integral fuel cycle

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Chapter 2 – Fuels and fuel cycles for High Temperature pebble-bed Reactor

2.1 Introduction

Nowadays nuclear power is the only greenhouse-free source that can face the always increasing world-wide energy demand. The LWR technology (at present the most widespread nuclear reactor technology) is secure and well- proven: its major “real” drawback is probably the limited exploitation of uranium resources and the consequently “large” production of waste. In fact, annually a 1000 MWe PWR produces about 30 tons of Spent Nuclear Fuel (SNF, with a burn-up around 30 GWD/tU) with the following average composition:

• 94% U²³⁸

• 1% U²³⁵ (please remember that natural uranium contains 0.7% of U²³⁵)

• 1% Pu

• 0.1% MA (Np, Am and Cm)

• 3÷4% FP

Considering the actual reprocessing technology this quantity of SNF contains about 1200 kg of "waste" (i.e. FP+MA). Once vitrified, this waste occupies about 20 m³, which correspond to less than 0.0001% of the total amount of domestic waste in UK for instance[2-2]. On the other hand, this is a relatively small quantity of waste even if compared with the conventional energy sources: a 1000 MWe coal-fired power plant discharges annually 6·10⁶ tons of CO₂, 2·10⁵ tons of ashes, 2·10⁵ tons of SO₂[2-1], that means a volume roughly 1.6·10⁸ times higher than that of vitrified nuclear waste¹.

The main issue for the nuclear power opponents is that the spent fuel radiotoxicity takes more than 100000 years to make its radiotoxicity equal to that of the uranium ore from which it descends. This long time is mainly due to Pu, but also if all the Pu would be transmuted, the time to reach the LOM (Level Of Mine) would still be around some tens of thousand years because of the presence of the MAs.

However, if all the actinides were fissioned directly or indirectly (by conversion fertile-to-fissile), the "real" remaining waste would be constituted

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Today Pu from SNF is partially recovered: France, UK, Japan, Russia, China and India reprocess their spent fuel in order to fabricate new MOX fuel elements, while other countries, as USA and Sweden, adopt the OTTO cycle.

From a global point of view[2-4], 16% of the world energy demand is covered by nuclear power, that supplies 350 GWe, roughly equally subdivided under USA, Europe and the rest of the world. These plants produce 10500 tHM/year of spent fuel, of which 3900 tHM/year are reprocessed. As already mentioned, reprocessing plants recover U and Pu only, while MA are treated with FP as waste. Of the Pu recovered form LWRs only 25% is consumed in thermal reactors, while 10% is converted in MA. MOX fuel can be reprocessed not more than 2 times in order to manufacture new LWR MOX fuel (Pu isotopic vector becomes more and more poor in fissile nuclides while many higher Pu isotopes and MA are poisons in the thermal spectrum) and its radiotoxicity increases because of Cm²⁴⁴ build-up[2-4].

Thus, it was clear since the beginning of nuclear age that different kind of reactors (e.g. HTR to use Th resources and fast reactors to increase U exploitation, by fertilization) would be necessary in order to increase the availability of nuclear fuel. Researches on fast reactor technology started in the past in order to multiply by 50 the availability of nuclear fuel resources.

According to the goals of sustainability, economics and proliferation resistance of the Generation IV Initiative[2-5], all the HM has to be seen as resource. The realization of the so-called

full actinide recycle

or

integral fuel cycle

(i.e. a self-sustaining fuel cycle in which the feed is constituted by fertile material only) improves 180 times the uranium resources exploitation with respect to a

once through

LWR cycle[2-3]. Using a closed fuel cycle the uranium resources would be sufficient for around 2500 years at the actual consumption rate (considering only the "Identified Resources", that are only a fraction of the total worldwide uranium availability)[2-6].

