Use of Thorium in the nuclear energy technology
- experiences in Germany
K. Kugeler, N. Pöppe, S. Jühe, O. Schitthelm
Institute of reactor safety and –technology
RWTH Aachen University
September 2007
Summary
The future importance of Thorium for the development of nuclear technology is high, because the amount of Thorium available worldwide is even higher than that of Uranium.
The conversion of Thorium to Uranium 233 allows to produce a fissile material, which has the best neutron economy in thermal reactors.
The absorption resonances of Thorium 232 are essential for the negative temperature- coefficient of reactors and form a necessary basis for the reactor safety.
In Thorium fuel cycles the forming of Plutonium is avoided and the production rate of minor actinides is reduced. This fact can be very important for the direct final storage of spent fuel elements.
In Germany there was an extensive programme to introduce Thorium into the nuclear energy over a time span of more than 30 years.
During this programme both High Temperature Reactors AVR (50 MWth) and THTR (750 MWth) have been built and operated.
Parallel to these activities many reactor concepts, based on the principle of the pebble bed HTR, have been worked out. These were dedicated to the steam cycle, the gas turbine cycle, to process steam production and to high temperature process heat applications.
Much work was dedicated to the question of high conversion and breeding with Thorium fuel cycles, too.
The AVR, as the first pebble bed reactor in the world, was operated very successfully for more than 21 years, in the last 10 years with an average helium outlet temperature of 950 °C.
The reactor showed a high availability and the concept of continuous loading and unloading of spherical fuel elements was successfully proven. Many fundamental safety experiments have been carried out, - action of a strong negative temperature coefficient and the self-acting decay heat removal after total loss of active cooling – and showed the high safety potential of this type of reactors. The mass test of many different fuel element types, carried out in this reactor, showed that coated particle fuels, especially with Thorium, can be operated in helium cooled reactors very successfully.
The THTR showed that a large pebble bed reactor with Thorium containing fuel elements can be licensed, built and operated in an adequate way. Despite of a very long construction time, during which the licensing requirements were changed drastically in Germany in the course of LWR-development, this reactor was taken into operation. After some difficulties in the beginning, which were caused by the prototype character of this plant, it worked corresponding to all specifications and all design data were fulfilled. After the catastrophic accident in Chernobyl the acceptance of nuclear energy in Germany dropped very much and this reactor was taken out of operation as the first one, mainly from financial considerations, which were caused by the prototype specific questions of this project and because of political reasons. The plant, however, delivered substantial knowledge on the fuel, on operation and on reactor components of HTR.
Today in many countries of the world the concept of small modular HTR is followed.
In these reactors with thermal power till 250 MW in case of cylindrical core and till 400 MW in case of annular core a maximum of inherent safety features will be realised. By using the concept of self-acting decay heat removal in all cases of loss of active cooling accidents, the maximal temperature of fuel will be limited to values below 1600 °C. In this case TRISO- coated particles will practical totally retain the fission products inside the fuel elements.
Catastrophic release of radioactivity from the core will not be possible.
An extensive fuel qualification programme has been carried out during the whole German HTR-development. These programmes covered coated particles with BISO- and TRISO- coatings, it started with Thorium carbide, changed then to (Th/U)O2, and finally UO2-LEU- Particles and fuel elements were qualified. Very high burn-up values (> 150 000 MWd/t), high fast neutron doses (> 6·5·1021 n/cm²), high operation temperatures of fuel (T > 1250°C) and very low fission product release (< 10-5 in normal operation) were characteristic results of this work. Therefore the helium circuits were very clean in the operation of these reactors. For the case of severe accidents (total loss of coolant and total loss of active decay heat removal) many heating-up experiments of irradiated fuel elements have been carried out. These showed that till a maximal temperature of 1600°C, which is foreseen for small modular HTR, the release share would stay below 10-5 of the inventory even during a heating time of some 100 hours. This behaviour was shown for all interesting fission products like I 131, Cs 137, Sr 90, Kr 85, Xenon isotopes and further. This fact allows to prove and to demonstrate the safety behaviour of inherently safe modular reactors. Further broad tests programmes on corrosion
(air, steam, H2, CO2) and on strength values and other relevant parameters qualified the spherical fuel element for industrial and commercial application.
Many fuel cycles have been studied during the whole HTR-development. Starting with HEU- cycles (93% enriched Uranium and Thorium), later MEU-cycles (with U/Th-mixtures) and then LEU-cycles (around 8 to 10% enrichment) have been analysed in all details. The later change to LEU was necessary after the results of the INFCE-study had been published.
INFCE declared all materials with higher than 20% enrichment as weapon grade material. All fuel cycles can be used in the pebble bed HTR and the main requirements of technical layout (power per ball, fuel temperatures, burn-up, fast neutron dose, fission product retention) can be fulfilled. In the same sense the safety requirements, even for very extreme conditions, can be fulfilled for all fuel cycles, too (always sufficiently strong negative temperature coefficients, integrity of fuel till high accident temperatures and effective retention of fission products at high temperatures).
High-converter systems and even breeder systems would be possible, if the burn-up of fuel would be reduced.
In Germany there was a broad study on use of Thorium for high converting reactors (HTR, D2O reactor, molten salt reactor), which showed that these systems could be realised in future, if breeding becomes important. Just for the molten salt system to that time difficulties with materials were seen.
Experiences with Thorium containing fuel in AVR were excellent. The stationary coolant activity was very low (< 1 Ci/MWth) and therefore the personal doses and the release to the environment was very low, too. 14 different types of spherical fuel elements have been inserted into this reactor and operated overlapping, there was just one very small batch of fuel elements (with too large heavy metal loading), which caused higher release of radioactivity.
All other, especially Thorium containing fuel elements, behaved very well and fulfilled all requirements. In totally AVR carried out a mass test of more than 300 000 fuel elements.
The experience with the fuel elements of THTR was very similar to that of AVR. The release rates of their HEU-elements ((Th/U)O2, 93% enriched) were very small (< 1 Ci/MWth), too. The same experience was made for personal doses and release in normal operation.
The fabrication of Thorium containing coated particles and spherical fuel elements today is a well established and in an industrial scale available process. Fluidised bed technology for coating and isostatical pressing processes for the fabrication of the spherical fuel elements are qualified and applicable. All specifications for the fuel elements can be fulfilled.
Especially the content of free Uranium in the matrix (this contamination causes main parts of release of fission products) was reduced by the development to values of around 10-5 of the total content of Uranium. The matrix material (synthetic graphite) has been optimised, the shrinkage by irradiation with fast neutrons was maximal 1% at full burn-up.
The intermediate storage of spent HTR fuel elements today is technically full established.
Concepts using cans and small containers have been tested and the temperatures, the tightness and the radiation doses have been measured in detail. The best solution for the future will be the dry storage in thick-walled cast iron or steel vessels, which are cooled from the outside just by radiation and natural convection.
The vessels will be arranged in concrete buildings and don’t need any active safety installations. The maximal fuel temperature will be below 200°C and maximal vessel temperature below 100°C. An optimal time span for the intermediate storage will be 50 to 100 years. Then the decay heat is very low and this offers more advantage for the final storage in geological depositories.
