Nuclear Engineering and Design Manuscript Draft Manuscript Number: NED-D-12-00380 Title: Sustainable Gas Cooled Fast Breeder Reactor through Thorium Utilization. Article Type: Full Length Article Abstract: A conceptual design of a (232Th +233U) fuelled gas cooled fast breeder reactor (TGFBR) has been developed to demonstrate self-sustainability, breeding capability, actinide recycling capability and feasibility. Simultaneous use of 232Th and LWR spent fuel in the core has been considered. Results confirm a neutron spectrum dominating in a fast range (peak at 100 KeV) with the design achieving a breeding ratio (BR) of 1.034 and burn up of 74.5GWd/MTH. Eventual shortage of 235U and nuclear waste management are the key issues faced by the Nuclear power community. TGFBR addresses these issues while providing clean energy. A mixture of Thorium and Uranium (232Th +233U) is used as fuel and "spent fuel" is used as blanket for breeding Plutonium (239Pu). 233U would be obtained from existing thorium based fast breeder reactors. Reactor calculations have been performed with MCNP/MCNPX. The reactor has to be fuelled once every 5 years with a thermal power output of 445 MW. Detailed analysis of control rod worth has been performed and different reactivity parameters have been evaluated as part of the safety analysis. The design demonstrates the sustainability of thorium, feasibility of 233U as an alternate to 235U and an alternate use for spent fuel as a blanket for production of Plutonium.

Breeding ratio of more than 1.03 with self-sustainability of 5 years and power of 445MWTh.

Usage of spent fuel for actinide recycling and production of plutonium.

Operation with fast neutron spectrum.

Usage of thorium for production of 233U.

Generation-IV Gas cooled fast breeder reactor with Helium as coolant.

Highlights (for review)

1. Introduction The thorium fueled gas-cooled fast breeder reactor (TGFBR) features a high temperature

helium cooled fast breeder reactor. It is designed to operate in a closed fuel cycle mode. Fast

neutron spectrum enables minimization of waste through actinide burning and also good for

better neutron economy. 235U+238U is the fuel currently used in most of the reactors and due to

its massive consumption the natural uranium resources would be exhausted very fast. However,

better utilization of uranium is possible in a fast breeder reactor. Also, extensive research is

being done and steps are being taken to obtain an alternative to 235U. This proposed TGFBR

uses 233U as fissile material. 233U can be obtained through neutron transmutation of 232Th in

existing reactors. Spent nuclear fuel containing U, Pu, Np, Am, and Cm from pressurized water

reactors (PWRs) is recycled as blanket material in TGFBR, which facilitates the transmutation of

238U to 239Pu and helps in burning long-lived actinides also called as actinide recycling. It has

been observed that there usage of thorium instead of uranium could improve the proliferation

resistance due to the fact that there is less production of weapons material plutonium. This has

been discussed in [1]. Helium is used as a coolant because it is chemically and neutronically

inert and being in single phase eliminates the issues with boiling. However due to lower thermal

inertia (gaseous coolant), there are possibilities of rapid heating up of the core when there is

depressurization and loss of forced cooling. A primary loop design based on upward core

cooling and moderate pressure drop across the fuel channel and a fuel element based on

refractory materials such as nitrides and carbides ensuring high thermal conductivity could

eliminate this problem. Operating at higher temperature helps to achieve higher thermodynamic

efficiency. This paper presents a detailed feasibility analysis of TGFBR and gives a comparative

analysis of different materials used.

*Manuscript

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2. Feasibility Analysis To obtain both controlled nuclear fission chain reaction and breeding, it is inevitable to have the

number of neutrons produced per fission to be greater than 2, neutron capture to fission ratio to

be low, low neutron leakage from the core and less neutron absorption in structural materials.

This is represented by a symbol η i.e. number of neutrons produced per neutron absorbed. It is

given by, where is the number of neutrons produced per fission and α is the neutron

capture to fission ratio. Figure 1 shows that 233U has got the best η as compared to 235U in 10-

100 KeV energy range; this is where TGFBR spectrum lies.

<Figure 1 >

Usage of thorium carbide ceramic helps in obtaining a harder neutron spectrum. In ThC one

carbon atom is attached to one thorium atom unlike ThO2. A harder neutron spectrum implies

larger mean free path of neutrons, hence more leakage out of central core and this helps in

neutron absorption in blankets.

