Analysis of measured isotopic compositions by CASMO5 coupled with a JENDL-4.0 base library for irradiated fuel of light water reactors

Figures & data

Table 1. Measurement data sets on UO₂ and MOX fuels of light water reactors.

				Enrichment	Burnup (GWd/t)	Number of	Measurement
Data set ID	Plant	Assembly	Fuel	(wt %)	(irradiation cycles)	samples^a	institute^b	Ref.
PWR-UO2
SF95	Takahama-3	17×17	UO2	4.1	14-37 (2cy)	5 (c)	JAEA	[8]
SF96(Gd2O3)			UO2-Gd2O3	2.6	16-29 (2cy)	4 (i)
SF97			UO2	4.1	31-47 (3cy)	5 (s)
91E07	Ohi-1	17×17	UO2	3.2	52 (4cy)	1 (i)	JAEA	[9]
89G(Gd2O3)	Ohi-2	17×17	UO2-Gd2O3	1.7	21-29 (2cy)	3 (i)	JAEA
89G			UO2	3.2	30, 39 (2cy)	2 (i)
GGU	Gösgen	15×15	UO2	4.3	46-72 (4cy)	3 (i)	CEA/SCK-CEN/PSI	[10]
GKN II	GKN II	18×18	UO2	3.8	53 (4cy)	1 (i)	SCK-CEN	[11]
GU	Gösgen	15×15	UO2	3.5, 4.1	31-60 (3, 4cy)	3 (i)	SCK-CEN/ITU	[12]
Vandellos	Vandellos	17×17	UO2	4.5	43-78 (5cy)	6 (s)^c	Studsvik	[13]
BWR-UO2
9×9-9FP	Fukushima-daini-1	9×9-9	UO2, UO2-Gd2O3	2.1-4.9	28-68 (3, 5cy)	4 (i), 1 (c)	JAEA	[14] [15]
9×9-9Ac					36-68 (3, 5cy)	6 (i), 2 (c)
SF98	Fukushima-daini-2	8×8-2	UO2	3.91	27-44 (3cy)	6 (i)	JAEA	[8]
SF99			UO2-Gd2O3	3.41	17-37 (3cy)	8 (i)
KLU	Leibstadt	10×10^d	UO2	3.90	60.5 (7cy)	2 (i)	SCK-CEN/PSI	[16]
MOX	MOX
PWR-GGM	Gösgen (PWR)	15×15	MOX	5.5 Pufissile	70 (4cy)	1 (i)	SCK-CEN	[10]
BWR-GRM	Gundremmin-gen (BWR)	9×9-1	MOX	5.52 Pufissile	52-77 (6cy)	4 (i)	SCK-CEN	[17]

^a Alphabets in blankets show the radial location of the fuel rods in the fuel assemblies, from which the fuel samples were taken: c: corner, i: insides, and s: side.

^b JAEA: Japan Atomic Energy Agency. CEA: Commissariat á énergie atomique, SCK-CEN: Stuiecentrum voor Kernergie-Centre d’Étude de l’ énergie Nucléaire, PSI: Paul Scherrer Institute, ITU: Institute for Transuranium Elements, Joint Research Centre, European Commission.

^c During the first four cycles of five cycles, the fuel rods were irradiated at the sides of the assemblies and, during the fifth cycle, they were irradiated at the inside of the different assembly (see text).

^d SVEA 96.

Table 2. Typical analytical methods applied in the isotopic compositions [5].

	Analytical method	Applied nuclides
1	TIMS/ICPMS for isotopic composition + IDMS for absolute isotopic content	U, Pu, Am, Nd, Sm, Gd, Eu
2	External calibration with ICPMS	^237Np, ^95Mo, ^99Tc, ^101Ru, ^103Rh, ^109Ag, ^129I
3	Gamma spectrometry	^106Ru, ^125Sb, ^129I, ^134Cs, ^137Cs, ^144Cs, ^154Eu, ^155Eu
4	Liquid scintillation counting for beta spectrometry	^90Sr, ^147Pm

TIMS: Thermal ionization mass spectrometry, ICPMS: Inductively coupled plasma mass spectrometry, IDMS: Isotope dilution mass spectrometry.

Figure 1. Illustration of fuel assembly gaps between 17×17 fuel assemblies. The cells in the assemblies show the locations of fuel rods and guide tubes for control rods and instruments.

Figure 2. Locations of the fuel rods from which the fuel samples of the GU data set (GU1, GU2 and GU3 fuel samples) in the 15×15 fuel assemblies irradiated in Gösgen [12]. ‘A’ shows the location of the fuel rod for GU1. ‘B’ shows the location of the fuel rod for GU2 and GU3 in the first and second cycles and ‘C’ that in the third cycle.

Figure 3. Locations of the fuel rods from which the fuel samples of the Vandellos data set. ‘A,’ ‘B’ and ‘C’ shows the locations in the first four cycles and ‘D,’ ‘E’ and ‘F’ those in the fifth cycle [13].

Figure 4. Illustration of the assembly configuration in the multiple-assembly model.

Table 3. Numbers of samples, averages and standard deviations (%) of C/E − 1's for actinide nuclides on the UO₂ samples.