For the final storage of spent HTR-fuel-elements containing Thorium broad development programmes have been carried out. Until now it is planned to fill in spent pebbles into canisters and to deposit these in salt. Therefore many measurements for the behaviour of these fuel elements in salt brine have been done. The result is that the leaching rates for the most important isotopes (Cs 137, Sr 90, Pu 239) are very small especially in case of TRISO coated particles. Vessels can be designed, which stay longer even in salt brine, than the content of fission products is remarkable in the fuel elements (because Cs 137 and Sr 90 decay practical totally in around 500 years).
Therefore the danger for the public caused by the final storage of pebble bed fuel should be very small, as all studies on release from final storage indicate.
Especially for the HEU-, MEU- and Nearbreeder- cycles reprocessing of Thorium fuels was necessary. Therefore in Germany over 2 decades an intensive programme for reprocessing took place. The THOREX-process was the basis of this activity. A technical facility for the
total process (JUPITER) was built and many different facilities have been built and operated for the head end, i.e. the destruction of the fuel matrix and the coatings. As a result of this programme it was stated, that reprocessing of Thorium fuels could be done, but the status of development stood behind that of the PUREX-process, which is used for the Uranium cycle today. After the INFCE-Study highly enriched fissile material was practically forbidden for power reactors and therefore the HEU-cycle was not possible anymore. The programme for reprocessing in Germany therefore was stopped at the end of the 80ties.
A major concern to avoid misuse of fissile material is the destruction of Plutonium by burning in thermal reactors.
Coated particle fuel together with Thorium as fertile material is very well suited to gain high conversion rates of the fissile content of Plutonium. Broad studies have been carried out as well for LWR-Plutonium as for weapon grade Plutonium. Conversion of more than 90% of the fissile material can be realised in suited HTR using Thorium containing coated particle fuel. This process can be carried out fulfilling all safety requirements of the core design, which are necessary.
In Germany there were broad analyses for innovative fuel elements with Thorium for different reactors, too. There are possibilities to insert Thorium in LWR with a significant reduction of the production of new Plutonium.
Especially in the case of D2O-cooling the use of pebble bed fuel elements with Thorium could be advantageous to realise a system without core melting in severe accidents. The principle of self-acting decay heat removal by radiation and natural convection could be used in this concept, too and would allow to limit the fuel temperatures in extreme accidents to values of
<1600°C as explained before for the small modular HTR.
To avoid proliferation and misuse of fissile material in the world is a major concern of nuclear safety and international politics. The Uranium 233 formed in Th-containing reactors requires similar attention as U 235 and Pu 239. Using the MEU-cycle with Thorium there is a chance to avoid material with higher enrichment than 20% in the whole fuel cycle. The conditions of proliferation are extremely bad in this case and there are much easier and cheaper methods to get the necessary amount of fissile material to build a nuclear weapon.
Overall the use of Thorium in strategies for future energy supply plays an important role.
In ADS-processes Thorium always is a necessary material for conversion into fissile material and to be able to supply the electricity demand of ADS-processes by operation of fission reactors.
Because of the large amount of Thorium available in the world for future use and because of many technical advantageous aspects this material might get a large importance for the nuclear technology.
Content
1 General aspects of the use of Thorium in nuclear reactors 2 Basic physical data of Thorium 232 and Uranium 233
3 Overview on German reactors containing Thorium fuel elements 3.1 AVR
3.2 THTR
3.3 New concepts of modular HTR
4 Results of German development programmes on fuel elements including Thorium as fertile material
5 Information regarding fuel cycles using Thorium 5.1 Overview
5.2 HEU-cycles 5.3 MEU-cycles
5.4 Near breeders and the U3O8-demand of different fuel cycles 5.5 Breeder systems (HTR)
5.6 Breeder studies for different reactors
6 Production of Thorium containing fuel elements
7 Experience with Thorium containing AVR-fuel elements 8 Experience with Thorium containing THTR-fuel elements 9 Innovative fuel elements with Thorium
9.1 HTR-fuel elements for future modular HTR 9.2 Thorium containing fuel elements for LWR-plants
9.3 Thorium containing pebble bed fuel elements for D2O-reactors 10 Intermediate storage of Thorium containing fuel elements 11 Direct final storage of Thorium containing fuel elements
12 Reprocessing of Thorium fuel elements in the THOREX-process and conditioning processes
13 Plutonium burners using Thorium
14 Non-proliferation aspects of Thorium fuel-cycles
15 Thorium as breeding material for ADS-processes 16 Importance of Thorium in energy strategies 17 Literature
1 General aspects of use of Thorium in nuclear reactors
There are some important aspects, why Thorium is interesting for long-term application in the field of nuclear energy [1.1 till 1.4]:
• There are higher resources of Thorium than of Uranium in the world. Geologists estimate a factor of 3 more Thorium than Uranium.
• Thorium 232 is converted in nuclear reactors into the fissile isotope Uranium 233. In thermal reactors U 233 has the best η-values regarding the neutron economy.
• Using Thorium as a fertile material the amount of Plutonium produced in the reactor is drastically reduced compared to the Uranium-cycle. The amount of other Transuranium isotopes is reduced, too. These changes in the composition of spent fuel are responsible for the reduction of the radiotoxicity in case of direct final storage of spent fuel elements.
• Several reactor concepts using Thorium have been developed in the past and successfully tested; further progress is possible, especially in Heavy-water-cooled systems, in HTR and in ADS-systems Thorium will be an important fuel. LWR-plants also can use this fuel cycle with some advantages.
Thorium is contained in the soil with an average content of 7 to 13 mg per kg material. It occurs in several minerals, the most common is Thorium-phosphate, which contains till 12%
ThO2 (monazite, as example in India).
Countries like Australia, India, Norway, USA, Canada, South Africa together today estimate reserves of around 1 Mio t as known. It was not reasonable from the standpoint of economics to carry out more detailed work of prospection until now.
Thorium 232 decays very slowly and has a very low specific activity – the half-life is 1,4 . 1010a –, but there are other Th-isotopes in the chain of the Thorium and of the Uranium – decay series, which have shorter half-lifes and therefore larger specific activities. However, their share in the Thorium is very small and therefore they are nearly unimportant.
A characteristic example is Th 228 with a half-life of 1,9 years.
As Uranium the isotope Thorium can be used in the nuclear fuel cycle as fertile material. Th 232 absorbs neutrons in the epithermal region and U 233 is produced via the decay of Proactinium. Fig. 1.1 shows the relevant isotopes in the Th-cycle.
This Uranium isotope has the highest neutron production (η-value) in thermal reactors.
Principally therefore Thorium allows the realisation of high converter – or even breeder – reactors. Closed fuel cycles including a special reprocessing of Thorium-containing fuels (THOREX-process) are possible.
In former reactor projects Thorium was used as Thorium carbide, in the last decades Thorium oxide was more interesting. ThO2 has a very high melting point (3300°C) and therefore the use of this material promises large advantages in nuclear technology for further application as fertile material. Additionally the heat conductivity is as high as that of UO2, and the thermal stability is very good.
