**6.3. Waste volume reduction and significance of P&T**

P&T is expected to reduce waste volume and repository footprint [38]. However, partitioning and/or transmutation cannot reduce the inventory of waste nuclides itself. P&T reduces waste package volume by conquering the technical problem of vitrified waste fabrication [39].

> There are two representative scenarios for P&T scheme [42]. Those are mainly optimized for the cooling time of Sr-Cs calcined waste. One hundred thirty and 300 years of cooling scenarios are employed. The specifications for LWR-SF (45 GWd/t) of 32,000 tIHM are listed in **Table 6**. Those are evaluated based on Ref. [38]. The P&T with 130 years of cooling can realize

Glass 50 V<sup>0</sup> 40,000 1,776,000

HWL glass 5 V<sup>1</sup> 8300 184,260 Sr-Cs 130 V<sup>0</sup> 5100 226,440 Total 13,400 410,700

HWL glass 45 C 8300 7885 Sr-Cs 300 C 5100 4845 Total 13,400 12,730

HWL glass 85 V<sup>0</sup> 8300 368,520 Sr-Cs 150 V<sup>1</sup> 5100 113,220 Total 13,400 481,740

/cani., V<sup>1</sup>

is 22.2 m2

/cani., and C is 0.95 m<sup>2</sup>

/cani [38].

is 44.4 m2

**Table 6.** Specifications of disposal for LWR-SF (45 GWd/t) of 32,000 tIHM.

**Cooling time (y) Configuration No. of package Footprint (m2**

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**)**

**Figure 16.** Number of waste packages per electricity generation [38].

Non-P&T

Partitioning

P&T with 130 years of cooling

P&T with 300 years of cooling

Footprint of emplacement configuration V<sup>0</sup>

For vitrified waste fabrication, there are limitations for heat generation (decay heat), waste content (FPs and MAs), platinum group metal (PGM) content, and molybdenum oxide content. The heat generation is limited to remain temperature of waste lower than 500°C during storage to prevent the phase transmutations such as crystallization and liquid-liquid phase separation at elevated temperatures. The waste content is limited to remain characteristics of glass for the confinement of the waste. The PGM content is limited not to shorten the lifetime of liquid-fed ceramic melter (LFCM). The molybdenum oxide content is limited to prevent the formation of molybdenum-rich phase, which is called yellow phase and degrades chemical durability of the vitrified form.

By partitioning [40], the PGM is recovered and used as resources. Strontium and cesium, whose decay heat is dominant, are partitioned and converted to Sr-Cs calcined waste. By employing high-waste-loading glass [41], high content of waste and molybdenum can be contained into the vitrified form.

To confirm the reduction of waste volume by P&T, numbers of waste package generation of four cycle schemes for LWR are compared as shown in **Figure 16**. The schemes are non-P&T, only transmutation, only partitioning, and P&T schemes. To reduce the number, partitioning is the most effective. The effect of high-waste-loading glass is dominant. The P&T scheme generates more waste packages than the partitioning scheme.

However, the partitioning is optimized to minimize the waste package generation and not optimized to minimize the repository footprint because the heat generation from Sr-Cs calcined waste is problematic to dispose. The repository footprint is mainly determined by heat generation from the waste. The buffer material of bentonite should be remained under the temperature of 100°C. The waste package pitches for disposal determined by the limitation of the buffer material temperature. In other words, the waste package with lower heat generation can realize lower footprint. Therefore, to dispose the Sr-Cs calcined waste, long cooling time is necessary.

Safety and Economics of Uranium Utilization for Nuclear Power Generation http://dx.doi.org/10.5772/intechopen.72647 41

**Figure 16.** Number of waste packages per electricity generation [38].

ADS also attracted a lot of attention. In this situation, an expert committee of Atomic Energy Society of Japan (AESJ) published the report for direct disposal [35]. In this report, the safety of geological disposal and public opinion were researched and discussed. It is emphasized that the potential toxicity cannot be the index directly to assess the safety, and the safety of geological disposal should be assessed by public exposure. The expert committee states its own view that the safety of geological disposal tends to be assessed by the potential toxicity

If the potential toxicity would be gotten public support as the hazard index and all MAs and LLFPs would be transmutated, the waste should be managed at least 300 years. Furthermore, if all radioactive nuclides would be transmutated to stable nuclides, the waste should be man-

P&T is expected to reduce waste volume and repository footprint [38]. However, partitioning and/or transmutation cannot reduce the inventory of waste nuclides itself. P&T reduces waste package volume by conquering the technical problem of vitrified waste fabrication [39].