Silicon Carbide (Si ) is used as cladding material due to high thermal heat conductivity

coefficient and corrosion resistance at high temperatures. However being a ceramic, Si is

brittle and there are chances of it getting broken at high temperature. Therefore Si is usually

used as composite in combination with fiber and graphite [9].

Spent fuel from PWRs is used in the blanket regions of the TGFBR. Some amount of

238U present in spent fuel will be converted to 239Pu. 239Pu produced in blanket, after

reprocessing can be used to fuel other PWRs. Operating in harder spectrum also helps in

fissioning of minor actinides in spent fuel; hence it helps solving the problem of management of

spent fuel to some extent.

Helium has the least moderating effect as compared to steam and saturated carbon

dioxide. This makes it most suitable to obtain a harder neutron spectrum. Thorium is present in

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higher concentrations by weight than uranium in their respective ores and with the elimination of

enrichment with using thorium as compared to uranium, it is economically more preferable.

However, thorium, in its decay chain produces 208Ti and 212Po, which are known to emit gamma

radiation of energies more than 2 MeV, leading to special storage requirements. But given the

relative economic benefits with regard to reprocessing, fabrication, mining etc.., thorium would

be more preferable as compared to uranium in terms of energy production costs [8]. Lead, used

as a reflector also contributes significantly in retaining fission gas release to atmosphere. It also

acts as a good gamma shield.

2.1 Single fuel pin cell calculation

To obtain the optimum configuration in terms of safety, efficiency, breeding capability and

economy, configuration of single fuel pin needs to be optimized. In a single fuel pin, the property

that is optimized is the pitch-to-diameter ratio (p/D). The p/D ratio is optimized based on the

following driving factors, infinite multiplication factor (k∞) , percentage of fission occurred by

neutrons in fast range (npct_f) , primary loop pressure drop (PPL) and average energy causing

fission (Eavg_f). Higher npct_f ensures higher fission hence criticality and better breeding. PPL

ensures efficient energy conversion and power generation. The above mentioned parameters

have been calculated with different p/D ratio by changing the diameter and keeping the pitch

constant. Fissile content is kept at 10% 233U in 232Th. <Figure 2 shows a single fuel pin used in

optimization studies.

<Figure 2 >

< Figure 3 >

From Error! Reference source not found.< Figure 3, it is evident that for higher efficiency and

economy p/D ratio should be closer to 1. It was obtained to be 1.092. Pitch (p) was chosen at

0.87 cm, radius of fuel (rf) at 0.308cm, thickness of gap (tg), 0.007 cm, thickness of cladding (tc)

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at 0.083 cm and primary loop pressure drop was calculated as 1.0856 bar. Theoretically this

sounds logical because lower p/D ratio means lower moderation hence neutrons maintain in fast

spectrum.

2.2 Lattice design

It is necessary to arrange fertile and fissile fuels for optimized breeding capability, minimized

fuel (fissile) inventory and maximized power density. Usually three types of lattice configuration

exists, external breeding (EB), internal breeding (IB) and scattered breeding (SB). In EB, fertile

material is contained in the blanket region surrounding the core. In IB, fertile material is

contained in the center of the core and fissile material surrounds it. Usually IB also has external

breeding. In SB, fertile material is mixed with fissile material in the core itself. It has been

observed that SB has the best breeding capability followed by IB and then EB. But, SB needs

the highest fissile inventory, softens the energy spectrum (due to moderation) and is associated

with core refueling difficulties. EB results in a hard spectrum but has rapid reactivity loss which

requires frequent refueling and also results in low burnup. IB with EB solves these issues and

ensures BR greater than 1. EB and IB are usually called homogeneous configuration because

fertile and fissile materials are localized while SB is referred as heterogeneous configuration.