The activity performed at the University of Pisa fits into this frame: for many years a DIMNP group is studying gas-cooled reactors, which have been largely recovered by Gen-IV initiative (two of the six concepts proposed are gas-cooled reactors). They seem to have the capability to reach the Gen-IV goals thanks to their unique characteristics. Particularly the ultra-high burn- up allowed by HTR fuel elements (up to 750 GWd/tHM) could lead to reduce the Pu quantity, and the total amount of HMs inside the core is strongly reduced. On the other hand, a consequence of a so long irradiation cycle is a net increase of MAs quantity. However recent neutronic studies demonstrated that is possible to exploit better the natural resources of nuclear fuel and to obtain a virtually integral use of nuclear fuel by means of optimized fuel cycles [1-9].

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2.2 Fuel General Description

The PBMR uses particles of heavy metal dioxide coated with silicon carbide and pyrolytic carbon. The particles are encased in graphite to form a fuel sphere or pebble of about the size of a billiard ball. Helium is used as coolant and energy transfer medium, to drive a closed cycle gas turbine-compressor and generator system. When fully loaded, the core would contain approximately 450 000 fuel spheres. Typically the packing (volume)fraction inside the core is of 61%.

Figure 2.1: Pebble-bed fuel[1-5]

The main feature of HTR cores is the fuel element: a micro particle with a less than 1 mm diameter, constituted of a central kernel surrounded by four layer (Figure 2.1 and Figure 2.2). Each layer has own specific function which is different from that of the others. Particularly, we have[1-7]:

1. The spherical central kernel made of HM Oxide, with a 100÷250 µm radius . Its density is around 10.5 g/cm³.

2. A Layer of porous graphite (buffer). It acts as a first barrier against

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4. The SiC layer, improving the mechanical strength of the micro particle.

Particularly, the SiC layer is able to resist to stresses due to FPs production.

5. Finally, an Outer Layer of Pyro-Carbon, with the same function as the internal one.

Figure 2.2: MCNP particle plot

A typical HTR fuel element is constituted of some thousands of coated particles embedded in a graphite matrix modelled in a spherical shape (Figures 2.1 and 2.3) .

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Figure 2.3: MCNP model of Pebble-bed fuel

2.3 PBMR-400 uranium pebble

The typical uranium loading is 9 g per fuel-sphere and the U²³⁵ enrichment is 9.6 wt%. The inner zone of the fuel sphere (with a 5 cm diameter) contains about 15000 UO₂ coated micro-spheres (TRISO) embedded in a graphite matrix. The pebble fuelled zone is surrounded by an outer 0.5 cm thick fuel free zone made of graphite as well. Each coated particle acts as a fission product barrier. All data are shown in Table 2.1.

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Table 2.1: Fuel Element Characteristics[1-2]

Description Value Fuel pebble

Fuel pebble outer radius 3.0 cm Thickness of fuel free zone 0.5 cm Total heavy metal loading per

fuel pebble

9.0 g

Matrix density 1.74 g/cm³

Packing fraction in pebble bed 61 % Coated Particle

Kernel Coating Material C / C / SiC / C

Layer thickness 95 / 40 / 35 / 40 µm

Layer densities 1.05 / 1.90 / 3.18 / 1.90 g/cm³

Fuel kernel

Fuel kernel diameter 250 micron Kernel material type UO2

U-oxide density 10.40 (95% TD)

U isotopes (%)

235 9.60

238 90.40

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2.4 PUMA project plutonium and plutonium-minor actinide fuels

The fresh fuel specifications [1-1] used in the PUMA Project are given in Table 2.2. A second reference composition, representative for a fuel containing also Am, is given in Table 2.2A.