For vitrified waste fabrication, there are limitations for heat generation (decay heat), waste content (FPs and MAs), platinum group metal (PGM) content, and molybdenum oxide content. The heat generation is limited to remain temperature of waste lower than 500°C during storage to prevent the phase transmutations such as crystallization and liquid-liquid phase separation at elevated temperatures. The waste content is limited to remain characteristics of glass for the confinement of the waste. The PGM content is limited not to shorten the lifetime of liquid-fed ceramic melter (LFCM). The molybdenum oxide content is limited to prevent the formation of molybdenum-rich phase, which is called yellow phase and degrades chemical durability of the vitrified form.

By partitioning [40], the PGM is recovered and used as resources. Strontium and cesium, whose decay heat is dominant, are partitioned and converted to Sr-Cs calcined waste. By employing high-waste-loading glass [41], high content of waste and molybdenum can be contained into

To confirm the reduction of waste volume by P&T, numbers of waste package generation of four cycle schemes for LWR are compared as shown in **Figure 16**. The schemes are non-P&T, only transmutation, only partitioning, and P&T schemes. To reduce the number, partitioning is the most effective. The effect of high-waste-loading glass is dominant. The P&T scheme

However, the partitioning is optimized to minimize the waste package generation and not optimized to minimize the repository footprint because the heat generation from Sr-Cs calcined waste is problematic to dispose. The repository footprint is mainly determined by heat generation from the waste. The buffer material of bentonite should be remained under the temperature of 100°C. The waste package pitches for disposal determined by the limitation of the buffer material temperature. In other words, the waste package with lower heat generation can realize lower footprint. Therefore, to dispose the Sr-Cs calcined waste, long cooling time is necessary.

in the recent society because it is easy to understand intuitively.

aged due to the toxicity of heavy metal.

the vitrified form.

**6.3. Waste volume reduction and significance of P&T**

40 Uranium - Safety, Resources, Separation and Thermodynamic Calculation

generates more waste packages than the partitioning scheme.

There are two representative scenarios for P&T scheme [42]. Those are mainly optimized for the cooling time of Sr-Cs calcined waste. One hundred thirty and 300 years of cooling scenarios are employed. The specifications for LWR-SF (45 GWd/t) of 32,000 tIHM are listed in **Table 6**. Those are evaluated based on Ref. [38]. The P&T with 130 years of cooling can realize


**Table 6.** Specifications of disposal for LWR-SF (45 GWd/t) of 32,000 tIHM.

1/4 of footprint compared with that of non-P&T. With 300 years of cooling, 1/100 of footprint can be realized. However, only the partitioning can also realize 1/4 of footprint with 150 years of cooling for Sr-Cs calcined waste and 85 years of cooling for the high-waste-loading glass. The waste package including MAs needs the cooling time to decay <sup>244</sup>Cm, whose half-life is 18.1 years. For the rest decay heat, 241Am, whose half-life is 433 years, is dominant and difficult to reduce by cooling. To realize more compact disposal, the transmutation is necessary.

The technology of partitioning is already demonstrated [40, 41]. On the other hand, transmutation should develop many innovative technologies concerning to neutron source by spallation reaction, Pb-Bi FR core, and pyro reprocessing for ADS and reactor core and advanced reprocessing for FBR. The partitioning technology without transmutation is preferable as early introduction option to suit uranium utilization.

#### **6.4. Environmental burden with P&T**

As described in the previous section, the safety of waste disposal should not be assessed by the potential toxicity. However, reduction of the potential toxicity is one of the objectives to develop fuel cycle system with FBR in Japan. All TRU nuclide is planned to recycled with the recovery ratio of 99.9% [43] to shorten the cooling time needs to decay the toxicity under the natural uranium level within 300 years.