Therefore EB needs higher fissile enrichment and thick blanket. To minimize fissile inventory

and ensure hard spectrum a tight fuel lattice is desirable. Therefore a higher fuel volume fraction

is necessary. A triangular/hexagonal lattice arrangement has a higher fuel volume fraction as

compared to square lattice. Therefore fast reactors usually have triangular or hexagonal lattice

structure while thermal reactors opt for square lattice structures (loosely packed square array

helps in moderation). It has also been observed that the triangular array ensures a higher power

density as compared to the square array. The power density in square array is given by ,

while in triangular array it is , where q’ is linear heat generation rate and P is array pitch. It is

clearly evident that power density in triangular array exceeds that of square array by 15% for the

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same linear heat generation rate. Keeping in view the above mentioned discussions, a

hexagonal lattice with internal as well as external breeding configuration has been chosen.

Following figure gives a glimpse of the lattice configuration,

< Figure 4 >

The following table summarizes the specific configuration of the lattice elements,

Parameter Fuel blanket

Lattice pitch (cm) 13 13

Pitch of Single pin (cm) 0.87 1.56

Number of pins in one assembly 271 61

Number of assemblies 138 199

Table 1 : Lattice element specifications.

To increase breeding capability axial blankets are also used. Since the neutrons in a

hard spectrum have a higher mean free path there are possibilities of leakage, hence having an

axial blanket allows us to use some of the leaked neutrons in breeding. Following figure shows

the exact position of inner blanket, outer blanket and axial blankets.

< Figure 5 >

For a safe operation in terms of preventing meltdown of fuel rod, peaking factors play an

important role. Peaking factor is the ratio between maximum local energy depositions to the

average energy deposition in reactor core. It has been assumed that external blanket and axial

blanket are not part of reactor core while determining radial peaking factors. Axial peaking

factors are also calculated to obtain the hottest region of the core. This helps in determining the

maximum temperature at the hottest region in the core. Following figures display radial and axial

energy deposition in the reactor core,

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Figure 6 : Energy deposition in fuel and inner blanket region for radial peaking factor determination.

Figure 7 : Energy deposition along axis of fuel element for axial peaking factor determination.

Radial peaking factor and axial peaking factors obtained are 1.76 and 1.697 respectively.

2.3 Reactivity analysis

In this design two types of control rods are used. One is used to regulate the neutron flux to

control fission and obtain stable power. (Regulating Control Rods (RC)) and the other is used to

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shut down the reactor (Shut down control rods (SC)). Based on the name and functionality, it

can easily be inferred that the control rod worth of SC should be higher than RC. There are 6

control rods of each of the above types. The following figure describes the position of the control

rods,

< Figure 8 >

Control rod type Reactivity (ρ) (pcm) (pcm)

Shutdown rod 11901.2 72.1

Regulating rod 8420.8 70.1

Table 2 : Reactivity change due to control rod insertion.

Borated graphite is used as neutron poison/absorber. SC has a higher concentration of

B4C as compared to RC. When RC is inserted it replaces fuel while SC is inserted in a gap.

Following figure gives an idea about SC and RC,

< Figure 9 >

<Figure 10 >

To understand the effect of insertion and withdrawal of control rods it is important to

calculate the integral and differential rod worth. This is done by calculating the change in

reactivity at different control rod positions. Since the SC rods are for shutdown and are only

used as either fully inserted or fully withdrawn, it is logical that integral and differential rod worth

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is calculated only for RC rods.

Figure 11 : Differential control rod worth.

It is clearly evident that the control rod worth is less near the ends and high at the center

because near the center neutron flux is high. Ideally the curve should be symmetrical at the

center but it has some variation at the top half of vertical axis due to presence of control rods.

Even though control rods are removed they are still closer to the upper half of the vertical axis.

Due to being high absorbers, the control rods insert some perturbations in reactivity. The

integral rod worth is given by,

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Figure 12: Integral control rod worth.

Following plot describes how neutron flux is affected due to insertion of control rods,

Figure 13 : Effect of control rods on neutron flux (Dip in neutron flux due to insertion of control rods).

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Figure 14 : Lateral view of power deposition along XY plane (Control rods removed).

Figure 15 : Lateral view of power deposition along XY plane (Control rods inserted).

