels ~680 K!.

Fig. 12. Burnup-dependent fuel temperature coefficient

at 900 K.

302 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
to the higher fuel enrichments of the LLFP-loaded assemblies than the standard one. Figure 13 shows that the

MTC of 99Tc0129I stabilizer is comparable to the base

case MTC, while the 99Tc burner has more negative MTC

values. For the Tc-burner case, the MTC becomes more

negative by ;14%, and it is considered that this amount

of change would not cause any safety problem. This enhanced MTC may help the core safety when the soluble

boron concentration is high at the beginning of cycle.

Figure 14 shows that the soluble boron worths are

slightly reduced in the LLFP-loaded cases compared with

the standard assembly. This boron worth reduction is

again due to the neutron spectrum hardening caused by

LLFP loading. The 99Tc0129I stabilizer has the smallest

boron worth because of the reduced coolant volume. Although the boron worth is slightly smaller in an LLFPloaded assembly, the critical boron concentration in the

core should not noticeably increase because the burnup

reactivity swing is sufficiently smaller in the LLFPloaded core, as shown in Fig. 9. Generally, the soluble

boron worth is a function of the boron concentration,

being larger at low concentration. However, it is expected that the trends shown in Fig. 14 would not change

noticeably at different boron concentrations.

The control rod worth is compared in Fig. 15 for the

standard and Tc-mixed fuel cases only, since the control

rod cannot be inserted when LLFPs are loaded in the

guide-tube locations. It can be seen that the control rod

worth of the LLFP-loaded assembly is marginally decreased ~by ;1%! again due to the hardened neutron

spectrum. Because this change is small, the worth of the

control rods might not be an issue in the LLFP-loaded

PWR core. Taking into account all these results, it is

concluded that 99Tc and0or 129I can be transmuted in a

UO2-fueled PWR core without any problem from the viewpoint of reactivity coefficients as long as the LLFP loading is bounded by the current uranium enrichment limit

of 5 wt% 235U.

Another important consideration for reactor safety

is the power peaking factor. Since the power distribution

is not changed when LLFP is uniformly mixed with fuel,

the LLFP-induced perturbation in the power distribution

was conservatively assessed by considering an assembly

loaded with the mixture of 20% Tc and 80% ZrH2 by

volume in the guide-tube locations. In this case, ;20 kg

of 99Tc is loaded in an assembly, which is much larger

than the 99Tc mass in the 99Tc burner case. Figure 16

compares the normalized pinwise power distribution at

zero burnup with that of the standard assembly. It is

observed that the power distribution is actually flatter in

Fig. 13. Burnup-dependent moderator temperature coefficient at 583 K.

Fig. 14. Burnup-dependent boron worth at 500 ppm. Fig. 15. Burnup-dependent B4C control rod worth.

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NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
the Tc-loaded assembly than in the standard one. As shown

in Fig. 16, the peak power in the standard assembly is

1.057, while it is 1.047 for the Tc-loaded case. This is

because the power peaking usually occurs in the vicinity

of the guide-tube cells in the standard fuel assembly. In

the PWR fuel assembly without burnable absorbers, the

power distribution changes only slightly over the depletion period. Furthermore, in the current PWR fuel assembly designs, any LLFP power peaking problem could be

mitigated by employing burnable absorbers. Therefore,

it can be said that a significant LLFP loading in the guide

tube does not adversely affect the assembly power distribution. However, the LLFP loading in the guide tube

would increase the heat flux from the guide tube to coolant due to additional gamma and neutron heating of the

target and hence affect the departure from nucleate boiling ratio ~DNBR! for the guide-tube channels. Although

the LLFP loading in the guide tube is not expected to

reduce the DNBR significantly since it reduces the neighboring pin powers as shown in Fig. 16, detailed thermalhydraulic analyses need to be performed to evaluate its

effect on the minimum DNBR.

IV. TRANSMUTATION OF 99Tc AND 129I IN AN

ACCELERATOR TRANSMUTATION OF

WASTE SYSTEM

Previous studies have shown that fast systems are

more attractive transmuters because of the high neutron

flux and the surplus neutrons that could be used for target transmutation.3,17 The high leakage fraction also provides more flexibility for LLFP loading. To take advantage

of higher capture cross sections for thermal neutrons, the

neutron spectrum in LLFP target assemblies can also be

moderated by employing metal hydrides such as ZrH2

and CaH2 ~Refs. 3 through 8!. These moderated target

assemblies can be positioned in the inner core to take

full advantage of the higher flux level of fast systems.

~However, it requires additional design measures to reduce the local power peaking in nearby fuel pins caused

by moderation in the target assembly.! Alternatively, neutron leakage from a fast reactor core can be utilized by

locating LLFP target assemblies at the core periphery

~e.g., in the reflector region!.

To devise an optimal strategy for transmuting LLFPs

in fast systems, trade studies were performed for an 840-

MW~thermal! sodium-cooled ATW system. Using metallic 99Tc and CaI2 target forms and ZrH2 moderator,

assembly design optimization studies were first done to

maximize the LLFP transmutation performance and to

mitigate the local power peaking problem in neighboring fuel assemblies caused by moderation in the target

assembly. Then, using optimized LLFP target assembly

designs, full-core depletion calculations were carried out

for various target loading options including the loading

of moderated target assemblies in the inner core and the

reflector region and the homogeneous mixing of LLFP

with fuel. For simultaneous loading of 99Tc and 129I target assemblies, further assembly design optimization was

performed to equalize the 99Tc and 129I support ratios to

the TRU support ratio for the burner design, ensuring

that neither the long-lived TRUs nor LLFPs accumulate.

In Sec. IV.A, the methods and results of the assembly design optimization studies are presented. The LLFP

loading option studies are discussed in Sec. IV.B along

with the LLFP transmutation performances.

IV.A. Long-Lived Fission Product Target

Assembly Design Studies

In the case of repeated recycle, the loss rate of

LLFPs to the repository is basically determined by the

discharge burnup ~i.e., the fractional consumption at

discharge! and the recovery factor per recycle pass. The

loss rate monotonically decreases as the discharge

burnup and recovery factor increase. Based on the ATW

roadmap study,18 the ~currently targeted! recovery factor

for LLFP elements is ;95%, while it is over 99.9% for

TRU elements. Considering this relatively low recovery

factor, the discharge burnup of LLFP target assemblies

should be maximized under various design and operation constraints.

For fixed total flux level and residence time, the

LLFP discharge burnup is primarily determined by the

effective one-group capture cross sections. To take advantage of higher LLFP capture cross sections for thermal neutrons, the LLFP target assemblies are generally

moderated with metal hydrides such as ZrH2 and CaH2.

Fig. 16. Power distributions of standard and Tc-loaded

assemblies.