Figure 1.1: Relevant isotopes in the Thorium cycle
ThC and ThO2 have been used with great success in high temperature reactors operated until now in different countries:
• AVR and THTR in Germany
• Peach Bottom and Fort St. Vrain in USA
• Dragon-reactor in Great Britain as a common European project.
In heavy water reactors the use of Thorium partly is common practice, too:
• PHWR (Pressurised heavy water reactors) in India
• Fast breeder system with Th to breed U 233 in India
Light water reactors have been tested with Thorium containing fuel elements in the past
• insertion of some Th-elements in the reactor Shippingport in USA
• insertion of Th-containing fuel elements in Lingen and Obrigheim in Germany
Additionally many reactor projects on the basis of the pebble bed concept and on the block- type fuel elements, too, have been studied in Germany with large effort and much work for the relevant fuel elements, mainly containing Thorium as fertile material:
• PR 500 for process steam generation
• Helium turbine systems (HHT 600, HHT 1000)
• PNP 500 for process heat applications
• HTGR 1160 (steam cycle, block-type fuel elements)
• HTR 500 (steam cycle)
For the future application there is much work preparing the use of Thorium in nuclear reactors in different types of reactors:
• use of Thorium to convert weapon grade Plutonium in a HTR-gasturbine-project (common activity of USA and Russia)
• potential future application of Thorium in the pebble bed HTR (South Africa, China)
• use of the potential for a application of MOX-fuel in LWR-plants in operating commercial reactors in different countries (PuO2/ThO2-mixtures)
• use of Thorium in molten salt reactor-concepts
• insertion of Thorium containing fuel elements in AHWR (Advanced Heavy Water Reactors)
• CANDU-reactors in Canada are planned to use Plutonium and Thorium
• in ADS-concepts (Accelerator Driven Systems) Thorium can be applied in a subcritical blanket. New fuel U 233 can be bred and actinides can be destroyed.
Beside all these applications of the past and of the possibilities of the future there is the need in the long term to establish a fuel cycle for Thorium parallel to that of Uranium. The main aspects to be discussed in this field are:
• production of Th- containing fuel elements
• intermediate storage of spent fuel elements, which contain Thorium
• final storage of spent Th- fuel elements
• reprocessing of Th-fuels
There are some further aspects, which have to be discussed for all fuel cycles. These are:
• proliferation aspects of fuel cycles containing Thorium and U 233 or other fissile isotopes
• general strategies of developments in nuclear energy economy, if large amounts of Thorium will be used in future
The following chapters discuss the topics mentioned before in more detail.
2 Basic physical data of Thorium 232 and Uranium 233
For the following discussions and analysis some physical aspects of Th 232 and U 233 should be remembered [2.1 till 2.5].
Thorium 232 shows many absorption resonances in the epithermal energy region (see figure 2.1). These resonances cause the strong negative temperature coefficient of all reactors which contain this material and are designed in an appropriate manner. Furthermore the absorption in Th 232 causes the breeding of U 233, which has a high fission cross section in the thermal energy region and which has a high η-value in a thermal neutron spectrum. This high η- value above 2 principally allows the realisation of high converter – or even breeder reactors.
Figure 2.1: Absorption cross section of Thorium 232 in the epithermal energy region
Based on these resonances the resonance integral for an homogeneous infinitely diluted medium, defined by the expression
o
( )
eff th
E a E
I E dE σ E
∞ =
∫
⋅can be defined. The parameters of the main resonances and the dependence of this integral on the moderation ratio in a homogeneous mixture can be seen from fig. 2.2
The limiting value at infinite dilution is around 70 barn. Characteristic designs of HTR with moderation ratios of 400 arrive at values of 40 barn.
a)
b)
Figure 2.2: Aspects of resonance absorption in Th 232 a) Resonance parameters
b) Resonance integral of the homogeneous infinite diluted medium dependent on the moderation (resonance integral of U 238 for comparison)
The resonance integral of Th 232 generally is smaller than that of U 238, which has important consequences on the different fuel cycles as will be explained later. The influence of the heterogeneity of fuel elements normally can be described by empirical equations. As example for ThO2 a suited equation is
2 3,6 37,6 /
IThO = − + ⋅ S M
where S/M is the surface to mass ratio of lumped fuel and IThO2 is expressed in barns. S/M has the dimension cm2/g. The equation is valid for values 0,2 ≤ S/M ≤ 1,5.
Especially for the cases of large values of the ratio S/M detailed and complex models to calculate the resonance absorption are necessary to get a good approximation for this important effect in neutron physics and reactor design. Fig. 2.3 shows the results of different model-calculations.
Figure 2.3: Effective resonance integral of Thorium 232 - lumps dependent on the ratio S/M
Thorium can be fissioned at high neutron energy too, however with a very small cross section above a threshold energy of 1,2 MeV (see fig. 2.4).
Figure 2.4: Cross section of fission at high energies of Th 232 and some Uranium isotopes
In the thermal energy region the absorption cross section of Th 232 is given by figure 2.5. The scattering cross section in this part of the spectrum has a value of around 12 barn.
Figure 2.5: (η, γ) cross section of Th 232 in the thermal region
The product of Thorium conversion, U 233 has very advantageous properties as a fissile material especially in the thermal energy region. As figure 2.6 indicates in a thermal spectrum U 233 has the highest η-values, where η is defined by the expression
( ) ( ) ( )
( ) ( )
= ⋅
+
f
f n
E E E
E γ E η ν σ
σ σ
Averaged over a thermal spectrum the differences are explained on behalf of table 2.1. These values show that the absorption per fission has the lowest value for U 233 and that the number of fission neutrons per absorption has the highest value in the thermal spectrum.
Figure 2.6: Dependence of the η-values of U 233, U 235, Pu 239 on energy
Energy region (eV)
U 233 U 235 Pu 239 Pu 241 absorption per
fission 0 … 107 1,12 1,24 1,59 1,41
fission neutrons
per absorption 0 … 107 2,24 1,98 1,81 2,10
thermal absorption cross section
<1,85 250 230 1100 900
Table 2.1: Parameters of fissile materials in a thermal neutron spectrum (HTR)
These parameters play a major role during the discussion of high converters and breeders based on Thorium in the next chapters.
The fission cross section of U 233 is comparable to that of U 235 in the thermal region, however there are resonances above an energy of around 1eV (see fig. 2.7): this has to be considered, if shifts in the spectrum during massive transients in a thermal reactor are regarded.
Figure 2.7: Cross sections for fission of the isotopes U 233, U 235, Pu 239 in the thermal energy region
U 233 shows further fission and absorption resonances in the epithermal energy region (see fig. 2.8) and this could influence the dynamical behaviour of a reactor loaded with large amounts of Th 232 and U 233, too.
At high energies U 233 shows a small cross section of fission as other fissile materials, too (see fig 2.9).