Table 2.2: PUMA Fuel Element Characteristics[I-5]

fuel pebble

2.0 g

Fuel kernel

Fuel kernel diameter 200 micron Kernel material type PuO_1.7

Pu-oxide density 10.89 (95% TD) Pu isotopes First Generation (%)

238 2.59

239 53.85

240 23.66

241 13.13

242 6.78

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Table 2.2A: Second reference fuel composition, originating from PWR spent fuel after five years of cooling[1-2]

Nuclide Fraction, wt%

Np-237 6.8 Pu-238 2.9 Pu-239 49.5 Pu-240 23.0 Pu-241 8.8 Pu-242 4.9 Am-241 2.8 Am-242m 0.02 Am-243 1.4

2.5 Thorium pebble configuration

In this paragraph, we will present a pebble for HTR using Thorium as fertile element in order to suggest an other possible composition for a better neutron economy and to increase the efficiency of the pebble. We will use this kind of mixture because:

• When the Th²³² captures a neutron becoming Th²³³, after decay Thorium becomes U²³³. This element is very dangerous for enviroment and human health because it generates four strong gammas. For this reason it is difficult to manipulate and the beast solution is to use it directly in the fuel.

• Using thorium enables the possibility to exploit an element that is more abundant in the environment than uranium (9 ppm vs. 3 ppm respectively), hence enhancing greatly the availability of nuclear fuel.

• It is possible to use Th²³² as fertile element with 1^st generation Pu as driver fuel, thus reducing the total amount of Plutonium and not producing MAs with higher mass numbers. In this way, nuclear power can be produced without creating large amounts of transuranic elements.

We will use 1^st generation Pu (i.e. Plutonium produced during irradiation of UO₂ fuel in a typical LWR) as driver fuel for our Pu-Th cycle. This kind of Pu is not useful for proliferation purposes, due to its content of even isotopes (Pu²³⁸, Pu²⁴⁰, Pu²⁴²) characterized also by a high self-fission rate. The next table 2.3 shows the characteristics of the studied fuel element, including the isotopic vector of 1^st gen. Pu

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Table 2.3: Thorium-Plutonium Fuel Element Characteristics [1-7]

fuel pebble

1.2 g

Fuel kernel

Fuel kernel diameter 240 micron Kernel material type (Th-Pu)O2.0

Pu-Th-oxide density 10.50 (95% TD) Th isotopes (%)

232 66.68%

Pu in total mixture (%) 33.32%

Pu isotopes First Generation (%)

238 2.61

239 53.99

240 23.63

241 13.06

242 6.71

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2.6 Transmutation physics

2.6.1 Theory

The term “transmutation” is applied to all the nuclear reactions which transform a nuclide into an other nuclide. In nuclear reactors transmutation happens by way of (n,c), (n,f) and (n,2n) reactions typically. It is clear that the probability of transmutation for a nuclide is strictly linked to its effective cross-sections for (n,c), (n,f), (n,2n) reactions in the considered core. What’s more, the transmutation rate depends obviously also from the neutron flux intensity. Let us call to mind some useful parameters in transmutation studies[I-4]. The neutron spectrum is defined as:

) , (

) , , ) (

, ,

( r t

t E t r

E

r r

ϕ

χ = ϕ

_(2.1)

where φ is the neutron flux, while the effective cross-section of the nuclide

n

for the reaction

x

is expressed by:

∫

=

MeV eV

x n x

n r t E r E t dE

10 001 . 0

,

, (r, ) σ ( )χ(r, , )

σ _(2.2)

The upper integration limit depends from the fission spectrum (typical of a given nuclide) while the lower limit considers that at the reactor temperature the neutron energy is higher than 0.001 eV (in a LWR the mean energy of the thermal neutrons is about 0.04 eV for instance).