304 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
In a moderated LLFP target assembly, as the moderator

loading fraction increases, the effective capture cross section ~i.e., relative transmutation rate! increases due to

the enhanced moderation and reduced self-shielding effects of LLFP isotopes. On the other hand, the increase

of moderator loading fraction results in a reduced LLFP

loading and hence a reduced absolute transmutation rate.

As a result, it is expected that there exists an optimum

moderator loading fraction that provides a good compromise between the goals of maximizing both the relative

and absolute LLFP transmutation rates.

Spectral optimization studies were first performed

for moderated 99Tc and 129I target assemblies. To account for the local heterogeneity effects, calculations were

carried out with the VIM continuous Monte Carlo code.19

A variety of thermal neutron filter concepts were examined to reduce the local power peaking in neighboring

fuel assemblies caused by target moderation.

IV.A.1. Spectral Optimization

For spectral optimization studies, a reference target

assembly design was developed based on previous studies and preliminary analyses. In this design, target pins

composed of annular technetium or iodine target material surrounding ZrH2 moderator within an HT-9 cladding are arranged in a triangular array inside a HT-9

hexagonal duct, as shown in Fig. 17a. Table VIII summarizes the assembly design parameters that were selected to be compatible with the ATW fuel assembly

design.11 Metallic technetium was assumed for the 99Tc

target. As the iodine target material, a CaI2 compound

was selected because of relatively high melting points of

CaI2 ~7838C! and calcium ~8428C!. ~NaI is expected to

melt when the sodium is liberated from iodine due to the

transmutation, and this might be a problem for achieving

a high discharge burnup.! Isotopic separation of 129I was

excluded, and an iodine vector of 77% 129I and 23% 127I

~typical of PWR spent fuel! was used for the 129I target.

For this reference assembly design, the optimum moderator fraction was first sought for each of the technetium and iodine target assemblies by adjusting the

moderator volume fraction inside the cladding. For a

given moderator volume fraction, two variations of

target configuration were also investigated: cylindrical

target surrounded by annular moderator and homogeneous mixture of LLFP target and moderator. To account for local heterogeneity effects, Monte Carlo

calculations were performed with VIM. In order to examine the spectral and LLFP resonance self-shielding

effects without introducing excessive computational

Fig. 17. ~a! LLFP assembly configuration and ~b! seven-assembly VIM computational model.

TABLE VIII

Design Parameters of Reference LLFP Target Assembly

Number of target pins per assembly 271

Pin diameter ~cm! 0.74422

Cladding thickness ~cm! 0.05588

P0D ratio 1.1969

Active length ~cm! 106.68

Duct outside flat-to-flat ~cm! 15.7099

Duct wall thickness ~cm! 0.3937

Interassembly gap ~cm! 0.4318

Assembly pitch ~cm! 16.1417

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 305

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
burden, two seven-assembly models were developed for

target assembly loadings in the inner core and reflector

regions; an LLFP target assembly is surrounded by six

fuel assemblies in the model for the inner core loading,

but it is surrounded by three fuel assemblies and three

HT-9 reflector assemblies in the model for the reflector

region loading as shown in Fig. 17b. The fuel assembly

was loaded with a Zr-based metal dispersion fuel, and

the TRU isotopic vector of LWR spent nuclear fuel was

used. In both models, one of the fuel assemblies was

evaluated in detail to assess the local power peaking due

to the thermal neutron leakage from the LLFP assembly.

With the reflective boundary conditions imposed, the VIM

calculations were performed with the ENDF0B-V crosssection library at 300 K.

Using these seven-assembly computational models,

Monte Carlo VIM calculations were carried out for various moderator volume fractions by adjusting the target

thickness. In order to separate out the spectrum softening effects, the VIM calculation for each target thickness

was repeated by setting moderator density to zero. The

results are summarized in Table IX and in Figs. 18, 19,

and 20. Table IX shows the effective one-group capture

cross sections of 99Tc, 127I, and 129I along with the power

peaking factor in the fuel assembly. The results show

that the capture cross sections of the three isotopes monotonically increase as the moderator volume increases.

This is due to the reduced resonance self-shielding and

the softened neutron spectrum. Even in the zero moderator density case, the capture cross section increases as

the target thickness decreases due to the reduced spatial

self-shielding effects. Compared to 99Tc, 129I shows a

substantially larger increase of the effective capture cross

section because of its higher thermal capture cross section. The increase of the 99Tc cross section is largely due

to the reduced self-shielding of the large resonance peak

at 5.6 eV. ~The resonance integral is ;340 b for 99Tc,

while it is ;36 b for 129I.! These energy effects are illustrated in Fig. 18, which shows the normalized capture

reaction rate distributions over a 27-group structure for a

60% moderator volume. The thermal group reaction rate

is much larger than those of other groups in the 129I

target. For the 99Tc target, the thermal reaction rate is

TABLE IX

Moderator Effects on Effective Capture Cross Sections and Power Peaking Factor

99Tc Target CaI2 Target

ZrH2

Volume

Fraction

One-Group

Capture Cross

Section of 99Tc

~b!

Power Peaking

Factor in

Fuel Assembly

One-Group

Capture Cross

Section of 127I

~b!

One-Group

Capture Cross

Section of 129I

~b!

Power Peaking

Factor in

Fuel Assembly

LLFP Target Loading in Reflector Region

0.0 0.43 ~0.43a ! 1.09 0.71 ~0.71a ! 0.37 ~0.37a ! 1.07

0.2 0.78 ~0.49! 1.08 2.19 ~0.83! 1.20 ~0.43! 1.54

0.4 1.23 ~0.57! 1.28 3.00 ~0.99! 2.11 ~0.51! 2.14

0.6 1.94 ~0.70! 1.64 3.70 ~1.24! 3.31 ~0.61! 2.74

0.8 3.39 ~0.91! 2.22 4.52 ~1.71! 4.98 ~0.79! 3.32

0.9 4.99 ~1.12! 2.81 5.18 ~2.12! 6.11 ~0.92! 3.67

LLFP Target Loading in Inner Core

0.6 1.84 1.87 3.54 3.10 3.44

aZero moderator density case.

Fig. 18. Groupwise capture reaction rate in LLFP assemblies with a 60% moderator volume.

306 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
slightly larger than that of the 24th group, which contains a large resonance. In the case of inner core loading,

the capture cross sections are slightly reduced relative to

the reflector loading case due to harder neutron spectrum.