Figure 2.8: Dependence of the total and fission cross section of U 233 on energy at thermal and epithermal energies
Figure 2.9: Cross sections of fission of the isotopes U 233, U 235, Pu 239 at high energies
3 Overview on German activities on Thorium containing reactors
3.1 AVR (Atom Versuchs Reactor)
In the AVR spherical fuel elements have been used for more than 20 years of operation and especially Thorium containing coated particle fuel has been tested extensively and with very good success [3.1 till 3.5]. Different forms of spherical fuel elements were inserted (see fig.
3.1) and different types of coated particles have been tested.
Figure 3.1: Forms of spherical fuel elements inserted in AVR:
a) Hollow spheres with a shell containing coated particles
b) Spheres with a ring gap and a closure, the ring contains coated particles c) Pressed synthetic graphite matrix, coated particles are contained in an inner zone
As example (U, Th)C2 with BISO-coating, UO2 and ThO2 as separated feed and breed particles, and finally UO2-low-enriched particles (TRISO) (after introduction of the consequences of INFCE-study in USA to exclude proliferation of weapon grade material) have been used. After this decision of INFCE high enriched material (e > 20%) was practically forbidden. Figure 3.2 shows some coated particle concepts, which have been tested successfully in the AVR in mass tests.
Figure 3.2: Types of coated particles, which have been tested in AVR a) BISO-particles (C/C, HTI), diameter approx. 1mm b) TRISO-particles (C/SiC/C, LTI), diameter approx. 1mm
Characteristic data of Thorium containing fuel elements operated in AVR are included in table 3.1. Results of the operation and of many accompanying experiments for testing fuel elements are given in chapter 7.
parameter dimension value remark
diameter mm 60
coated particle - BISO
fuel - UO2/ThO2
fissile material g/fuel element 1 93% enriched U 235
fertile material g/fuel element 5 Thorium
average burn-up MWd/t 100 000
maximal fuel temp. °C 1250
max. surface temp. °C 1050
fast neutron dose
(E > 0, 1MeV) n/cm2 6 . 1021
residence time in core
full power days 1000
Table 3.1: Some important data of Thorium containing fuel elements in AVR
The first pebble bed reactor AVR was built corresponding to the figure 3.3. Because at the time of planning of this reactor coated particles were not available, it was thought, that the contamination of the cooling gas helium would be very high. Therefore the reactor has a double-walled vessel and all penetrations through the vessel are realised using the principle of two closures, too. The reactor used continuous loading and unloading of fuel elements and therefore excess reactivity for the compensation of fuel burn-up was not necessary. This is very important for considerations on very severe reactivity accidents. The fuel elements were recycled several times before they reached their final burn-up.
The steam generator was arranged above the core and two helium circulators were positioned at the bottom of the vessel. The flow of helium through the core was from the bottom to the top upwards. Table 3.2 contains some important data of the plant.
parameter dimension value
thermal power MW 46
electrical power (net) MW 15
average core power density MW/m3 2.6
number of fuel elements - 100 000
helium inlet – and outlet –
temperature °C 275/950
helium pressure bar 10.8
steam conditions (temp./press.) °C/bar 500/72
burn-up MWd/t till 140 000
fuel cycle - MEDUL
(several passages) Table 3.2: Some data of AVR-plant
The AVR has been operated with high availability during 20 years and has demonstrated that the pebble bed reactor concept works and that such reactors can be operated relatively easy.
Many useful and new safety experiments have been carried out and especially the experiments to demonstrate the concept of self-acting decay heat removal are important for all new modular HTR worldwide. Overall in the AVR around 200 kg of Thorium were inserted, and the operation showed, that there can be attractive features to use this material in the future in nuclear technology.
Figure 3.3: Overview on the vertical section of the AVR-primary system
3.2 THTR (Thorium High Temperature Reactor)
The THTR (Thorium High Temperature Reactor) used spherical fuel elements loaded with U 235 (93% enriched) as fissile material and Th 232 as fertile material, both in form of oxides.
The heavy metal was used in form of coated particles (BISO-particles) [3.6 till 3.11].
Fig. 3.4 shows a section through the fuel element and gives an impression of the coated particle. The small BISO-particles are embedded in the graphite matrix of the fuel elements.
The BISO-particle has a small kernel made from UO2/ThO2 and is surrounded first by a porous graphite layer and then by 2 dense pyrolytic graphite layers. Additionally there is an overcoating around the particle. Around 20000 particles are contained in one spherical fuel element, which has an outer diameter of 6 cm.
The enrichment of the Uranium was 93% and around 1 g fissile (U 235) and 10 g fertile material (Th 232) were contained in one ball.
Figure 3.4: Spherical fuel elements of THTR
a) Section through a fuel element (6 cm diameter)
b) View into a BISO-coated particle (total diameter: around 1 mm)
Table 3.3 contains some important data of THTR fuel elements and coated particles.
Experiences with these fuel elements have been gained in the AVR, in the THTR itself and in many test rigs in different reactors. These results on the mass tests of Thorium containing fuel elements are summarised later in the chapters 7 and 8.
The THTR has been designed and the fuel fabrication had been started, before as a consequence of the INFCE-study all fuel cycles with higher enrichment than 20% were practically forbidden in the most countries of the world. Therefore the reactor was started with this HEU-fuel (High Enriched Uranium).
a) b)
parameter dimension value
diameter mm 60
type of particles - BISO
number of coated particles
per fuel element - 20 000
diameter of coated particle mm 1
form of heavy metal - oxides
U 235 – content g/fuel element 1,11
enrichment % 93
Th 232 – content g/fuel element 10,3
burn-up MWd/t 100 000
average power kW/fuel element 1,1
maximum power kW/fuel element 3,9
max. fuel temperature °C 1250
max. surface temp. °C 1050
fast neutron dose
(E > 0,1 MeV) 2
n cm
6 . 1021
Table 3.3: Data of THTR fuel elements
Fig. 3.5 gives an overview on the THTR itself, it shows a vertical and horizontal section through the primary system, mainly formed by the large pre-stressed concrete reactor pressure vessel. It contains not only the core and its surrounding structures (reflector and thermal shield), but 6 steam generators for the production of hot steam and 6 helium circulators integrated into the cylindrical wall of the pre-stressed reactor vessel. The fuel elements were loaded continuously during operation and unloaded continuously, too. They were recycled several times to gain the final burn-up. Table 3.4 contains some important data of the plant.
parameter dimension value
thermal power of reactor MW 750
electrical power of plant (net) MW 300
efficiency (net)
(dry air cooling tower) % 40
helium inlet – and outlet
temp. reactor °C 250/750
helium pressure bar 40
core power – density
(average) MW/m3 6
fuel cycle - MEDUL
(several passages through core) steam condition (life steam) °C/bar 530/180
steam condition (reheat) °C/bar 530/45
cooling system - dry air cooling
Table 3.4: Some important data of THTR
Figure 3.5: Primary system of THTR 300: vertical section
(1: core, 2: side reflector, 3: top reflector, 4: bottom reflector, 5: steam generator, 6: hot gas duct, 7: fuel element discharge tube, 8: reflector rods, 9:
helium circulator, 10: core rods, 11: pre-stressed concrete reactor pressure vessel, 12: liner with cooling and insulation, 13: fuel element unloading system, 14: fuel element loading system)
The reactor was taken in operation in 1986. After some difficulties, caused by the principle of a prototype of a totally new reactor concept, it went to successful operation. As example in the first phase of operation the rods of the second shut-down system often were inserted into the pebble bed and caused damages on the surfaces of the fuel element. This procedure was totally unnecessary from the standpoint of safety. In new modular HTR-concepts all operations of the first and second shut-down systems are done just from the side reflector.