The transmutation rate for a nuclide

i

is determined by the Bateman’s Equations:

∑

^→

+ +

−

+

+ +

+

+ +

+

−

+

−

=

k

i k k

i n n i i

c i

i n

n i a i

i i i

t r N

t r t r N t

r t r N

t r t r N

t r N t

r dt N

d

λ

ϕ σ

σ

λ

) , (

) , ( ) , ( )

, ( ) , (

) , ( ) , ( ...) (

) , ( )

, (

1 ) 2 , ( , 1 1

, 1

) 2 , ( , ,

r

r r

(2.3)

Let us name σ_i_,_a +σ_i_,₍_n_,₂_n₎ +σ_i_,₍_n_,₃_n₎ +..._asσ_i,_t, effective transmutation cross- section of the nuclide

i

. The previous equations show that destruction rate is as high as:

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• The effective transmutation cross-section of the nuclide

i

is high

• The effective transmutation cross-sections of the precursory nuclides is small or, if it is high, its most important term is the fission cross-section

• The fluence is high (i.e. the intensity of the neutron flux is high or the irradiation time is long)

• The decay constant of the nuclide

i

is high while the ones of the precursors are small

The (n,xn) cross-sections of the actinides are not so important if compared with the capture or fission cross-sections, so we can neglect them in transmutation analysis. Similarly the decay constants are small if compared with the transmutation rates. Then we can use the following simplified equation:

) , ( ) , ( )

, ( ) , ( )

,

(r t _, N r t r t ₁_, N ₁ r t r t dt N

d

i c i i

a i

i r =−

σ

r

ϕ

r +

σ

₋ ₋ r

ϕ

r _(2.4) In order to reduce the long-term radiotoxicity of the SNF, the best transmutation way is fission (direct, if the nuclide is fissile, or indirect, by way of conversion from fertile to fissile): in fact, capture reactions (also followed by decay) cause the build-up of other very radiotoxic actinides, while fission produces elements with shorter (except few cases) half-lives.

Please note that the effective cross-sections are obtained by weighting over the neutron spectrum.

2.6.2 Innovative gas cooled thermal reactors

HTRs are helium-cooled, graphite-moderated high temperature reactors.

These peculiar characteristics have many positive consequences also from the neutronic point of view:

• The use of helium as coolant and of graphite as moderator and structural material entails reduced parasitic captures and then brings to a very good neutron economy

• The moderator separated from a neutronically inert coolant allows to change the lattice parameters without changing dramatically the cooling conditions. This implies a very large flexibility in the choice of

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implies higher fluence than in LWR, with a beneficial effect on the transmutation capability. What’s more, HTR has a slightly epithermal spectrum if compared with LWR (Figure 2.4).

The neutron density in the slowing down range is higher in HTR than in LWR, because the mean lethargic increase of the neutrons is higher with hydrogen than with carbon. Nevertheless the thermal peak in LWR is smaller because of neutron capture by hydrogen nuclei. Thus, HTR seems to have a better transmutation capability than LWR.

It is useful to show the

R.P.

for some relevant actinides, calculated over the spectrum of a pebble-bed HTR (Table 2.4).

Nevertheless, in order to obtain an efficient transmutation to fulfil the previous conditions is necessary but not enough, as we will illustrate in the next paragraph.

Figure 2.4: Comparison between LWR and HTR neutron spectrum[I-2]

Table 2.4: Reactivity Parameter for some actinides in the high temperature gas reactor[I-3]

Np²³⁷ Pu²³⁸ Pu²³⁹ Pu²⁴⁰ Pu²⁴¹ Pu²⁴² Am²⁴³ Cm²⁴⁴ R.P.

HTR -27.0 -53.1 +181.1 -143.3 +212.6 -46.0 -82.5 -25.5

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2.6.3 Challenges

TRU loading is challenging from the safety point of view. In fact, TRU are material characterised by a low delayed neutron fraction² and are composed by lots of nuclides with different behaviour in the resonance region. What’s more, concerning thermal reactors, if a Pu+MA fuel is used, a higher enrichment in fissile nuclides is requested than in UOx fuel, because TRU nuclides have generally a higher capture-to-fission ratio.

Fertile material like U²³⁸ or Th²³² has a positive effect both on delayed neutron fraction and Doppler feedback, so it can “counterbalance” the effect of the TRU load. The spectrum of HTRs makes the Doppler coefficient strong negative, so that HTRs can be loaded with fertile-free fuel[2-2]. In GCFR the high mass of DU in the core has a beneficial effect on the safety parameters, nevertheless further studies are still needed.