As shown in Table IX, the ZrH2 moderator in the

LLFP target assembly induces a large local power peaking in the adjacent fuel assemblies. The power peaking

problem is more pronounced in the iodine target assembly case. This is mainly due to the larger mean-free-path

of thermal neutrons in CaI2 than in technetium; the thermal neutron mean path is about three times larger in CaI2

than in technetium since the iodine atom density in the

CaI2 target is ;3.8 times smaller than the 99Tc atom

density in the metallic technetium target. The extra neutron moderation by the calcium nuclides also contributes. Compared to the reflector loading case, the power

peaking of the inner core loading case is considerably

higher for the 129I target assembly and marginally higher

for the 99Tc target assembly.

In Figs. 19 and 20, the capture reaction rates and

relative transmutation rates normalized to the unmoderated case are plotted as a function of the moderator volume fraction. It can be observed that a moderator

fraction in the vicinity of 0.4 provides the maximum

capture rate, that is, the maximum absolute transmutation rate ~LLFP mass transmuted per unit time!. With

this optimal moderator volume fraction, the capture rate

is increased by ;15% for 99Tc and by ;150% for 129I

relative to the unmoderated case. The 129I capture rate is

increased much more significantly by moderation because the capture cross section of 129I is greatly enhanced by the ZrH2 moderator due to its higher thermal

cross section. Comparing the capture rates of the normal

and zero moderator cases, most of these capture rate

increases can be ascribed to the moderator. These results

show that a desired absolute transmutation rate can be

achieved with a much smaller LLFP loading by employing moderator. For a fixed amount of LLFP loading per

target assembly, the maximum transmutation performance is achieved by arranging the LLFP target pins in

such a way that the moderator volume is maximized.

This suggests that total volume for the target and moderator should be maximized subject to assembly thermal

design constraints.

Figures 19 and 20 also show that the relative transmutation rate ~fraction of LLFP transmuted per a unit

time! increases monotonically with the moderator volume fraction. As a result, the LLFP loading needs to be

minimized in order to obtain the maximum target discharge burnup, which is desired to reduce the LLFP loss

to the repository. On the other hand, in order to maximize the LLFP incineration, the moderator volume fraction needs to be selected in the vicinity of 0.4 as discussed

above; in the region of moderator volume fraction above

0.4, the absolute transmutation rate decreases monotonically with the moderator volume fraction. Furthermore,

the power peaking in nearby fuel assemblies also increases monotonically as the moderator volume fraction

increases. Considering these three conflicting performance indices, it is believed that a moderator volume

fraction of 0.6 to 0.7 is a good compromise.

For the 60% moderator volume fraction, two variations of target configuration were also investigated:

~a! central cylindrical target surrounded by annular moderator and ~b! homogeneous mixture of LLFP target and

moderator. The effective one-group capture cross sections are compared in Table X. In terms of the relative

transmutation rate, the homogeneous mixture provides

the best performance ~due to the minimized self-shielding

effect!, increasing the rate by ;5%, and the central

cylindrical target gives the worst performance, but the

Fig. 19. Transmutation rates of 99Tc versus moderator volume fraction ~reflector loading case!.

Fig. 20. Transmutation rates of 129I versus moderator volume fraction ~reflector loading case!.

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 307

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
differences are small. Furthermore, the power peaking

problem is exacerbated in the homogeneous mixture and

the annular moderator options. In addition, it is not clear

at this moment whether the homogeneous mixing concept is technically feasible from the fabrication0irradiation

point of view.

For the same moderator volume fraction in the LLFP

target assembly, these target designs ~Fig. 17a! generally

provide larger LLFP capture cross sections due to the

reduced self-shielding effects, compared to previous studies3,8 where distinct LLFP target and moderator pins are

employed either with a uniform pin diameter or with an

irregular pin diameter ~larger moderator pin diameter!.

Therefore, in terms of the effective LLFP capture cross

sections, the LLFP target assembly based on the annular

target concept could be considered as a near-optimal design. However, the material compatibility between LLFP

targets and ZrH2 moderator needs to be confirmed, although no serious compatibility issue is anticipated.

IV.A.2. Thermal Neutron Filter

As discussed in Sec. IV.A.1, the moderated LLFP

target assembly substantially enhances the LLFP transmutation rates. On the other hand, the moderation in the

target assembly also results in an increased local power

peaking in the nearby fuel assemblies. To mitigate the

power peaking, a thicker duct wall has been proposed.4

However, this option is not attractive from the neutron

utilization viewpoint. To reduce the thermal neutron leakage from the LLFP assembly to the neighboring fuel

assemblies, a 99Tc plate could be placed between the

duct wall and the target pins.8 This latter scheme has an

advantage in that the leaking thermal neutrons can be

utilized to transmute 99Tc in the thermal neutron filter,

although its transmutation rate is relatively lower compared to the interior targets. In this work, a variant of the

latter option was examined by replacing the target pins

in the outer rings with unmoderated 99Tc pins along the

boundary between the LLFP and fuel assemblies. For the

129I target assembly, the CaI2 target pins in two outer

rings were replaced with cylindrical 99Tc target pins. In

the case of the 99Tc target assembly, the internal moderator was removed from the target pins in the outermost

ring. Figure 21 shows the configurations of the 99Tc and

129I target assemblies with thermal neutron filters.

Using the seven-assembly VIM model for the reflector loading case, the effects of the 99Tc thermal neutron

filter were investigated for 99Tc and 129I target assemblies. For comparison, each calculation was repeated by

replacing the 99Tc pins with HT-9 pins. The resulting

local power peaking factors are presented in Table XI.

By employing the unmoderated 99Tc pins as the thermal

neutron filter, the initial local power peaking factor was

reduced from 1.64 to 1.43 for the 99Tc target assembly

and from 2.74 to 1.40 for the 129I target assembly. On the

other hand, the effective capture cross sections were

slightly reduced because of reduced moderation, as expected. The results also show that HT-9 pins are not as

efficient as 99Tc pins as the thermal neutron filter, especially for the 129I target assembly; for sufficient filtering

of thermal neutrons, more 129I target pins would need to

TABLE X

Effective Capture Cross Sections of Different Target Configurations*

99Tc Target CaI2 Target

Target Configuration

One-Group

Capture Cross

Section of 99Tc

~b!

One-Group

Capture Cross

Section of 127I

~b!

One-Group

Capture Cross

Section of 129I

~b!

Annular LLFP target surrounding cylindrical moderator ~reference! 1.94 3.70 3.31

Central cylindrical LLFP target surrounded by annular moderator 1.87 3.46 3.30

Homogeneous mixture of LLFP and moderator 2.01 3.93 3.37

*Moderator volume fraction 5 60%.

TABLE XI

Local Power Peaking of Fuel Assembly Neighboring to

LLFP Target Assembly with Thermal Neutron Filter*

Power Peaking in Neighboring

LLFP Fuel Assembly

Burnup

~at.%!