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The plots are built using a java based tool named “SDP Java Application “. This tool interprets

data from MCNP output file and develops an Excel spread sheet file with calculated data. The

plots show how neutron flux and power deposition is affected due to insertion of control rods.

2.4 Burnup Calculations

The reactor core is burned for 5 years using MCNPX and it has stayed critical for the whole term

and has given us the desired burn up of 74.49 . The breeding ratio obtained is, 1.0373.

Following plot describes the criticality of the reactor during the burn duration,

Figure 16 : K-Effective vs Burn duration

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Figure 17 : Power fraction versus burn days.

It is evident from the above figure that contribution of blanket to reactor power increases with

burn days. This is due to the fact that 238U transmutes to 239Pu due to neutron capture. Similarly,

contribution of fuel to reactor power decreases with burn days due to fissioning of pre-loaded 233U.

Figure 18 : Axial burning of 233U in different time steps, mass of 233U remaining in the fuel

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Figure 19 : Axial Burnup of fuel in different time steps.

The above figures describe that there is higher burnup of fuel towards the center of fuel

rod axially. This is due to the fact that neutron flux is higher towards the center. This is also

agreed by the control rods differential worth plots. Due to this behavior there is a higher fraction

of fuel remaining un-utilized towards the axial end of the fuel rods as compared to the central

region. To minimize wastage fuel enrichment variation axially can be opted. However, this would

affect neutron flux distribution and overall peaking factors. This aspect is a prospective area of

research.

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Figure 20 describes the closed cycle followed in this reactor design. 233U is obtained from fast

breeder reactors, where thorium is used as blanket; it is clearly described in [3] . It is assumed

that FBR has the required amount of fuel (1830 kg of 239Pu). FBR is operated for 3 years and

then we get the desired amount of 233U for this TGFBR. This flow is described by the bold black

line. After operating this TGFBR for 5 years 1723kg of 233U and 562.3 kg of 239Pu are obtained

following reprocessing. The subsequent cycle flow is represented by dashed line. This amount

of 233U is again mixed with thorium and used as fuel and the required balance i.e. 285 kg can be

obtained from FBR by supplying it 236.85 kg of 239Pu. These 285 kg of 233U can be obtained by

burning the FBR for 0.4 years. Remaining amount of 239Pu can be used to fuel LWR with would

FBR: obtained from Pu/Th – Carbide fuel in fast

breeder reactor (Burned for 3 years).

Spent fuel obtained from

existing LWR.

TGFBR – operated for 5 years with

100% power and 100% availability.

Reprocessing

233U – 1723kgs

239Pu – 562.3kgs

LWR

Figure 20 : Fuel cycle.

Reprocessing

Separation of

zirconium based

materials from spent

fuel.

236.85 kgs 239Pu

239Pu

325.45 kgs 239Pu

2008 kgs 233U

285 kgs 233U

First cycle

Subsequent cycles

1830 kgs of 239Pu

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give us our desired amount of spent fuel for blanket. If extrapolated to 4 burn cycles and

assuming that the structural materials can withstand irradiation and burnup limits for 20 years,

we would need the FBR to run for 4.2 years and we would be supplying 1301.788 kg of 239Pu to

LWR, along with 236.85 kg of 239Pu to FBR and 1723 kg of 233U for future reactor operation. The

only external input would be 1830kgs of 239Pu needed to obtain the initial amount of 233U from

FBR. Similar type fuel cycle is being employed by Bhabha Atomic Research Centre, Mumbai,

India by the design and development of AHWR [6] , where a combination of (Th-Pu)O2 and (Th-233U) O2 as fuel and initial loading of 233U is obtained from existing FBRs. Spent fuel obtained

from LWR cannot be used directly, even though they don’t need reprocessing but separation of

zirconium should be done due to its lower burnup and irradiation limit.

2.4.1 Breeding ratio

Breeding ratio (BR) is obtained by calculating the amount (mass) of fissile material at beginning

of time step and then at the end of the time step. Dividing the later by the former gives the

breeding ration. Following table highlights the masses of fissile materials.