All specified data have been reached, especially the power, temperatures and the total efficiencies.
After the decision of a large German political party to go totally out of use of nuclear energy after the accident in Chernobyl, the operation of THTR was finished. This party formed the government in the state North Rhine Westphalia, were the THTR was built. The decision was made because of political and financial reasons, caused by the prototypic structure of the project.
3.3 New concepts of modular HTR
Nuclear reactors of the future have to be designed with highest safety standards, best fuel optimisation and for many applications in the energy market. In the generation IV – programme the modular HTR (VHTR = Very High Temperature Reactor) is one of the candidates to fulfil these requirements.
The HTR–module has been developed as a modular heat source for cogeneration and for process heat applications [3.12 till 3.18]. In case of combination with cogeneration processes the helium outlet temperature from the core should be 700°C and the steam condition 530°C/180 bar. For process heat applications like combination with the steam reforming process or with intermediate heat exchangers the helium outlet temperature should be 900 to 950°C. The fuel elements could contain mixtures of Uranium, Plutonium and Thorium – coated particles in different compositions. The basic design scheme of such a modular unit, which is combined with a steam generator, is shown in figure 3.6.
Figure 3.6: HTR–Module (200 MWth),
a) Vertical section through the primary system b) Horizontal section through the primary system
(1: core, 2: core internals, 3: reactor pressure vessel, 4: steam generator, 5: helium circulator, 6: coaxial duct, 7: outer surface cooler, 8: drive of control rods, 9: concrete structures of inner cell)
a) b)
The reactor core as well as the steam generator are arranged in separate steel vessels which are connected by another connecting vessel, which contains a coaxial hot gas duct. The helium circulator is arranged at the top of the steam generator vessel and forms a part of it.
The reactor core consists of 360 000 fuel elements, which contain the fuel in form of TRISO–
Coated particles. The core has a diameter of 3 m and a height of 9.43 m. This choice of H/D deviates from all reactor designs known until now, which use a H/D–ratio of nearly 1, because of the optimisation of the neutron balance.
The reason is that the decay heat in a loss of coolant accident shall be transported out of the core just by heat conduction, heat radiation and natural convection with limitation of the maximal fuel temperature to values less than 1600°C. This concept of inherent safe decay heat removal is a basic condition for fission product retention. A broad experimental programme has shown that till these temperatures the fission products stay practical totally inside the coated particles.
Furthermore it was required to operate all shut–down elements inside the side reflector. This limits the core diameter, too, to the indicated value of 3 m. The average power density in the core is chosen by these boundary conditions as 3 MW/m3.
The arrangement of the steam generator prevents its overheating and its damage due to a fai- lure of the circulator. To realise higher power ratings, multiple HTR 200 modules each are arranged in separate primary concrete cells. These concrete cells containing the pressure vessels are equipped with surface cooler systems to remove the decay heat in case of an unavailable steam generator loop. These surface cooler systems are permanently in operation.
All modular systems together with their concrete shields are situated inside a common reactor building protecting the nuclear system against outer impacts and designed without an inner liner. In order to assess the safety performance as well as the economics of the reactor, it is important to consider that the service loops and the supply systems are not relevant to safety.
All systems and components outside the primary cells can be designed applying conventional technology. This is substantial precondition to achieve justifiable investment costs also in case of nuclear plants with a small power capacity.
In case of total loss of coolant and active decay heat removal the heat is transported just by self-acting processes to the outer surface cooler or the concrete. The figure 3.8 shows the characteristic time dependence of the temperature in the hottest part of the reactor.
Figure 3.7: Reactor building of the HTR-Module a) Vertical section,
b) Horizontal section (2 modules)
(1: reactor pressure vessel, 2: steam generator vessel, 3: inner concrete cell, 4: outer surface cooler, 5: reactor building)
Figure 3.8: Time dependence of the hottest part of the core in case of total loss of coolant and active decay heat removal
4 Results of German development programmes on fuel elements including Thorium as fertile material
In totally, HOBEG at Hanau has manufactured a million fuel spheres with (TH, U)O2 particles for AVR and THTR, and 50000 spheres with UO2 particles for AVR and irradiation tests in Material Test Reactors (MTRs) [4.1 till 4.7]. Some characteristic data of particles and of fuel elements are summarised in the tables 4.1 and 4.2. With this experience, the fuel manufacturing and performance data pertinent to fission product source terms have been established (see table 4.3).
UO2 kernel Diameter
506 µm
Density 10.80 Mg/m³
Coating layers Thickness Density
Buffer 96 µm 1.01 Mg/m³
Inner PyC 42 µm 1.91 Mg/m³
Silicon Carbide 35 µm 3.19 Mg/m³
Outer PyC 39 µm 1.91 Mg/m³
Table 4.1: Characteristic data of coated particles (TRISO)
Property Parallel
to equator Perpendicular to equator
Density 1.75 Mg m-3
Coefficient of heat
conduction at 1000°C 41 37 W m-1 K-1
Corrosion rate * at
1000°C 0.79 mg cm-2 h-1
Crushing strength * 26.3 23.7 kN
Abrasion rate * 1.85 mg h-1
Fraction defective SiC
layers 1.4 x 10-5
Table 4.2: Characteristic data of spherical fuel elements
The manufacturing specifications can be reproduced in production scale equipment: therefore they are being used in ongoing reactor design activities worldwide (China, Japan, South Africa, Korea, USA, EU-countries).
Irradiation testing of HTR fuels utilised the AVR at Jülich and various material test reactors.
Important performance related parameters are irradiation temperature, heavy metal burn-up, and the accumulated fluence of fast neutrons. Irradiation testing to date has been performed in the range 800 to 1250°C, 8 to 14% FIMA, 1 to 8·1025 neutrons/m² (E > 0,1 MeV). During
irradiation the continuous monitoring of released fission gases indicates the integrity of particles.
Figure 4.1 shows the release rate of Kr 85m from a test with spheres irradiated for 400 full power days in the Jülich DIDO reactor. Similar results have been obtained from the
measurements of Kr 88, Kr 87, Xe 133 and Xe 135. The results show that no sphere contained failed particles (neither manufactured defects, nor irradiation induced failures). The only fission product source term derives from traces of Uranium in the fuel element graphite at a level below 1 ppm equivalent to natural trace contamination.