However, with the recovery, the public dose from MAs would not be reduced. The MAs in 4N+1 decay series are problematic. 229Th is the daughter of 237Np, and other MAs in 4N+1 decay series, that is, 241Pu, 241Am, and 245Cm, are decay to 237Np. These nuclides should be recovered with high recovery ratio. It is said that the recovery ratio should be higher than 99.998% [44]. The relations of the recovery ratio to the molar flux of radioactive nuclide from EBS are shown in **Figure 17**. 135Cs can be reduced with higher recovery ratio because that

**High decontamination**

**MOX**

**MOX**

**+Np**

> FBR-MOX

(114.9 GWd/t)

LWR-UOX

Standard burnup

(BWR 45 GWd/t)

High burnup (PWR 60 GWd/t)

LWR-MOX

Standard burnup

(BWR 45 GWd/t)

High burnup (PWR 60 GWd/t)

Hand

0.4 \*The dose rates are normalized by the dose in representative fuel fabrication with spent fuel from Fugen [43].

\*\*Dose rate lower than 1 for the whole body and hand: fabrication in GB is possible with the current technology.

Dose rate lower than two for the whole body and hand: fabrication in GB is possible with the current technology improving the process.

Dose rate lower than 10 for the whole body and hand: automation is necessary for fabrication in GB.

Dose rate higher than 10 for the whole body and hand: fabrication in GB is impossible [43].

**Table 7.**

Feasibility of fuel fabrication in globe box [43].

0.6

20

50

200

200

200

200

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Whole body

2

2

10

400

700

700

900

1000

Hand

0.3

0.43

8

20

100

100

100

200

Whole body

2

2

8

100

600

600

700

800

Hand

0.2

0.3

4

10

200

200

200

200

Whole body

1

1

4

60

700

700

900

1000

Hand

0.2

0.2

3

7

100

100

100

100

Whole body

0.8

0.9

3

40

600

600

700

700

Whole body

Hand

0.1

0.1

3

9

300

300

300

300

0.7

0.7

9

60

1000

1000

2000

2000

**+Np + Am**

**+Np + Am + Cm**

**MOX**

**MOX**

**MOX**

**MOX**

**+Np**

**+Np + Am**

**+Np + Am + Cm**

**MOX**

**MOX**

**Low decontamination**

**Figure 17.** Relation of recovery ratio to the molar flux of radioactive nuclide from EBS [44].


**Figure 17.** Relation of recovery ratio to the molar flux of radioactive nuclide from EBS [44].

1/4 of footprint compared with that of non-P&T. With 300 years of cooling, 1/100 of footprint can be realized. However, only the partitioning can also realize 1/4 of footprint with 150 years of cooling for Sr-Cs calcined waste and 85 years of cooling for the high-waste-loading glass. The waste package including MAs needs the cooling time to decay <sup>244</sup>Cm, whose half-life is 18.1 years. For the rest decay heat, 241Am, whose half-life is 433 years, is dominant and difficult to reduce by cooling. To realize more compact disposal, the transmutation is necessary.

The technology of partitioning is already demonstrated [40, 41]. On the other hand, transmu

tation should develop many innovative technologies concerning to neutron source by spall

early introduction option to suit uranium utilization.

42 Uranium - Safety, Resources, Separation and Thermodynamic Calculation

**6.4. Environmental burden with P&T**

natural uranium level within 300 years.

ation reaction, Pb-Bi FR core, and pyro reprocessing for ADS and reactor core and advanced reprocessing for FBR. The partitioning technology without transmutation is preferable as

As described in the previous section, the safety of waste disposal should not be assessed by the potential toxicity. However, reduction of the potential toxicity is one of the objectives to develop fuel cycle system with FBR in Japan. All TRU nuclide is planned to recycled with the recovery ratio of 99.9% [43] to shorten the cooling time needs to decay the toxicity under the