Thermal

Neutron

Filter 99Tc Target CaI2 Target

0 No filter 1.64 2.74

HT-9 1.49 2.04

99Tc 1.43 1.40

40 99Tc 1.46 1.55

*Moderator volume fraction 5 60%.

308 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
be replaced by the HT-9 pins, penalizing the 129I transmutation performance.

By neutron capture, technetium and iodine isotopes

are converted mainly into 100Ru and 130Xe, respectively.

Since the capture cross sections of these product isotopes are significantly smaller than those of target isotopes, the power peaking is likely to increase with target

burnup. Therefore, the power peaking factors were estimated for a 40 at.% burnup of 99Tc, 127I, and 129I. To get

a conservative estimate, it was assumed that 127I and 129I

are converted into void and 99Tc is converted into 101Ru,

which has a slightly smaller thermal capture cross section than 100Ru. As shown in Table XI, the high LLFP

burnup increases the power peaking slightly. However,

this power peaking increase is considered tolerable based

on the conservative peaking factor evaluation.

The power peaking factors presented in Table XI

provide conservative estimates for the LLFP assembly

loading in the reflector region because a reflective boundary condition was imposed. In the actual situation, a steep

flux gradient exists across the fuel, target, and reflector

assemblies that would likely reduce the power peaking

in the adjacent fuel assembly. In addition, the LWR spentfuel TRU vector used in these calculations has a high

239Pu fraction ~.50%!, enhancing the thermal neutron–

induced power peak; in an equilibrium core, the 239Pu

fraction would be significantly depressed, helping reduce the peaking factor. With appropriate coolant flow

allocation, the power peaking factors shown in Table XI

can be tolerated since the power densities in the peripheral fuel assemblies are much lower than the core average power density. On the other hand, the power peaking

needs to be reduced further when LLFP target assemblies are loaded in the inner core.

IV.B. Long-Lived Fission Product

Transmutation Performance of an

Accelerator Transmutation of Waste System

Using the LLFP target assembly designs discussed

in Sec. IV.A, the 99Tc and 129I transmutation capability

of a sodium-cooled ATW system11 was investigated by

performing full-core depletion calculations for various

target loading options including the moderated target assembly loading in the inner core and in reflector regions

and a homogeneous mixing of LLFP with fuel. ATW

system design parameters are based on the studies in

Ref. 11. For simultaneous loading of 99Tc and 129I target

assemblies, further assembly design optimization was

performed such that all 99Tc and 129I produced in conventional PWRs supported by an ATW system can be

incinerated in the same ATW system without being accumulated ~i.e., to equalize the 99Tc and 129I support

ratios to the TRU support ratio!.

The LLFP transmutation performances were evaluated by REBUS-3 ~Ref. 20! equilibrium cycle analyses.

Region-dependent 27-group cross sections were generated using the MC2-2 code21 based on ENDF0B-V data.

Since MC2-2 is not designed to generate the thermal

group data, the thermal group ~E , 0.414 eV! cross

sections for the moderated LLFP target assemblies were

estimated for the seven-assembly models shown in Fig. 17

using the VIM code based on the ENDF0B-V library.

Fig. 21. Thermal neutron filters for 99Tc and 129I target assemblies.

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 309

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
Flux calculations were performed with triangular-Z models using the finite difference diffusion theory option.

The core region is modeled with two “enrichment” zones

with a higher TRU0Zr ratio in the outer zone. Fuel regions are modeled with batch-averaged compositions for

a seven-batch fuel management scheme without fuel shuffling. Target regions are also modeled with batch-averaged

compositions over the target assembly lifetime constrained by the peak fast fluence limit on the cladding.

IV.B.1. Long-Lived Fission Product

Transmutation Objectives

The main design objective of the sodium-cooledATW

system is to consume TRU as efficiently as possible. To

limit the burnup reactivity swing, a half-year cycle ~135

EFPD! is adopted. The charged fuel contains the TRUs

recovered via recycle from the dischargedATW fuel, supplemented by LWR-dischargeTRU to make up for theTRU

consumed by fission. During recycle and refabrication, it

is assumed that all TRU discharged are recovered and

5% of the rare-earth FPs are carried over with the recycled TRU. A schematic configuration of the ATW core

is shown in Fig. 22. The detailed core design characteristics of the sodium-cooled ATW can be found in Ref. 11.

Unless special systems dedicated to the LLFP incineration are employed, simultaneous transmutation of both

TRU and LLFP ~in the appropriate proportion! appears

to be highly desirable. In this respect, the LLFP transmutation goal was defined as achieving 99Tc and 129I

support ratios equal to the TRU support ratio. In order to

estimate these support ratios, the 99Tc, 129I, and TRU

production rates in a typical PWR and the ATW system

were calculated using the ORIGEN2 code. A UO2 fuel

composition ~typical of a PWR! with 3.2 wt% 235U enrichment was depleted up to 33 GWd0tonne U burnup,

and a TRU fuel composition was depleted up to 270

GWd0tonne HM burnup, which is the typical fuel discharge burnup of the ATW system. The results are summarized in Table XII; assuming that the PWR and ATW

system capacity factors are 80 and 74%, respectively, the

TRU support ratio of the 840-MW~thermal! ATW system is estimated to be 3.2. Therefore, in order to achieve

the same support ratios while accounting for LLFP production in the ATW system, 22.8 kg of 99Tc and 5.6 kg of

129I need to be consumed per year ~i.e., 11.4 kg of 99Tc

and 2.8 kg of 129I per cycle! in the target assemblies.

IV.B.2. Heterogeneous Transmutation

The heterogeneous transmutation approach was first

investigated by focusing on 99Tc transmutation. Two loading options were analyzed. In the first option, 12 moderated 99Tc target assemblies were loaded into positions

R4 and R6 of the reflector ~see Fig. 22! and their symmetric counterparts. In the second option, six target assemblies were loaded in the O3 position within the core

and its symmetric counterparts.

In the reflector loading case, the R4 and R6 positions

were selected because they have higher flux levels than

the other reflector positions. The moderated 99Tc target

assembly with the 99Tc thermal neutron filter was used;

the 28 target pins at the interface between the fuel and target assemblies are composed of the annular 99Tc target

without moderator, and the other target pins are composed of the annular 99Tc target and ZrH2 moderator with

a 60% moderator volume fraction. The number of target

assemblies was determined such that the 99Tc transmutation rate in the target assemblies is roughly 11 kg0cycle.

In the case of the core interior loading, the location

of the target assembly should be determined in such a

way that a high transmutation rate can be achieved while

preserving the central source multiplication. For example, although the target assembly loaded in the innermost fuel ring near the spallation target would produce

the highest 99Tc transmutation rate due to the flux level

Fig. 22. Planar layout of 840-MW~thermal! sodiumcooled ATW core ~one-sixth core!.