2.7 The Level Of Mine (LOM) concept as reference parameter:

definition and discussion 2.7.1

Radiotoxicity

Radiotoxicity is a measure for the dose that a human suffers when a certain amount of radioactive nuclides enters the body[1-2]. The activity of the material, the half-life of its constituent nuclides, the type of radiation, the energy of the emitted particles, the way the radionuclides enter the human body (inhalation/ingestion), the organs that are exposed to the radiation and the time that the nuclides stay in different organs (biological half-lives) determine the value of this parameter. However we should not forget that radiotoxicity indicates a potential hazard upon ingestion and/or inhalation, thus if the radionuclides are sufficiently separated from the biosphere, no dose is suffered[1-2].

2.7.2

Level of Mine Balancing Time

Nevertheless, it is difficult to guarantee the perfect integrity of a human- made confinement beyond 10000[1-4], then in international context the Level Of Mine (LOM) is used as reference parameter. A reactor or fuel cycle has a positive impact on long term spent fuel management if it reduces the time that the spent fuel takes to reach the radiotoxicity level of the original uranium ore from which is extracted (LOMBT[2-31], Level Of Mine Balancing Time). Because the Level Of Mine is the radiotoxicity of a quantity of natural

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In order to find the LOMBT of the spent fuel from the irradiation cycles shown in the following paragraphs, we have had to calculate the corresponding LOM. We assumed 20 mSv/gU as radiotoxicity of natural uranium[I-2], while we have neglected the radiotoxicity of thorium, if present.

The radiotoxicity of SNF will depend by the initial fuel composition, the kind of reactor and the burnup. The corresponding level of mine is obtained calculating the mass of natural uranium that generated the considered quantity of SNF.

Consider for instance a quantity of natural uranium that is enriched and then used as fuel in a chain of reactors (Figure 2.5).

Figure 2.5: Sketch for LOM calculation

This equivalent natural uranium is given by going backward in the following recursive system:

( )

 



⋅ +

=

⋅

=

₋

i i i i

i 1 i i

b F I Q

f Q I

(2.5) where:

•

R

i is the ith reactor

•

I

f

ⁱ is the fuel loaded in the n_th reactor (then let us assume I₀=U_nat)

• i is the ratio between the input HM mass for the ith reactor and the output HM mass of the (i-1)_th reactor

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•

F

i is the natural uranium quantity equivalent to be added to the spent fuel from the (n-1)th reactor in order to manufacture the fuel of the nth

reactor

•

b

i is the ith reactor burn-up expressed as mass of HM discharged/mass of HM loaded (of course b0=1)

•

Q

i is the fuel discharged from the i_th reactor 2.8 The HTR as plutonium burner

In order to demonstrate the capability of HTRs to contribute in reducing nuclear waste (actinides in particular), it was decided to analyse Pu-based fuel[2-31].

In order to validate the used code (namely MONTEBURNS), an international code-to-code benchmark was performed. The obtained results have shown a good agreement between the codes adopted by the different participants, as reported in [2-31].

Particularly, 1^st generation Pu[2-32] (i.e. Pu deriving from reprocessing typical LWR spent fuel) and 2^nd generation Pu[2-32] (i.e. Pu deriving from the reprocessing of typical MOX spent fuel) were considered.

Therefore the following calculations were performed:

• HTR loaded with 1g/pebble 1^st generation Pu, 600 GWd/ton (600 EFPD)

• HTR loaded with 1.5g/pebble 2^nd generation Pu, 430 GWd/ton (645 EFPD)

To check the advantages in using different fuel cycles, the radiotoxicity evolution of spent fuel vs. time was evaluated. As reference for radiotoxicity, we, as described above, originally defined[2-31] the

LOMBT (level of mine balancing time)

, i. e. the interval of time necessary for the radiotoxicity of the exhausted fuel to return to the values of the mineral extracted from the mine (in this case of uranium) from which the fuel was generated.

