Utilization of rock-like oxide fuel in the phase-out scenario

Figures & data

Figure 1. Layer structure of the NMB code.

Figure 2. Layer structure of the waste management part.

Table 1. Nuclides in the NMB code.

Nuclide	Half-life	Nuclide	Half-life
Th-232	1.405E+1y	Pu-240	6.564E+3y
Pa-231	3.276E+4y	Pu-241	14.35y
Pa-233	26.967d	Pu-242	3.733E+5y
U-232	68.9y	Am-241	432.19983y
U-233	1.592E+5y	Am-242m	141y
U-234	2.455E+5y	Am-243	7.37E+3y
U-235	7.038E+8y	Cm-242	162.8d
U-236	2.342E+7y	Cm-243	29.1y
U-237	6.75d	Cm-244	18.1y
U-238	4.468E+9y	Cm-245	8.5E+3y
Np-237	2.14E+6y	Cm-246	4.73E+3y
Np-238	2.117d	I-129	1.57E+7y
Np-239	2.3565d	Tc-99	2.111E+5y
Pu-238	87.74y	Other FPs	-
Pu-239	2.411E+4y

Figure 3. Effect of time mesh of the multiple expansion method.

Figure 4. Effect of time step to update neutron flux.

Figure 5. Effect of the order of expansion.

Figure 6. Comparison between the NMB and ORIGEN codes (PWR).

Figure 7. Composition change of PWR ROX.

Figure 8. Comparison between the NMB and SRAC codes (PWR ROX).

Table 2. Scenarios.

Index	Note	Reprocess	Waste
OT	Once-through	No^*	Glass, UO2 SF
M0	MOX loading to LWRs up to 30%.	Yes	Glass, MOX SF
M1	Full-MOX loading to LWRs
M2	MOX loading up to 30%, two full-MOX BWRs
R0	ROX loading to LWRs up to 30%	Yes	Glass, ROX SF
R2	ROX loading up to 30%, two full-ROX BWRs

^*No, but reprocessing in oversea facilities is counted.

Table 3. Core parameters.

Plant type	Plant life (yr)	Thermal efficiency	Load factor	Batch length (EFPY)	Batch number
PWR-LB	40	33%	85%	0.803	3
PWR-MB	40	33%	85%	1.095	3
BWR-LB	40	33%	85%	1.176	3
BWR-MB	40	33%	85%	1.203	4

Notes: “LB” = low burn-up; “MB” = middle burn-up.

Table 4. Fuel parameters.

Fuel type	Plant type	Power density (MW/t)	k-infinity	^235U enrichment	Limit of Pu
UO2	PWR-LB	37.5	1.354	3.4%	–
	PWR-MB	37.5	1.394	4.1%	–
	BWR-LB	25.6	1.424	3.0%	–
	BWR-MB	25.6	1.495	3.7%	–
MOX	PWR-LB	37.5	1.253	–	12%
	PWR-MB	37.5	1.253	–	12%
	BWR-LB	25.6	1.349	–	12%
	BWR-MB	25.6	1.349	–	12%
ROX	PWR-MB	250	^*		45%
	BWR-MB	171	^*		45%

^*Pu ratio is fixed at 45% in the ROX fuel.

Figure 9. Electricity generation and reprocessed amount of the “OT” scenario.

Figure 10. Pu and MA inventory of the “OT” scenario.

Figure 11. Electricity generation and reprocessed amount of the “M0” scenario.

Figure 12. Pu and MA inventory of the “M0” scenario.

Figure 13. MOX ratio in LWRs except for the full-MOX reactor in the “M0” and “M1” scenarios.

Figure 14. Pu and MA inventory of the “M1” scenario.

Figure 15. Electricity generation and reprocessed amount of the “M2” scenario.

Figure 16. Pu and MA inventory of the “M2” scenario.

Figure 17. Electricity generation and reprocessed amount of the “R0” scenario.

Figure 18. Pu and MA inventory of the “R0” scenario.

Figure 19. Electricity generation and reprocessed amount of the “R2” scenario.

Figure 20. Pu and MA inventory of the “R2” scenario.

Figure 21. Comparison of Pu inventory in 2100.

Table 5. Nonproliferation parameters of Pu in SF.

	WG^*1	UO2	MOX	ROX
BCM (kg)^*2	10.6	13.3	15.6	18.9
SF (n/s/kg)^*3	7.2E+4	4.1E+5	6.5E+5	8.9E+5
Heat (W/kg)	2.3	16.9	27.6	30.7

^*1: weapon grade; ^*2 BCM; ^*3 spontaneous fission neutron production.

Figure 22. Isotopic vectors of Pu in the SFs after 5 yr cooling.

Figure 23. BCM of Pu in each SF.

Figure 24. Spontaneous fission neutron generation of Pu in each SF.

Figure 25. Heat generation of Pu in each SF.

Figure 26. Weight of SF in 2100.

Figure 27. Number of SF casks in 2100.

Figure 28. Repository footprint in 2100.

Figure 29. Potential radiotoxicity.

Figure A.1. Model of SF cask in the case of N = 4.

Table A.1. Minimum value of x (= x_min).

Rock type^*	Hard	Soft
Vertical	2.0	3.0
Horizontal	2.0	2.5

^*Hard = crystalline rock; soft = sedimentary rock.

Figure A.2. Model of emplacement.

Figure A.3. Determination of the number of assemblies per cask and interval of emplacement to minimize footprint.

Figure B.1. Model of rock around repository.

Table B.1. Comparison of the maximum buffer temperature to the reference [10] (PWR 45GWd/tHM, 54-yr cooling).

Rock type^*		Soft				Hard
Emplacement type		Vertical		Horizontal		Vertical		Horizontal
N per cask		2	4	2	4	2	4	2	4
Reference	°C	67.8	90.0	90.0	90.0	90.0	95.0	90.0	95.0
Present	°C	68	89	90	90	89	99	91	99
Present reference	°C	−0.1	−0.7	0.5	−0.2	−0.5	3.9	1.0	3.9

^*Hard = crystalline rock; soft = sedimentary rock.

Japan Atomic Energy Commission. Report on comparison of nuclear fuel cycle cost in the basic scenario [in Japanese]. Tokyo: Atomic Energy Commission; 2004. Available from: http://www.aec.go.jp/jicst/NC/iinkai/teirei/siryo2004/kettei/sakutei041124.pdf

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