TABLE XII

Production Rates of TRU, 99Tc, and 129I

in PWR and ATW Systems

Production Rate

@kg0GW~thermal!{yr#

Fuel Type TRU 99Tc 129I

PWR ~33 GWd0tonne U burnup! 107.9 8.6 2.0

ATW ~270 GWd0tonne HM burnup! 2369.1 7.1 2.2

310 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
being highest, it reduces the source multiplication efficiency significantly by directly absorbing the source neutrons. In this study, the O3 location was selected as a

compromise between the flux level and the negative impact on the source multiplication. To reduce the local

power peaking, solid cylindrical 99Tc pins were loaded

in the two outermost rings of the target assembly, and the

moderator volume fraction of the internal target pins was

increased to 70% to balance the transmutation rate.

The lifetime of the moderated target assembly would

likely be constrained by the fast neutron irradiation damage to the cladding ~fast fluence limit! and0or irradiation

damage to the moderator. Based on the worldwide operation histories of TRIGA reactors in which ZrH2 moderator is incorporated with enriched uranium metal in the

fuel elements, it is expected that the ZrH2 disintegration

would not be a main obstacle to achieving this irradiation time. Thus, the irradiation time of the 99Tc target

assembly was constrained by the fast fluence limit of

HT-9 cladding ~4 3 1023 n0cm2 !. From the REBUS-3

equilibrium cycle analysis, it was estimated that the 99Tc

target assembly could be irradiated for 14 consecutive

fuel cycles ~5.2 yr! before the fast fluence limit is reached.

Thus, the target burnup was determined by the REBUS-3

equilibrium cycle analyses with a target residence time

of 14 cycles and seven- and eight-batch fuel management schemes for the inner and outer zones, respectively.

The resulting 99Tc transmutation rates and selected

core performance parameters are presented in Table XIII.

It can be observed that the two 99Tc transmutation options could meet the required 99Tc support ratio by slightly

increasing the 99Tc loading. The in-core burning of 99Tc

provides a 25% higher relative transmutation rate than

the reflector loading case, due to the higher flux level.

Although the flux level is almost doubled in the in-core

burning case, the increase in the transmutation rate is not

proportionally high. This is attributed to the fact that the

effective capture cross section of 99Tc is significantly

smaller in the in-core case due to the larger loading per

assembly and the harder neutron spectrum than in the

reflector loading. For the in-core loading, the 46 at.%

discharge burnup might be a maximum value due to the

fluence limit. However, in the case of the reflector loading, the target discharge burnup could be significantly

increased with a longer irradiation time: The peak fast

fluence is only ;62% of the limit.

As shown in Table XIII, the in-core burning option

requires a slightly higher TRU fuel inventory than the

reflector loading case and hence results in 4% lower fuel

discharge burnup, although the absolute 99Tc transmutation rate ~kilogram0cycle! is comparable in both cases.

This is due to the relatively large neutron leakage in the

ATW core. The in-core burning requires additional neutrons to compensate the neutrons absorbed by the 99Tc

target, whereas the 99Tc target in the reflector is mainly

transmuted by the neutrons leaking out of the core. This

indicates that 99Tc placement in the reflector is preferable to the in-core burning from the neutron economy

point of view.

As discussed above, the local power peaking factor

of 1.43 for the reflector loading case is considered to be

acceptable since the power densities in the core peripheral fuel assemblies are much lower than the core average power density ~240 W0cm3 !. However, the in-core

burning option increases the average core power density

to 252 W0cm3 because of the reduced number of fuel

assemblies. This increased power density makes the local power peaking problem worse, and hence the local

TABLE XIII

Transmutation Performance Parameters of Heterogeneous 99Tc Loading Options in Equilibrium Cycle ATW Core

Tc Assembly Location

Parameter Core Reflector

Transmutation rate of 99Tc over 14 cycles ~1890 EFPDs!

BOECa inventory ~kg! 232.2 316.7

Average burnup ~%0cycle! 3.26 2.61

Average burnup ~kg0cycle! 9.91 10.1

Discharge burnup ~at.%! 45.6 36.5

TRU fuel

BOECa inventory ~kg! 3054 2911

Discharge burnup ~at.%! 25.8 26.9

Local power peaking in seven-assembly model 1.39 1.43

Initial total flux in Tc target assembly ~n0cm2{s! 2.66E115b 1.37E115

Peak discharge fast fluence in Tc assembly ~n0cm2 ! 3.95E123 2.46E123

aBOEC 5 beginning-of-equilibrium-cycle. bRead as 2.66 3 1015.

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 311

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
power peaking factor of 1.39 might not be acceptable. In

order to reduce the power peaking, the thermal neutron

leakage should be further reduced by decreasing the moderator volume and0or by increasing the number of peripheral filter pins. This modification would reduce the

99Tc transmutation rate. Alternatively, the power peaking problem can be mitigated by reducing the TRU fuel

enrichment in the adjacent fuel assemblies. However,

this makes the core design and fuel management more

complicated and increases the fuel fabrication cost.

The in-core LLFP transmutation has another potential problem from the safety point of view. In general,

the reactivity effect of the LLFP target assembly is quite

large when it is loaded in the middle of the core. As a

result, an accident involving the loss of target materials

could inject a large amount of reactivity into the core.

Meanwhile, the corresponding reactivity of an LLFP assembly in the reflector zone is much smaller, compared

with the in-core case. Consequently, any melting or ejection of an LLFP assembly from the reflector would be

more benign.

Transmutation of LLFP in the ATW reflector displaces steel reflectors, while still providing adequate

shielding. Therefore, the LLFP burning in the reflector

reduces the generation of steel activation products. This

is viewed as another advantage of reflector loading over

in-core target placement.

IV.B.3. Homogeneous Transmutation

Another option for 99Tc transmutation is direct mixing of 99Tc with fuel. In this so-called homogeneous

transmutation option, the high neutron flux in the fast

reactor core can effectively be used without adverse selfshielding of 99Tc. One disadvantage of this approach is

that the effective capture cross section is small due to the

hard neutron spectrum. This option is not considered for

iodine transmutation because of the xenon gas production, which could adversely affect the fuel performance.

As shown in Fig. 22, the sodium-cooled ATW core

employs two TRU enrichment zones ~low-enrichment

inner zone and high-enrichment outer zone! to flatten

the radial power distribution. Thus, if 99Tc is loaded uniformly in the whole core, the required outer zone TRU

enrichment would become too high. On the other hand,

since the flux level of the ATW core is significantly higher

in the inner zone than in the outer region, the 99Tc would

be more effectively burned in the inner zone; however,

too much loading of 99Tc in the inner zone may seriously

reduce the multiplication efficiency of the external neutron source. Taking these factors into account, 99Tc was

loaded in such a way that the loading density of 99Tc in

the inner core is about twice that of the outer zone. In

addition, 99Tc was not loaded in the outermost fuel ring

to enhance the burnup of 99Tc and flatten the radial power

distribution.