The results show that 1^st generation Pu behaves (in terms of LOMBT) better than 2^nd generation Pu.

In [2-26] it was considered the possibility of adopting Th-Pu fuel cycle and, for the first time, the use of a (relatively) simplified LWR-HTR-GCFR symbiotic fuel cycle (Figure 2.6).

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It should be noted that, in the case of HTR, it was considered an once through fuel cycle. Consequently it was assumed that the waste contains Pu, MA and FP. Instead in the case of GCFR, it was considered that all the Pu will be recycled at the end of (each) GCFR step. Consequently, it was assumed that the final waste contains only MA and FP.

The obtained results are summarized in tab. 3. In column 7 the ratio between inlet and outlet transuranic elements (Pu+MA) and in column 8 the ratio between inlet and outlet actinides (Th+U+Pu+MA) are shown.

In the case of LWRs spent fuel, the LOMBT is about 250000 years[I-2]. From table 2.5 we can see that this figure of merit (the most important in terms of safety) is significantly reduced (an order of magnitude) using HTR and further (recycling Pu as fuel) by using GCFR to close the cycle.

Table 2.5: Summary of the obtained results Case Reactor Fuel LOMBT

[y] gPu

Energy

in out

Pu Pu

in out

TRU TRU

in out

Act Act

1 2 3 4 5 6 7 8

a HTR 1^stg-Pu/Th

(1/1) 37982³ 60.16 GJ 24.14% 31.75% 63.88%

b HTR 1^st g-Pu/Th

(1/2) 37547¹ 59.40 GJ 27.75% 33.88% 76.25%

c HTR 1^st g-Pu/Th

(1/3) 37117¹ 58.89 GJ 29.06% 34.37% 81.94%

d GCFR 1^st g- Pu⁴/MA/U²³⁸

(10/1/40) 9204⁵ 154.44 GJ 85.66% 86.77% 57.78%

e GCFR 2^nd g- Pu⁶/MA/U²³⁸

(10/1/40) 9310³ 155.26 GJ 85.85% 87.77% 57.57%

f GCFR⁷ 2^nd g- Pu⁴/MA/U²³⁸

(10/1/40) 11860³ 105.57 GJ 98.02% 98.95% 70.64%

3 In the case of HTR, it was assumed a once through fuel cycle. Consequently the waste contains Pu, MA and FP

4 1^st generation Pu added to Pu already present in the spent fuel of HTR loaded with 1^st generation Pu

5 In the case of GCFR, we assumed that all the Pu is recycled. Consequently the waste contains only MA and FP

6 2^nd generation Pu added to Pu already present in the spent fuel of HTR loaded with 2^nd generation Pu

7 In this case it was assumed that the thickness of the external layer of the CP is 0.003 mm instead of 0.005 mm (like cases d and e).

Therefore the spectrum is not so hard as in case e.

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2.9 Symbiotic fuel cycles

Although[2-23] water-cooled thermal reactors have reached a high stage of development and can (economically) give a significant contribution to the world energy supply, the efficient utilization of U or Th resources and the long-term management of the waste are still a challenge. As stated before, in the short-term, even if U availability seems to be adequate to fuel the continuing installation of thermal reactors, this cannot be sustained for the next centuries. Future improvements in the performance of thermal reactors can be reached particularly in terms of increased efficiency through higher temperatures of the HTR system. Furthermore, by using the fast reactors features and multiple symbiotic cycles, the utilization of uranium resources could be highly enhanced in addition to a further waste reduction.