Expected Design Limit
Fuel element manufacture
• Fraction particles with defective SiC
• Heavy metal
contamination per fuel element
3 ⋅ 10-5
10 µg
< 6 ⋅ 10-5
< 50 µg Normal operations
• End-of-life particle failure fraction
2 ⋅ 10-5 < 2 ⋅ 10-4 Design basis accidents
• Particle failure fraction
5 ⋅ 10-5 < 5 ⋅ 10-4
Table 4.3: Data of fuel manufacturing and performance
Figure 4.1: Release of Krypton 85 during irradiation of spherical fuel elements
As part of the post-irradiation examinations, KFA has constructed high temperature annealing furnaces for accident simulation tests with spherical fuel elements. One furnace is characterised by a tantalum heating element and a water cooled trap for the accurate assessment of released metallic fission products (figure 4.2). This furnace is being used for isothermal heating tests at 1500 to 1800°C. Another furnace with a graphite heater is used for temperature ramps to 2500°C. Both furnaces are located in hot cells and are connected to a helium sweep circuit to enable continuous gas release measurements outside the cell.
Figure 4.2: Heating furnace used in accident simulation tests with irradiated spherical fuel elements
Irradiated fuel elements have been heated at constant temperatures of 1800, 1700, 1600 or 1500°C for times between 25 and 1000 hours. Fuel particles were (Th, U)O2 TRISO and UO2
TRISO irradiated in the AVR and different material test reactors. At 1500 and 1600°C the SiC layer stays completely intact (figure 4.3).
The initial Cs 137 release fraction of AVR elements at the 2 ⋅ 10-5 level is caused by Caesium contamination deposited on the sphere surface from old releasing fuels in the AVR. With spherical fuel elements irradiated in the MTR, Cs 137 fractional release is 1 ⋅ 10-6 and initial Kr 85 release is 1 ⋅ 10-7. Both figures are consistent with the assumption that there are neither defects (from manufacture) nor failed (form irradiation) coatings, and a heavy metal contamination of 1 ⋅ 10-6 dispersed throughout.
For the temperature transients of new modular HTR designs, where the peak fuel temperature is less than 1600°C, the above results indicate that no particle failure would be observed. For
the TRISO particles from several spheres, this translates into a statistical statement of less than 5 ⋅10-5 failure fraction.
Figure 4.3: Fission product release from spherical fuel elements irradiated in HFR Petten and heated at constant 1600/1800°C
Figure 4.4: Average Krypton 85 release at 1600°C from different types of fuels
Figure 4.3 shows the measured fractional release of key radio-nuclides as a function of heating after corrections for contamination. These data were obtained on UO2 TRISO fuel elements heated at 1600 and 1800°C after MTR irradiation to values of burn-up, fluence, and temperature well beyond the requirements of the ongoing HTR designs. Fig. 4.4 showed the influence of the type of coatings and the height of the burn-up. In fig. 4.5 there are some further information on the influence of temperature and time on the release.
Figure 4.5: Release rates of different fission products from spherical fuel elements (TRISO-Coated Particles)
a) Release of 85Kr dependent on temperature
b) Release of 137Cs, 90Sr, 85Kr dependent on time at a heating temperature of 1600°C
In an accident involving depressurisation and total loss of active cooling, the temperature would rise to 1600°C in 20-100 hours and then decrease to 1000°C in 1000 hours, while in the simulation tests the fuel is maintained continuously at 1600°C. Therefore the integral release of radioactivity from the spherical fuel elements of modular HTR (cylindrical cores till 250 MWth, annular cores till 400 MWth) is restricted to values of less than 10-5 of the total inventory in core.
5 Information regarding fuel cycles using Thorium
5.1 Overview on cycles
The HTR can be operated with different fuel cycles. The fuel consists of the nuclides U 233, U 235, Pu 239, Pu 241 as fissile material and Th 232, U 234, U 238 as fertile material. In the pebble bed reactor these materials can be used in mixed or separated form. Dependent on the today economic conditions and regarding the situation and perspectives of fuel supply for the next decades, the ”open” cycles will have advantages for a time schedule of some decades.
With rising costs of Uranium ore the closed cycles will become interesting in far future. Table 5.1 contains some information on different fuel cycles of HTR.
Characterisation Explanation Fuel loading
“open” cycles
Th/U (93%) (HEU) Thorium cycle with Uranium (93% enriched) as fuel
ThO2/UO2 (93%) Th/U (20%) (MEU) Medium enriched (∼ 20%)
Thorium/Uranium-cycle
ThO2/UO2 (20%) U (8%) (LEU) Low enriched (∼ 8%)
Uranium cycle
UO2 (8%) closed cycles
Th/U Thorium cycle ThO2, 93% enriched UO2
from reprocessed fuel
Th/denat. U Denatured cycle ThO2, 20% enriched UO2
U from reprocessing, enriched 15%
PB Prebreeder ThO2, 93% enriched UO2
in separate fuel element
NB Netbreeder ThO2, UO2 of reprocessed
Fuel, mainly 233U Table 5.1: Fuel cycles of HTR (pebble bed system)
Cycles with different enrichment and different breeding materials have been developed and proved in mass tests in AVR and THTR. Cycles with highly enriched Uranium (93%) and Thorium, so called HEU–cycles, as well as cycles with 8% enriched Uranium with U 238 as fertile material (LEU) are fully developed.
Because of the INFCE–regulations cycles with more than 20% enrichment today are not allowed because of non-proliferation requirements. This excludes nowadays the use of HEU- cycles. For all new HTR–projects therefore today LEU–cycles are foreseen with around 8%
enrichment and a burn–up of 80 000 MWd/t to 100 000 MWd/t of heavy metal.
Especially for near breeders the HEU–cycle with lower burn–up is feasible, however this is an option for a far future, if the Uranium ore costs become much higher. The MEU–cycle with medium enrichment (< 20%) would be still proliferation resistant. The pebble bed HTR allows a high cycle flexibility. In the AVR during normal operation fuel elements corresponding to the different options mentioned above have been tested without any major changes of plant parameters in a continuous change. 10 different types of fuel elements have been tested. More details of fuel fabrication, storage, aspects of breeding and proliferation are explained later in chapter 12.
5.2 HEU-cycles
The High-enriched Uranium-Thorium cycle [5.1 till 5.4] was mostly used until now in HTR plants. Uranium with 93% enrichment was the initial fuel and Thorium was inserted as fertile material. The U233 produced from the Thorium should be recycled in subsequent loadings to reduce the U235 makeup requirement. Until now spherical fuel elements using this fuel have been used in Germany and block-type fuel elements in USA.
HEU/Th-fuel cycles promised a good neutron economy and rescource utilisation, however the potential nuclear weapon proliferation causes a problem and after the results of the INFCE- study this fuel cycle in the today situation is practically forbidden. The fuel cycle allows high conversion ratios, because the graphite moderator of HTR has very low parasitic absorptions.
U 233 has nuclear properties close to that of U 235, therefore burn-up changes are minimised and fuel zoning and power shaping are simplified. Fig. 5.1 contains operations and principal of material flow in case of the U 233/Th-cycle of HTR.
Figure 5.1: Operations and principal of material flow of the U 233/Th-cycle of HTR
The fuel of the pebble bed-HTR is contained in coated particles, which have a UO2/ThO2- mixed oxide-kernel. In USA in the block-type reactors UO2 is used as fissile and ThO2 as fertile material.