The 99Tc transmutation performance of ATW for

the homogeneous loading option was assessed by the

REBUS-3 equilibrium cycle analyses. Seven- and eightbatch fuel management schemes were utilized for the

inner and outer zones, respectively. The results are summarized in Table XIV. As discussed before, the 99Tc discharge burnup decreases as the 99Tc loading increases,

while the absolute transmutation rate ~kilogram0cycle!

increases with the 99Tc loading. By comparing Tables XIII

and XIV, it can be observed that for the same 99Tc loading, the homogeneous loading option results in transmutation performance inferior to the heterogeneous loading

options. The significantly lower 99Tc discharge burnup

is mainly due to the short irradiation time: 945 days in

the inner core and 1080 days in the outer core. The lower

cycle burnup is caused by the small capture cross section

resulting from the hard neutron spectrum. The homogeneous loading option also requires a larger TRU fuel

inventory than the heterogeneous loading in the reflector

for a comparable 99Tc transmutation rate.

IV.B.4. Simultaneous Transmutation

of 99Tc and 129I

In order to achieve support ratios for 99Tc and 129I

equal to the TRU support ratio, the simultaneous transmutation of both 99Tc and 129I in the ATW core was studied. Based on the above loading option studies, it was

decided to load moderated LLFP target assemblies in the

reflector region. The moderated LLFPtarget assembly designs with the 99Tc thermal neutron filter discussed in

Sec. IV.A.2 were used. The reflector positions that have

relatively higher flux levels were selected for LLFP assembly loading. Twelve 129I target assemblies were loaded

into the R4 and R6 positions and their symmetric counterparts ~see Fig. 22!, which have the highest flux levels

in the reflector region, and twelve 99Tc target assemblies

were loaded into the R2 and R8 positions and their symmetric counterparts, which have the next highest flux levels.

The resulting LLFP transmutation rates showed that

the 99Tc transmutation rate is sufficient to meet the support

ratio goal, but the 129I transmutation rate is not. Therefore,

TABLE XIV

Transmutation Performance Parameters of Homogeneous

99Tc Loading Option in Equilibrium Cycle ATW Core

99Tc Transmutation Rate

Initial

99Tc

Inventory

~kg! %0cycle kg0cycle

Discharge

Burnup

~at.%!

Initial

Fuel

Inventory

~kg!

Fuel

Discharge

Burnup

~at.%!

0 — — — 2744 28.3

45 2.90 1.30 19.8 2775 28.0

89 2.81 2.50 19.1 2813 27.7

176 2.68 4.72 18.2 2878 27.2

343 2.42 8.33 16.5 3014 26.1

312 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
to increase the absolute transmutation of 129I and to enhance the LLFP discharge burnups at the same time, the

above “standard” assembly designs were further optimized. As discussed in Sec. IV.A.1, the target discharge

burnup increases monotonically as the moderator volume fraction increases ~i.e., as the target volume fraction

decreases!. On the other hand, for a moderator volume

fraction near that of the standard design ~60%!, the absolute transmutation rates decrease as the target loading

decreases. As a result, in order to increase the discharge

burnup while satisfying the support ratio goal, it is necessary to increase the moderator volume while maintaining the 99Tc loading and increasing the 129I loading. The

only possible way to accomplish this without introducing additional target assemblies is to increase the total

target pin volume by reducing the coolant volume.

Accordingly, modified assembly designs were developed by reducing the pitch-to-diameter ~P0D! ratio of

target pins to its minimum value of 1.0, keeping the same

number of target pins per assembly. In these modified

designs, the target pin diameter was increased from the

reference value of 0.74 cm to 0.89 cm with the same clad

thickness. To maintain the 99Tc loading, the moderator

volume fraction within each 99Tc target pin was increased from 60 to 72%. Whereas, the 129I loading was

increased by applying a moderator volume fraction of

65%; this increases the loading per assembly by ;18%

relative to the reference target design. Although the moderator volume fractions inside target pins are not increased much relative to the reference designs, the

moderator volume fraction of the assembly is significantly enhanced, and hence the neutron spectrum is much

softer in these modified designs. In the modified 99Tc

target assembly, the thermal neutron filter located at the

core0reflector boundary was enhanced by replacing the

central void of the annular target pins with HT-9; this

increases the parasitic absorption of low-energy neutrons.

The resulting LLFPtransmutation rates and some core

performance parameters are summarized in Table XV. It

can be seen that the transmutation rates obtained with the

modified LLFPassembly designs are high enough to meet

the support ratio goals for both 99Tc and 129I. The modified assembly design increases the absolute and relative

transmutation rates of 129I by ;50 and ;10%, respectively, but it increases the 99Tc transmutation rates only

by ;1%. This is attributed to moderator trends observed

for the PWR LLFP transmutation analyses, which indicated that the effective capture cross section of 129I is enhanced more efficiently. It is noteworthy that the total flux

levels in the modified design are significantly lower than

those in the standard design, due to the softened neutron

spectrum. This reduced flux largely offsets the increase in

capture cross sections, particularly for 99Tc.When the modified LLFP designs are used, the TRU inventory must be

increased by ;1.6% mainly because of increased 129I capture rate. The local power peaking factors obtained from

the seven-assembly model VIM calculations are ,1.45 for

all cases. With appropriate coolant flow allocation, these

power peaking factors can be tolerated since the power

densities in the peripheral fuel assemblies are much lower

than the core average power density ~240W0cm3 !, as shown

in Fig. 23.

The LLFP target assemblies loaded in the reflector

are transmuted mainly by the neutrons leaking out of the

core. As a result, the impact on the core performance

parameters was generally minor. Compared to the base

design without any LLFP loading, the loading of modified LLFP assemblies in the reflector region increases

the TRU inventory only by ;10%, with a commensurate

8% reduction in the TRU discharge burnup. This increased inventory reduces the burnup reactivity swing

slightly. The peak power density still occurs within the

core ~near the interface between inner and outer cores in

Fig. 22!, and the global peaking factor is basically the

same as the base design.

In the modified LLFP assembly designs, the P0D

ratio of target pins is 1.0, and hence the coolant flow area

is reduced by a factor of ;4 relative to the ATW fuel

assembly. Thus, adequate cooling of target pins must be

considered. However, as shown in Fig. 23, the power

densities in the LLFP target assemblies are ,10% of the

core average power density ~240 W0cm3 !. Considering the

excellent thermal properties of sodium coolant, it is expected that the LLFPtarget pins can be adequately cooled.