A symbiotic fuel cycle is a strategically planned chain, where the output of a reactor is the input of the following. Each link of such a chain is a different kind of reactor (e.g. LWR or HTR or FR, etc.), because each one is able to do a different task. Of course, between two different steps, the fuel has to be cooled and reprocessed. Further data on chemical aspects of these reprocess steps are available in the international literature (e. g. documents from EU projects on HTR[1-6][2-26])

In this way, the waste radiotoxicity growth can be reduced by recycling both Pu and MA, producing, at the same time, energy.

2.9.1 A symbiotic LWR-HTR GCFR fuel cycle -

As shown, in the past the University of Pisa research group on innovative gas cooled reactors[1-9] already performed similar calculations using LWR-HTR- GCFR fuel cycles[2-26]. In figure 2.7 a flowchart of the proposed symbiotic cycle is shown.

Figure 2.7: Symbiotic fuel cycle LWR-HTR-GCFR[2-23]

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performed until k_effbecame less than 1 (about 500 GWD/tHM). The obtained isotopic composition constitutes the fissile material to be added to depleted uranium in the GCFR fuel. Finally, it has been performed the burn-up calculation of a GCFR core until k_effbecame less than 1 (about 188 GWD/tHM) and evaluated the radiotoxicity of the final waste (i.e. only the FPs, in the framework of the

full actinides recycle

strategy). For criticality reasons it was chosen for GCFR a fuel composition constituted by 30% as atomic fraction of Pu+MA (TRU) and 70% of DU (DU contains 0.25% in U²³⁵). It is important to note that DU represents a material not considered useful in order to produce energy in a "classical" way. Therefore it has a very low cost. The isotopic vectors of TRU resulting respectively from LWR and HTR that we used are shown in table 2.6.

Mass fraction Spent LWR Spent HTR Np²³⁷ 0.046 0.056 Pu²³⁸ 0.023 0.073 Pu²³⁹ 0.490 0.074 Pu²⁴⁰ 0.215 0.313 Pu²⁴¹ 0.114 0.18 Pu²⁴² 0.062 0.20 Am²⁴¹ 0.028 0.0001 Am²⁴³ 0.016 0.04 Cm²⁴² - 0.013 Cm²⁴⁴ 0.005 0.043 Cm²⁴⁵ - 0.0079 Table 2.6: TRU isotopic vectors

2.9.2 LOMBT evaluations for LWR-HTR-GCFR fuel cycles

To check the advantages of the chosen fuel cycles, it was mainly taken into account the radiotoxicity evolution of the spent fuel vs. time. As already anticipated, it was defined[2-30], as reference for radiotoxicity, the

LOMBT (Level Of Mine Balancing Time)

. This value is linked to the LOM definition, value characteristic of the type of fuel cycle and it is not easily defined (due to its oscillation range). However, in this paper, it was adopted the LOM only as a conceptual reference level.

At the end of the single steps we consider as waste the amount of:

• Pu+MA+FP in the case of HTR, because HTR has a "single step" fuel cycle (i.e. a given quantity of material spends only a burnup cycle in this kind of reactor)

• FP in the case of GCFR. In fact, in GCFR the

full actinide recycle

is foreseen (i.e. all the actinides from a irradiation step will be mixed to new DU in the “fresh” fuel for cycle continuation)

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However, in this research the analyses were limited to the first irradiation cycle in GCFR without considering what happens with the multiple recycles of the fuel.

Table 2.7 gathers the found results, while in figures 2.8 and 2.9 two typical CARL output graphs are showed.