Material Mass (kgs) – Beginning Mass (kgs) – End (5 years)

92235 (Blanket) 191 160.8

94239 (Blanket) 148.8 562.3

94241 (Blanket) 44.36 32.16

92233 (Fuel) 2008 1723

92235 (Fuel) 0 3.179

Total 2392.18 2481.43 Table 3: Breeding Ratio Data

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2.5 Neutron flux Spectrum

Figure 21 : Neutron energy spectrum

Neutron flux spectrum is obtained by using F4 tally in MCNP5. Normalized flux obtained is

.The plot shows that there is no thermal peak. There are very few

neutrons in thermal and epithermal region, which implies that there is less moderation. This

remains in line with the motivation of operation in a fast spectrum. There is a peak at higher

energy range near 0.1 MeV which is desired to obtain higher fission yield.

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3. Safety assessment Reactivity coefficients are calculated to ensure safe operation. Negative coefficient of reactivity

is desired. Two types of reactivity coefficients calculated are temperature coefficient of reactivity

( ) and void coefficient of reactivity . determines how the reactivity of reactor would

change if the temperature of fuel increases during the course of operation. Ideally if the

temperature is increased the multiplication factor should decrease due to Doppler broadening

and swelling of fuel. Hence for stable operation a negative value for is desired. This

TGFBR has a of -6.38E-06 . determines reactivity behavior during changes in

density of coolant i.e. depressurization incidents. Ideally a decrease in density should decrease

the reactivity of reactor to ensure safe operation. TGFBR has an of -0.339647557 .

Delayed neutron fraction, β is obtained as 0.00313±0.0007. Prompt neutron lifetime, lp is

1.8167E-06±1.35E-08 seconds.

4. Actinide recycling Presence of minor actinides (Americium (Am), Curium (Cm), and Neptunium (Np)) in spent fuel

causes long term maintenance issues due to high gamma activity. A fast neutron spectrum

enables recycling of these actinides because of higher probability of transmutation at higher

neutron energy due to favoring cross-section. Spent fuel used in blanket had a (MA: Total

mass of minor actinides, Pu: Mass of plutonium vector) ratio of 9.3% and after burning this

TGFBR for 5 years this ratio decreased to 5.47%. Hence 41.1% efficiency in actinide recycling

is obtained.

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5. Summary of Core Characteristic Parameters Following table summarizes the final parameters and their values of TGFBR,

Parameter Value Units

Reactor Power 445 MWth

Burn up 74.49

Thermal Efficiency 41.43%

Maximum Fuel Centerline Temperature Limit 2125 Current maximum temperature achieved 1689.571 Coolant temperature at reactor inlet. 200 Coolant temperature at reactor outlet. 1000.2 Pumping power 148.814 W

Primary loop pressure drop 1.0856 bars

Coolant flow rate 107.084 kg/s

Mass of initial content of fissile material 2008 Kg

BR 1.0373

Core life time 5 Years

Radial power peaking factor 1.76

Axial power peaking factor 1.69

Height of core 2.3 M

Outer radius of core 2.6 M

Table 4: Design parameters and their values.

6. Software tools and simulation code The model has been developed using MCNP and burn details have been obtained using

MCNPX. MCNPX is a general purpose Monte Carlo radiation transport code designed to track

particles over a wide range of energies [10]. It uses predictor-corrector approach to obtain time

dependent simulation results. For efficient retrieval and representation of data from MCNP

output file, an in-house java application has been developed. This application has the capability

to export burn details (F6 Tally), mesh values and flux details (F4) into excel sheet. It also has

capabilities to build input files for single fuel pin cell analysis.

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7. Conclusions Single fuel pin cell calculation was done by optimizing fission yield only, but if this optimization is

coupled with optimization of breeding gain then BR achieved can be higher. This TGFBR can be

used for 20 years till the irradiation limit of structural materials is reached. TGFBR qualifies as a

Generation – IV goals in terms of sustainability, economics, safety, reliability and enabling

proliferation resistance. From Figure 20, it is evident that this reactor can have self-sustainability

with the use of an existing FBR that can supply 233U. To run this TGFBR for 20 years we just

need to run the above stated FBR for 4.2 years and in the meantime, this GCFR would also

supply 1301.79 kgs of 239Pu to LWR .At the end it would still generate 1723 kgs of 233U and

562.3 kgs of 239Pu for future reactor operation. With such self-sustainability, usage of spent fuel,

usage of thorium and actinide recycling efficiency of greater than 40%, this is a prospective

candidate for future energy needs.