However, the power density will be much increased compared to the typical steel reflector assemblies. Thus, appropriate coolant flow will need to be allocated.

V. IMPACTS OF LONG-LIVED FISSION

PRODUCT TRANSMUTATION ON YUCCA

MOUNTAIN REPOSITORY

Current repository performance evaluations of the

Yucca Mountain repository show 237Np as the largest

Fig. 23. Assembly average power densities ~watts per

cubic centimeter! of equilibrium cycle ATW with 99Tc0129I

target assemblies ~P0D 5 1.0! in the reflector region.

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 313

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
contributor to the long-term dose release rate with 99Tc

and 129I the major LLFP contributors. In addition, previous studies reveal that removal of 99.9% actinides from

the waste might reduce the long-term ~beyond 100 000 yr!

peak dose rate by a factor of 156 ~Ref. 22!. It was found

that 99Tc was the dominant contributor to the dose risk

when 99.9% actinides were simply removed without taking into account the conversion of the actinides to FPs.

Based on this study, the impacts of LLFP transmutation

on the Yucca mountain repository were analyzed by comparing three cases:

1. direct disposal of spent fuel

2. transmutation of TRU into FPs

3. transmutation of both TRU and LLFP.

The additional FP inventory generated by actinide

transmutation depends on the specific transmutation scenario and method. However, to estimate the impact, it

was assumed that among 99.9% actinides removed from

the spent fuel, all TRUs were transmuted into FPs, and

the uranium elements were stored for future utilization.

The production rates of 99Tc and 129I from actinide transmutation were approximated by their weight fractions

among the total FPs in the typical PWR spent fuel with

33 GWd0tonne U burnup and 25-yr cooling. That is, it

was assumed that 2.3 and 0.52% of TRU were converted

to 99Tc and 129I, respectively.

When the 99Tc and 129I are recycled and transmuted

into short-lived isotopes, the cumulative fractional loss

to the repository was estimated by summing the initial

spent-fuel processing loss and the cumulative loss

TABLE XV

Transmutation Performance Parameters of Simultaneous Loading of 99Tc and 129I

in Reflector Region of Equilibrium Cycle ATW Core

LLFP Assembly

Parameter

Standard

~P0D 5 1.1969! P0D 5 1.0

LLFP transmutation rate over 14 cycles ~1890 EFPDs!

BOECa inventory ~kg! 99Tc 531.1 530.1

129I 69.2 91.5

Average %0cycle 99Tc 2.08 2.10

129I 2.40 2.65

Average kg0cycle 99Tc 12.9 13.0

129I 2.00 2.99

Discharge burnup ~at.%! 99Tc 29.1 29.4

129I 33.7 37.1

TRU fuel

BOECa inventory ~kg! 2980 3027

Discharge burnup ~at.%! 26.4 26.0

Local power peaking in seven-assembly model 99Tc target 1.43 1.45

129I target 1.40 1.44 ~1.88b !

Initial total flux in LLFP assembly ~n0cm2{s! 99Tc target 9.28E114c 7.02E114

129I target 1.35E115 1.09E115

Peak discharge fast fluence in LLFP assembly ~n0cm2 ! 99Tc target 1.77E123 1.65E123

129I target 2.40E123 2.25E123

aBOEC 5 beginning-of-equilibrium cycle. bHT-9 rods are used instead of 99Tc rod as the thermal neutron filter. cRead as 9.28 3 1014.

314 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
resulting from the recycling of LLFP targets. Denoting

the fractional discharge burnup of LLFP and the fraction

of LLFP lost in recycle0refabrication by B and l, respectively, the cumulative fractional loss LT can be approximately represented as

LT 5 l

B 1 ~1 2 B!l . ~2!

Under these assumptions, the 99Tc and 129I inventories sent to the repository were evaluated for the three

specific cases. The initial inventories were obtained from

previous Yucca Mountain performance assessment studies. The total initial inventory consists of 7860 commercial spent nuclear fuel packages from PWRs and boiling

water reactors and 3890 codisposal packages for U.S.

Department of Energy spent fuel and high-level waste

glass. The inventories for the recycled LLFP case were

estimated using two reprocessing loss fractions and two

sets of discharge burnups. As the reprocessing loss fraction, a recommended value of 5% and a low loss of 1%

were used. In the first set of discharge burnups, the LLFP

discharge burnups in PWRs given in Sec. III were used,

i.e., 19 and 10 at.% for 99Tc and 129I discharge burnups,

respectively. In the second set, both discharge burnups

were assumed to be 30 at.%, which could be achieved by

using a special moderated LLFP target assembly in the

fast ATW system. Table XVI summarizes the resulting

99Tc and 129I inventories to be sent to the repository.

The impact of 99Tc and 129I transmutation on the

repository performance was estimated by evaluating the

dose rate resulting from use of water from a well located

;20 km from the repository, which is one of the key

parameters considered in the total system performance

assessment for the site recommendation of the Yucca

mountain repository. Since 99Tc and 129I are highly soluble in water, the release rates of these isotopes from the

engineered barrier system ~EBS! are proportional to their

inventories. Furthermore, since they are nonsorbing elements, the temporal shapes and magnitudes of the EBS

release rates and the release rate at the 20-km well are

very similar with a time delay of only one or two thousand years. Therefore, the dose rates of 99Tc and 129I for

different inventories were estimated in a simple way by

linearly combining the temporal shapes of individual

isotopes provided from previous repository analyses.22

In Fig. 24, both total dose rate and major dose contributors are reproduced for the direct disposal case. It

can be seen that the long-term dose rate is dominated by

237Np, and the dose contributions of 99Tc and 129I are

marginal, although 99Tc is the major dose contributor

during the early time period. Figure 25 shows that 99Tc

becomes the dominant contributor to the dose rate if

99.9% of actinides are removed from the spent fuel and

all the removed TRUs are transmuted into FPs. In this

TABLE XVI

Inventories of 99Tc and 129I to Be Sent to the Repository*

TRU and LLFP Transmutation

5% LLFP Loss Factor 1% LLFP Loss Factor

Isotope

Direct

Disposal

TRU

Transmutation

~No LLFP

Transmutation!

Low

Burnupa

High

Burnupb

Low

Burnupa

High

Burnupb

99Tc 6.37E17c 8.04E17 1.77E17 1.20E17 4.02E16 2.62E16

129I 1.45E17 1.82E17 6.37E16 2.72E16 1.64E16 5.93E15

*The inventories are given in grams.

a19 at.% for 99Tc and 10 at.% for 129I. b30 at.% for both 99Tc and 129I. cRead as 6.37 3 107

.