HTR (513 EFPD) GCFR (4950 EFPD) Symbiotic cycle⁸ LOMBT [years] 63682 (waste=FP+HM) 159 (waste=FP⁹) 159 (waste=FP⁷)

Np²³⁷out/ Np²³⁷in 66% 27% 18%

Pu²³⁸_out/ Pu²³⁸_in 165% 75% 123%

Pu²³⁹_out/ Pu²³⁹_in 81% 306% 248%

Pu²⁴⁰_out/ Pu²⁴⁰_in 77% 73% 56%

Pu²⁴¹_out/ Pu²⁴¹_in 80% 26% 20%

Pu²⁴²out/ Pu²⁴²in 170% 72% 122%

Am²⁴¹out/Am²⁴¹in 21% 243% 51%

Am²⁴³out/Am²⁴³in 126% 83% 104%

Cm²⁴⁴_out/Cm²⁴⁴_in 412% 102% 420%

Table 2.7: Summary of the obtained results

Total Radiotoxicity and Time (LOMBT=63682 years)

1000000 10000000 100000000 1000000000 10000000000

0,1 1 10 100 1000 10000 100000 1000000

Time (y)

Radiotoxicity (Ingestion) Sv

Spent Fuel Radiotoxicity LOMBT

Figure 2.8: Spent fuel radiotoxicity vs. time for HTR, intermediate step (“waste”= Pu+MA+FP):

LOMBT=63682 years

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Total Radiotoxicity and Time (LOMBT=159 years)

10000 100000 1000000 10000000 100000000 1000000000 10000000000

0,1 1 10 100 1000 10000 100000 1000000

Time (y)

Radiotoxicity (Ingestion) Sv

Spent Fuel Radiotoxicity LOMBT

Figure 2.9: Spent fuel radiotoxicity vs. time for GCFR, "final" step (“waste”=FP¹⁰): LOMBT=159 years¹¹

This strategy couples the advantages of both the HTR neutron spectrum (good performance in Pu, Np²³⁷ and Am²⁴¹ burning) and the GCFR spectrum (capability of burning the higher Pu isotopes and of decreasing the Cm²⁴⁴ growing rate).

2.10 Conclusions

PBMR-400 is very flexible (as each type of HTR) from fuel type point of view.

Indeed in this chapter we have illustrated 3 different type of fuel suitable for this kind of reactor. In addition it is important to note that the discharge burn-up is not due to fuel material constraints, but only to criticality considerations. Starting from these considerations it is important to underline that this reactor can be used to transmute HLW.

The shown results[1-9] demonstrates that HTRs reduce waste radiotoxicity of an order of magnitude burning Pu and supplying energy. Adopting Th-Pu based cycle, it appears that the total amount of produced energy is quite large. Regarding the future perspectives on these topics, the material technology is probably the most challenging aspect in the field of innovative reactors and fuel cycles. In fact, the fuel has to be stable under high temperature and high fast fluence, and simple to be reprocessed and re- fabricated. These open questions are, at the moment, still to be fully solved.

10 The fission products considered here are the following: Rb⁸⁷, Sr⁹⁰, Zr⁹³, Nb⁹⁴, Tc⁹⁹, Pd¹⁰⁷, Sn¹²⁶, I¹²⁹, Cs¹³⁵, Cs¹³⁷, Sm¹⁴⁷, Sm¹⁵¹, Eu¹⁵⁴

11 If we would consider as waste besides FP, also MA, the LOMBT would become equal to 20491 years

(21)

At the end of the described symbiotic cycle, the waste results to be virtually made up only by fission products, that was separated by the actinides, whose best repository seems to be a new GCFR core itself, in order to further reduce the HLW and best exploit the uranium resources.

If the illustrated fuel cycle were feasible (mainly from a technological point of view) the LOMBT would be reduced to less than 200 years[1-9]. Please remember that in a LWR, using a

once through

cycle (even if a MOX technology is adopted), the LOMBT will be reached after more than 100000 years[1-9].

It is important also to highlight that, in the shown recycling scheme only chemical and not isotopic separations were proposed: in this way it is possible to reduce the costs and to avoid proliferation concerns linked to the use of plutonium.

Finally, considering the positive characteristics of HTR in terms of Pu burning due to their excellent neutronic economy, and coupling it with GCFR (fast neutronic spectrum and high fluence) in a symbiotic fuel cycle, it would be possible to say that the geological disposal issues concerning high level radiotoxicity of actinides can be considerably reduced.