However, with the recovery, the public dose from MAs would not be reduced. The MAs in 4N+1 decay series are problematic. 229Th is the daughter of 237Np, and other MAs in 4N+1 decay series, that is, 241Pu, 241Am, and 245Cm, are decay to 237Np. These nuclides should be recovered with high recovery ratio. It is said that the recovery ratio should be higher than 99.998% [44]. The relations of the recovery ratio to the molar flux of radioactive nuclide from EBS are shown in **Figure 17**. 135Cs can be reduced with higher recovery ratio because that



**Table 7.** Feasibility of fuel fabrication in globe box [43].

Dose rate higher than 10 for the whole body and hand: fabrication in GB is impossible [43].

dissolved in groundwater congruently with glass form dissolution. On the other hand, the concentration of 237Np is limited by solubility. Then, the inventory of 237Np should be reduced lower than the amount corresponding to the solubility. The MA recycling may not contribute to reduce public dose with the recovery ratio of 99.9%.

For environmental burden, the safety of geologic disposal for existing LWR waste is secured by evaluating public dose with a sufficient margin. However, P&T is planned to reduce the potential toxicity, which the index should not be used for safety assessment. To reduce waste volume, P&T is effective. Only with partitioning, the repository footprint is reduced to 1/4 times. However, transmutation of MAs cannot reduce the public dose with the recovery ratio of 99.9% determined to reduce the potential toxicity. MA recycling with FBR increases the

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As discussed above, uranium utilization in thermal reactor can achieve safe and sustainable

[1] Walter A, Reynolds A. Fast Breeder Reactors. New York: Pergamon Press; 1981. 700 pp [2] IAEA. Passive Safety Systems and Natural Circulation in Water Cooled Nuclear Power

[3] Bouteille F, Azarian G, Bittermann D, et al. The EPR overall approach for severe accident

[4] AESJ. The Fukushima Daiichi Nuclear Accident: Final Report of the AESJ Investigation

[5] Ohashi H, Sato H, Tachibana Y, et al. Concept of an inherently-safe high temperature gas-cooled reactor. ICANSE 2011; 14-17 November 2011; Bali; 2011. pp. 50-58

[6] The Cabinet of Japan. Strategic Energy Plan [Internet]. 2014. Available from: http://www. enecho.meti.go.jp/en/category/others/basic\_plan/pdf/4th\_strategic\_energy\_plan.pdf

[7] Lewis E. Nuclear Power Reactor Safety. New York: John Wiley & Sons inc.; 1977. 648 pp [8] Massimo L, Physics of High Temperature Reactors. New York: Pergamon Press; 1979. 220 pp [9] Sun K. Analysis of Advanced Sodium cooled Fast Reactor Core Designs with Improved Safety Characteristics. Swiss Federal Institute of Technology Lausanne (EPFL) doctoral

[10] Wade D, Chang Y. The integral fast reactor concept: Physics of operation and safety. Nuclear

working environmental burden due to the increased dose.

energy supply with acceptable environmental burden.

\*Address all correspondence to: fukaya.yuji@jaea.go.jp

Plants. IAEA-TECDOC-1624, IAEA, Vienna; 2009

Committee. Tokyo: Springer; 2015. 560 pp

Science and Engineering. 1988;**100**:507-524

mitigation. Nuclear Engineering and Design. 2006;**236**:1465-1470

Japan Atomic Energy Agency, Japan

[Accessed: 02-10-2017]

thesis; 2012. N5480

**Author details**

Yuji Fukaya

**References**

Furthermore, MA recycling increases working environment burden like the concept of PPP. MA recycling makes difficulty not only for spent fuel but also for fuel fabrication. **Table 7** lists the feasibility of fuel fabrication in globe box (GB) [43]. MOX fuel and neptunium-doped MOX fuel with high decontamination can be fabricated in GB with the current technology. However, for americium- and/or curium-doped fuel, automation is necessary, or fuel fabrication in GB is impossible. Fuel with low decontamination cannot be fabricated in GB. In this context, there is the opinion that MA recycling should not be performed [45]. For nuclear proliferation, safeguard should be enhanced by increasing the transparency of society instead of MA recycling [45].