Fig. 24. Estimated dose rates at the 20-km well of the

direct disposal case.

LONG-LIVED FISSION PRODUCT TRANSMUTATION STUDIES 315

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
case, the peak dose rate is ;3 mrem0yr, which is only

;20% higher than the simple removal of 99.9% actinides. Figure 26 shows the impact on the dose rate of the

base 99Tc0129I transmutation scenario ~discharge burnups

of 99Tc and 129I are 19 and 10 at.%, respectively, with

recycle loss factor of 5%!. In this case, the 99Tc and 129I

transmutation reduces the peak dose rate from ;3 to

;1 mrem0yr. The factor of ;3 reduction in LLFP dose

rate could also be achieved using special tailored waste

forms. In fact, previous Argonne National Laboratory

studies on the metallic waste form ~where Tc naturally

resides for pyroprocessing! indicated superior leach resistance compared to typical glass waste forms. It can be

deduced from the inventories presented in Table XVI

that a higher LLFP discharge burnup and a lower reprocessing loss fraction would further reduce the dose rate

of the LLFPs significantly in terms of relative value.

Given a repository dose limit of 15 mrem0yr, the

current Yucca Mountain release evaluations do not indicate a compelling need to transmute 99Tc and 129I, where

the maximum doses rates are only 1 to 3 mrem0yr. However, it is important to note that the LLFPs are much

more mobile than the TRU species, and the dose rates for

the first 40 000 yr following closure are completely dominated by the LLFPs, and 99Tc in particular. The current

repository design relies on robust, long-lived waste containers that do not even begin to fail until the 10 000-yr

time frame. This allows retention of the bulk of the 99Tc

inventory until its elimination by decay ~10 000-yr halflife! becomes important. Furthermore, the 99Tc release

and transport is not limited by solubility; the “short term”

99Tc dose rates are directly tied to the container and waste

form failure rates. Therefore, elimination of the LLFP

inventory could allow significant relaxation of the waste

form and container performance criteria, with associated

economic benefits. In addition, future utilization of the

remaining uranium inventories could result in large increases in the LLFP waste inventory; thus, some development of either specialized waste form or transmutation

target is prudent.

VI. CONCLUSIONS

A systematic study on the transmutation of LLFPs

has been performed with the aim of devising an optimal

strategy for their transmutation in critical or subcritical

reactor systems and evaluating impacts on the geologic

repository. First, 99Tc and 129I were confirmed to have

highest transmutation priorities in terms of transmutability and long-term radiological risk reduction. Then,

the transmutation potentials of thermal and fast systems

for 99Tc and 129I were evaluated by considering a typical

PWR core and a sodium-cooled ATW system. To determine the best transmutation capabilities, various target

design and loading optimization studies were performed.

Finally, based on these results, the impacts of 99Tc and

129I transmutation on the Yucca mountain repository were

assessed in terms of the dose rate.

For 99Tc transmutation in a PWR core, thin 99Tc

film coating of fuel pellet, homogeneous mixing of 99Tc

with fuel, and mixture of 99Tc with ZrH2 are preferable

loading options from the incineration point of view. However, the Tc-bearing fuel options need to be evaluated

carefully in terms of fuel fabrication, performance, and

Tc recovery rate. In the case of 129I targets loaded in the

guide-tube locations, annular iodine targets filled with

ZrH2 moderator have the best transmutation performance. However, considering the xenon gas pressure

buildup, an annular target with inner void zone would be

preferable to other options.

Fig. 25. Estimated dose rates at the 20-km well of the

TRU transmutation case.

Fig. 26. Estimated dose rates at the 20-km well of the

TRU and LLFP transmutation case.

316 YANG et al.

NUCLEAR SCIENCE AND ENGINEERING VOL. 146 MAR. 2004
Both 99Tc and 129I can be stabilized in the same

PWR core at the cost of higher fuel enrichment. However, due to the enrichment constraint, it appears that

only one of them can be incinerated in a given burner

core with a support ratio of ;0.5 for 99Tc and ;3 for

129I. In a UO2 core, the discharge burnups of 99Tc and

129I are ;20 and ;12%, respectively, for an 18-month

fuel cycle with an 80% capacity factor. For a MOX core,

the LLFP discharge burnup is ;19% smaller for 99Tc

and ;50% smaller for 129I, relative to the UO2 fuel case.

The safety-related reactivity coefficients are not affected

significantly by the LLFP loading, if the LLFP loading is

restricted by the fuel enrichment constraint ~5.0 wt%

235U!.

In the sodium-cooled ATW system, the 99Tc and 129I

transmutation rates are significantly enhanced by employing ZrH2 moderator. The annular target configuration filled with the moderator reduces the self-shielding

effects and increases the effective capture cross sections

relative to the distinct target and moderator pin configuration. There exists an optimal moderator volume

fraction to maximize the absolute transmutation rate.

However, the fractional LLFP transmutation rate increases monotonically with the moderator volume fraction ~i.e., as the target loading decreases!. For a fixed

LLFP loading, the moderator volume needs to be maximized to increase the transmutation rate. The local power

peaking in the neighboring fuel assemblies can be effectively mitigated by employing unmoderated 99Tc target

pins at the interface between fuel and target assemblies.

The heterogeneous loading of LLFP target material

in special moderated assemblies is more favorable than

the homogeneous mixing of target and fuel in terms of

the relative transmutation rate. Among various heterogeneous loading options, the loading in the reflector region

seems to be the best option from the neutron economy

and safety point of view. This option provides sufficient

transmutation rates to meet the LLFP support ratio goal

to balance TRU transmutation of the ATW system. By a

simultaneous loading of twelve 99Tc and twelve 129I target assemblies in the reflector region, 99Tc and 129I can

be consumed with a support ratio equal to the TRU support ratio ~;3.2! at the cost of increased fuel inventory

~and thus decreased discharge burnup! by ;10%. Discharge burnups of ;29 and ;37% are achieved for 99Tc

and 129I target assemblies with an ;5-yr irradiation period. The discharge burnup could be further increased by

increasing the number of target assemblies with the correspondingly reduced loading per assembly.

The current Yucca Mountain release evaluations do

not indicate a compelling need to transmute 99Tc and

129I because the dose rates are well below current regulatory limits for the spent-fuel LLFP inventory. However, elimination of the LLFP inventory could allow some

relaxation of the stringent waste form and container performance criteria, with associated economic benefits.

Therefore, some development of either specialized waste

forms or transmutation target for the LLFP is prudent,

especially when sustained fission power production ~with

accumulating FP inventory! is envisioned for future

applications.

ACKNOWLEDGMENT

The authors are grateful to E. E. Morris for providing the

isotopic dose rates of the Yucca mountain repository.

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