References [1] Sunil S. Chirayath et al. “Neutronic and nonproliferation characteristics of (PuO2-UO2) and (PuO2-ThO2) as fast reactor fuels” in

Nuclear Engineering and Design, Vol 239, issue 10. [2] S. Peterson et al., “PROPERTIES OF THORIUM, ITS ALLOYS, AND ITS COMPOUNDS”, ornl-tm-1144. [3] R Ramanna, S M Lee “The thorium cycle for fast breeder reactors” in J. Phys., Vol.27, Nos 1 & 2, July & August 1986, pp. 129-

137 [4] P.Anzieu et al., “Gas-cooled Fast Reactor (GFR): OVERVIEW AND PERSPECTIVES”, GIF Symposium – Paris (France) – 9 -

10 September; 2009. [5] Alan E. Waltar, Albert B. Reynolds “Fast Breeder Reactors” PERGAMON PRESS. [6] R.K. Sinha, A. Kakodkar “Design and development of the AHWR – the Indian thorium fuelled innovative nuclear reactor” in

Nuclear Engineering and Design Vol 236 (2006) 683-700 [7] Turan Unak “WHAT IS THE POTENTIAL USE OF THORIUM IN THE FUTURE ENERGY PRODUCTION TECHNOLOGY”

Progress in Nuclear Energy Vol. 37, No 1-4 , 137-144, 2000 [8] P. Rodriguez, C. V. Sundaram “NUCLEAR AND MATERIALS ASPECTS OF THE THORIUM FUEL CYCLE” in Journal of

Nuclear Materials 100 (1981) 227 – 249. [9] Sang Pill Lee, Jin Kyung Lee, In Soo Son, Dong Su Bae, Akira Kohyama “Thermal shock properties of 2D-SiCf/SiC

composites” in Fusion Engineering and Design (In Press). [10] MCNPX USER’S MANUAL, V 2.6.0, April 2008, Denise B. Pelowitz, editor.

[1] Sunil S. Chirayath et al. “Neutronic and nonproliferation characteristics of (PuO2-UO2) and (PuO2-ThO2) as fast reactor fuels” in Nuclear Engineering and Design, Vol 239, issue 10.

[2] S. Peterson et al., “PROPERTIES OF THORIUM, ITS ALLOYS, AND ITS COMPOUNDS”, ornl-tm-1144. [3] R Ramanna, S M Lee “The thorium cycle for fast breeder reactors” in J. Phys., Vol.27, Nos 1 & 2, July & August 1986, pp. 129-

137 [4] P.Anzieu et al., “Gas-cooled Fast Reactor (GFR): OVERVIEW AND PERSPECTIVES”, GIF Symposium – Paris (France) – 9 -

10 September; 2009. [5] Alan E. Waltar, Albert B. Reynolds “Fast Breeder Reactors” PERGAMON PRESS. [6] R.K. Sinha, A. Kakodkar “Design and development of the AHWR – the Indian thorium fuelled innovative nuclear reactor” in

Nuclear Engineering and Design Vol 236 (2006) 683-700 [7] Turan Unak “WHAT IS THE POTENTIAL USE OF THORIUM IN THE FUTURE ENERGY PRODUCTION TECHNOLOGY”

Progress in Nuclear Energy Vol. 37, No 1-4 , 137-144, 2000 [8] P. Rodriguez, C. V. Sundaram “NUCLEAR AND MATERIALS ASPECTS OF THE THORIUM FUEL CYCLE” in Journal of

Nuclear Materials 100 (1981) 227 – 249. [9] Sang Pill Lee, Jin Kyung Lee, In Soo Son, Dong Su Bae, Akira Kohyama “Thermal shock properties of 2D-SiCf/SiC

composites” in Fusion Engineering and Design (In Press). [10] MCNPX USER’S MANUAL, V 2.6.0, April 2008, Denise B. Pelowitz, editor.

*BibliographyClick here to view linked References

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