The HEU/Th fuel cycle is complicated by the fact that U 232 is also generated. The chain leading to U 232 is: Th 232 (n,2n) Th 231 (β-) Pa 231 (n, γ) Pa 232 (β-) U 232. U 232 has a relatively short half-life, it decays into Th 228 and in a series of short lived intermediates to stable Pb 208. The chain is shown in fig. 5.2. The most significant of the intermediates are Rn 220. This is a gas, which is transported through filters and Tl 208 and Bi 212, both emitting energetic γ -rays. The γ -activity associated with these isotopes must be accommodated in the refabrication plant, since it is not possible to separate the U 233 chemically from U 232. On the other hand these activities form a self-protection, which makes the misuse of the material more complicated.
Figure 5.2: Decay products of Th 232, U 232
The neutron physical evaluation of the HEU-cycle is based on the conversion chain of Thorium as shown in fig. 5.3. The isotope Pa 233 plays a major role, because it decays to U 233 and because of relatively important parasitic absorption.
Figure 5.3: Chain of conversion of Th 232
Based on an HEU/Th-concept similar to the THTR-fuel elements (moderation ratio: NC/NHM
= 325, heavy metal loading: 11,2 g/ball, burn-up: 100 000 MWd/t, enrichment: ≈ 10%) one gets characteristic data of the equilibrium core corresponding to table 5.2:
Parameter Dimension Value
average enrichment
(at start) % 7,16
burn-up MWd/t 100 000
conversion rate - 0,59
fissile material
inventory Kg/GWel 917
U3O8-demand
(0,25 tail) Kg/GWdel 462
loading U 235 Kg/GWdel 1,81
discharging U 233 + U 235 +
Pa 233 Kg/GWdel 0,51
discharging Np 239 + Pu 239
+ Pu 241 Kg/GWdel 0,002
Pu (fiss)/Pu (tot) in
discharged mat. % 19
In situ use of Pu 239 + Pu 241
% 98
Fast dose 1021 n/cm2 5,1
Table 5.2: Some characteristic data of an HEU-cycle in equilibrium
These data show that a production of Plutonium is almost entirely avoided and that the power of balls and the fast neutron dose stay within allowed limits.
5.3 MEU-cycles
In a Thorium cycle with denatured Uranium the optimal proliferation resistance is realised (fig. 5.4). 20% enriched Uranium and Thorium are used in one fuel element type as mixed oxide [5.5 till 5.7]. The loading can be chosen with 9,7 g of heavy metal/fuel element.
Figure 5.4: The fuel cycle with Thorium and denatured Uranium
The inserted Uranium 235 is converted by fission corresponding to the Uranium 233, which is bred from Thorium. Fig. 5.5 shows the dependence of the isotopes depending on the passage through the HTR-core. The enrichment of the fissile material U 233 and U 235 is reduced from 20% to 14% during the passage of the fuel elements though the reactor.
Figure 5.5: Balance of fissile material during the passage through the reactor in the Thorium/denatured Uranium-cycle
After unloading the fuel elements would be reprocessed. The Thorium would be separated and the Uranium isotopes would be inserted in a second fuel element type. During their passage through the reactor their enrichment is reduced to 2,7%. Therefore a further reprocessing is not interesting from the economical standpoint. Furthermore Uranium 236 is built up till an enrichment von 3,7%. U 236 causes difficulties from the radiological standpoint. These elements would be brought into a final storage.
In no position of the fuel cycle Uranium is available in a form, which would be suited for misuse and would be weapon-grade. The content of fissionable Plutonium is very small. At the discharge the value is 0,024 g/ball and this Plutonium is denatured to 52% with Pu 240 and Pu 242, which are not fissionable.
To generate an amount of 10 kg fissionable Plutonium the total amount of fuel discharged yearly from a 1000 MWel-reactor would have to be reprocessed. Because of the mixing with Pu 240 and Pu 242 a complicated technology of production of bombs would be necessary.
Additionally the energy release would be small, around 10% compared to a weapon from pure Plutonium 239 (see table 5.3).
Figure 5.6: Plutonium content of spherical fuel elements dependent on burn-up
Term Characterisation Content of fission material [%]
Critical mass [kg fission
material]
remarks
Weapon Plutonium 239Pu + 241Pu 100 ~ 8 With reflector (Beryllium)
Weapon Uranium 235U 93 ~ 25 With reflector (Beryllium)
Weapon Uranium 233U 100 ~ 8 With reflector (Beryllium)
Reactor Plutonium 70% Pu-content ~ 33 ~ 60 Burn-up ~ 35000 MWd/t Reactor Uranium 238U + 235U ~ 10 ~ 1500 15t heavy metal Reactor Uranium Uranium + Plutonium
20% content ~ 13 ~ 250 Very low release of energy
Table 5.3: Characteristic parameters of weapon grade materials
5.4 Near breeders and the U
3O
8-demand of different fuel cycles
The question of fission material supply for nuclear reactor systems over long time periods is relevant since the beginning of the development of nuclear technology. As already indicated before in chapter 1, for the HTR different fuel cycles may be used. Especially Thorium, which is available as a resource in comparable extent like Uranium, is usable as a breeding material in HTR cycles [5.8 till 5.12]. The complex breeding chain for Thorium is presented in figure 5.6.
It is of great importance in this context that the isotope 233Pa not only may decay to the fission material 233U to be gained, but may also be converted to 234Pa with a high reaction cross section (110 barn) by a (n, γ)–reaction. This leads to a significant loss with regard to the recovery of new fission material. The height of the neutron flux and the fuel element design are additional parameters of great importance with regard to the utilisation of Thorium in a HTR. The neutron flux should be as small as possible to realise high breeding gains.
The 233U gained during breeding processes in thermal reactors is a very good fission material and due to its high η–value it is superior with regard to the fissionability in comparison to the other fission materials 235U and 239Pu in the thermal energy spectrum as can be seen from figure 5.7. The average η–values for a thermal spectrum can be defined as
0
0
0
0
( ) ( ) ( )
E
E
E E dE
E dE η φ η
φ
⋅ ⋅
=
⋅
∫
∫
with E0 as the upper boundary of the thermal spectrum.
Then the following average values may be derived (with E0 ~ 1 eV) for a thermal reactor:
235 233 239
( U) 1.98; ( U) 2.24; ( Pu) 1.81
η = η = η =
It can be stated that cycles with utilisation of 232Th and with the production of 233U offer the best possibility to reach high conversion or even breeding with a thermal spectrum.
Theoretically breeding with η >2.2 is possible. For Thorium cycles, the formalism for the characterisation of doubling time, breeding ratio and breeding gain for breeding systems may be used. A simplified term for the breeding rate C is
1 1 C η V 1
= − − ⋅ α + with the definitions:
η= average neutron gain by fission (thermal spectrum)
V = number of neutrons per fission, which are lost by other processes than absorption in fuel
a/ f
α σ σ= (averaged over the thermal spectrum)
Figure 5.7: Breeding chain of Thorium
Figure 5.8: η–values dependent on energy for the different fissile materials
The breeding gain, defined as the relation of new generated fission material to the total mass of fissioned material is
( 1) (1 ) G = C− ⋅ +α
The doubling time TD, which is a characteristic parameter for the assessment of the quality of a breeding system is defined by
(1 )
f D
f
T m
m ε
= ⋅ +
Δ
where mf is the fission material inventory of the reactor, Δmf is the mass of surplus fission material generated in the time period TD and ε is the relation of the outer inventory to the core inventory.
It is obvious that this relation should be as small as possible with regard to short doubling times. This means that the interim storage time for spent fuel should be short, which is disadvantageous for the reprocessing process, because in this case fuel with a very high activity has to be handled. The fissile material surplus Δmf is dependent on the parameters breeding gain G, utilisation time T, total efficiency ηtot as well as on the specific demand on fission material ζ:
f
tot
m G T ζ Δ = ⋅η ⋅
With this term, the doubling time may be derived from the formula:
(1 ) 1
(1 ) ( 2)
f tot
D
T m
T V
ε η
ζ α η
⋅ + ⋅
= ⋅
⋅ + ⋅ − −
In order to reach a low doubling time in the breeding process of the Thorium cycle, the following requests should be fulfilled: low inset of fission material, high utilisation time of the plant, small values of V and small ε.
Extensive system studies concerning open and closed Uranium– and Thorium cycles demonstrated the potential of high temperature reactors with regard to the achievement of high conversion factors or even the possibility of breeding. Some results are presented in the following.
The nuclear characteristics of fission nuclides and breeding nuclides of the Thorium cycle are particularly favourable in the thermal spectrum of graphite moderated high temperature reactors. Open Thorium cycles show a comparable demand for ore as an open Uranium cycle.
Figure 5.8 shows the results, related to an annual electrical work of 1 GWel for different cycles.
It can be seen that the demand for ore at open cycle is 30% lower than that for a light water reactor. This fact is also due to the higher thermodynamic efficiency of the energy conversion in the circuit process.
During the open Uranium cycle of the HTR, about 90% of the Plutonium bred during operation of the reactor is burned directly in–situ in the reactor. Therefore the Plutonium content of spent fuel elements is very low and additionally is denatured by higher isotopes.
For closed Thorium cycles conversion rates of 0.8 to 1 and with further reduction of the burn–
up breeding factors C > 1 may be realised. But the fission material inventory in the reactor has to be raised significantly for this purpose, leading to reduced burn–up rates of about 20 000 MWd/t heavy metal (fig. 5.9).
Figure 5.9: Annual demand on Uranium ore related to an electrical work of 1 GWel as a function of burn–up for different fuel cycles
The demand on ore may be reduced significantly by these strategies, compared to the former calculated demand e.g. for the THTR. With regard to the aspects mentioned above for such cycles, reprocessing of HTR fuel elements and a refabrication for irradiated fuel has to be planned. Higher fabrication costs for fuel elements have to be taken into account concerning the evaluation of such fuel cycle concepts.
Closed cycles, which can be operated with enriched Uranium ( 20% ) and Thorium and which show a very good proliferation resistance, may show high conversion rates and therefore a reduced Uranium ore demand. For the realisation of closed fuel cycles for HTR as well as for other reactor concepts economic reprocessing capacities have to be available.
Near-breeder
5.5 Breeder Systems
Thorium allows principally the realisation of breeding in helium cooled HTR, in D2O-cooled systems and in molten salt reactors. These three options have been studied extensively in Germany in the past [5.13 till 5.16]. The HTR fuelled with Thorium and with Uranium 233 may be discussed here as an example.
As indicated in table 5.4 Uranium 233 has the highest η value in a thermal spectrum. Low moderation ratios cause a harder neutron-spectrum and reduce this value.
Nc/Nm
= 246 Nc/Nm
= 180 Nc/Nm
= 110 Nc/Nm
= 80
U 233 2,24 2,24 2,22 2,21
U 235 1,98 1,97 1,93 1,87
Pu 239 2,1
Table 5.4: Average η-values of fissile isotopes in thermal reactors (0…107eV) dependent on the moderation ratio
This favours the Thorium cycle for the HTR, if high conversion factors or even breeding is the main goal of the reactor design.
Broad studies on thermal near breeders or breeders have shown the result given in fig. 5.10.
Parameters are the burn-up and the moderation ratio.
Figure 5.10: Conversion ratio of HTR versus heavy metal burn-up; moderation ratio as parameter (thermal power: 3000 MW, power density in core: 5 MW/m3)
The figure explains, that at a small burn-up value of 20000 MWd/t and at low moderation ratio (NC/NHM ≈ 80) a conversion ratio of around 1 would be possible. The decreasing values at higher burn-up are caused by the increasing parasitic neutron absorption in the fission products.
Furthermore the calculations and the results in fig. 5.10 show, that with decreasing heavy metal loading, which means increasing the moderation ratio, the conversion factor becomes smaller. The reason for this behaviour is as follows: The fissile inventory of the core increases nearly proportional to the heavy metal inventory. Therefore the core has a high absorption cross section and this causes a reduction of the leakage of neutrons.
Moreover a high density of fissile material in the core results in a relatively low neutron flux at a given power of the core and by this two effects are caused: There will be low neutron losses in Proactinium 233. More Pa 233 nuclei can decay to the fissile material U 233 and by parasitic absorption in Pa 233 is reduced. The second effect will be a reduced neutron absorption in Xenon 135, which is proportional to the power density and the neutron flux, too.
Further studies showed, that the use of a radial blanket, containing just Thorium loaded fuel elements, would cause a higher conversion ratio (around 2%), caused by the reduction of the radial leakage.
The results in fig. 5.10 have been found for a large cylindrical core. As is well known, such a large core would arrive at fuel temperatures of maximal 2500°C in case of extreme cooling accidents (total loss of coolant + total loss of active decay heat removal + total loss of liner cooling). This accident would cause relatively high fission product release from the fuel elements.
Today HTR-cores of smaller power (modular HTR: cylindrical core with a thermal power of 200 MW, annular core with a thermal power of 400 MW) are considered; they would have maximal fuel temperatures of less than 1600°C in the very severe accidents mentioned before.
In this case the release of fission products from the fuel elements would be restricted to values of less than 10-5 of the total inventory in the core.
This same safety standard could be realised for large thermal power if an annular core is applied and if pre-stressed reactor pressure vessels which allow much larger diameters than forged steel vessels are used to install the primary circuit.
This is a very important topic for future application of HTR: by this different design reactors can be realised with higher conversion factors and with the limitation of the fuel temperatures in severe accidents to values below maximal 1600°C.
During the analysis of breeding systems a “symbiotic HTR-system” consisting of two pebble bed HTR fuelled with Thorium has been considered.
One system, named pre-breeder, employs a moderation ratio, which is suitable for Uranium 235, which is used as feed fissile material. This pre-breeder produces U 233, which is used as the fissile material for the second system (see fig. 5.11)
In the second reactor a breeding design is realised in which a low moderation ratio and a low burn-up corresponding to the conditions shown in fig. 5.11 are used.
Naturally these systems require negative temperature coefficients and therefore some Thorium is mixed with the U 235 of the pre-breeder.