	PWR			BWR
Nuclide	Number of samples	Average	Standard deviation	Number of samples	Average	Standard deviation
^234U	32	12.12	13.27	29	5.75	14.33
^235U	32	2.73	5.20	29	6.67	7.61
^236U	32	−0.94	2.64	29	−1.45	1.24
^238U	27	0.09	0.47	29	−0.05	0.23
^237Np	23	−6.57	10.21	16	−3.36	10.85
^238Pu	33	−4.40	7.75	29	5.21	8.86
^239Pu	33	0.32	3.85	29	2.54	5.17
^240Pu	33	0.73	3.62	29	−0.60	4.12
^241Pu	33	−2.40	3.65	29	0.44	4.72
^242Pu	32	−1.72	4.35	29	−0.13	4.44
^241Am	29	3.89	16.52	16	−6.58	15.09
^242mAm	23	−6.58	13.81	16	8.01	17.20
^243Am	30	−4.64	8.36	16	−3.42	5.18
^242Cm	23	−10.35	17.28	15	−44.46	22.54
^243Cm	20	−7.37	13.87	16	3.68	38.79
^244Cm	33	8.05	35.99	16	−2.91	7.86
^245Cm	23	−5.74	11.04	16	1.12	7.97
^246Cm	27	−3.41	36.94	15	−17.43	26.33
^247Cm	6	−29.30	3.62	4	−64.60	29.68

Table 4. Numbers of samples, averages and standard deviations (%) of C/E − 1's for fission product nuclides on UO₂ samples.

	PWR			BWR
Nuclide	Number of samples	Average	Standard deviation	Number of samples	Average	Standard deviation
^90Sr	6	−4.57	10.03	2	−2.00	0.67
^95Mo	7	−3.93	9.55	7	0.45	9.69
^99Tc	9	7.01	10.01	2	24.77	14.73
^101Ru	7	9.42	13.99	7	4.92	11.67
^106Ru	31	9.13	22.63	16	63.22	57.83
^103Rh	8	6.77	11.38	7	4.33	7.80
^109Ag	6	17.58	54.08	7	−7.66	7.08
^125Sb	25	−24.48	16.52	11	−23.81	18.83
^129I	5	−17.47	11.97	–	–	–
^133Cs	12	1.89	4.44	7	−0.23	2.89
^134Cs	32	−1.24	5.81	21	−1.09	6.10
^135Cs	12	−6.21	8.37	2	−2.73	0.70
^137Cs	33	−0.94	3.60	21	−5.22	3.82
^139La	4	6.73	2.66	–	–	–
^140Ce	6	−1.99	3.08	–	–	–
^142Ce	6	−1.36	1.03	–	–	–
^144Ce	32	6.38	11.79	15	−4.18	12.78
^142Nd	22	−2.24	15.95	2	1.91	4.31
^143Nd	33	−0.04	4.01	29	2.07	2.97
^144Nd	27	−4.25	5.91	29	−2.26	5.63
^145Nd	33	−0.54	1.38	29	−0.13	0.65
^146Nd	33	0.19	1.77	29	0.48	0.62
^148Nd	33	−1.14	1.42	29	−0.31	0.29
^150Nd	33	−0.54	2.43	29	0.52	0.68
^147Pm	6	14.69	19.41	7	11.08	5.22
^147Sm	17	2.93	4.04	18	−9.66	9.71
^148Sm	18	−4.52	5.05	18	−9.25	4.41
^149Sm	18	1.49	10.07	18	1.41	11.41
^150Sm	18	−5.97	3.47	18	−7.65	3.69
^151Sm	18	12.38	6.59	18	9.84	7.22
^152Sm	18	9.29	3.49	18	4.14	7.21
^154Sm	18	−0.11	8.11	18	−6.85	7.93
^153Eu	12	4.35	5.91	7	11.12	3.70
^154Eu	32	9.38	11.35	21	1.79	6.96
^155Eu	12	6.25	12.23	7	−4.74	6.45
^154Gd	6	42.49	16.52	6	4.53	11.15
^155Gd	12	14.34	22.57	7	1.56	10.10
^156Gd	6	7.33	6.09	6	3.28	4.35
^157Gd	–	–	–	5	−46.13	20.75
^158Gd	6	25.41	6.37	6	13.35	8.84
^160Gd	6	3.25	9.25	6	−2.75	11.23

Table 5. C/E − 1's and uncertainties (%) in C/E − 1's for the PWR GGM MOX sample, and averages and standard deviations (%) of C/E − 1's for the BWR GRM MOX samples on actinide nuclides for the single-assembly (Single A) and multiple-assembly models (Multi. A).

	PWR				BWR
	Single A		Multi. A			Single A		Multi. A
Nuclide	C/E−1	Uncertainty	C/E−1	Uncertainty	Number of samples	Average	Standard deviation	Average	Standard deviation
^234U	5.28	0.47	2.18	0.46	4	−0.14	2.95	−4.71	2.22
^235U	6.21	0.41	4.70	0.41	4	11.72	2.67	9.43	1.47
^236U	−2.35	0.32	−3.30	0.31	4	−1.33	1.91	−2.60	1.31
^238U	0.52	0.33	0.85	0.33	4	−0.26	1.27	−0.06	1.28
^237Np	9.74	9.15	−0.91	8.26	4	30.45	7.99	17.69	7.55
^238Pu	−2.93	1.37	−4.95	1.34	4	0.99	3.54	−1.92	2.58
^239Pu	13.57	0.38	1.96	0.34	4	21.41	6.79	10.17	6.99
^240Pu	0.17	0.34	−1.38	0.33	4	−0.11	0.97	−4.54	0.92
^241Pu	5.08	0.39	2.13	0.38	4	8.24	1.95	7.58	3.88
^242Pu	−3.19	0.33	1.13	0.34	4	−6.80	2.65	−2.38	2.70
^241Am	49.29	2.41	40.42	2.26	4	20.31	4.05	15.66	4.94
^242mAm	36.17	3.47	14.75	2.92	4	33.62	11.57	15.84	10.57
^243Am	16.29	2.11	13.36	2.05	4	0.79	1.11	1.35	0.71
^242Cm	–		–		4	37.27	21.47	32.44	20.96
^243Cm	13.33	5.33	14.66	5.39	2	7.78	0.13	9.29	0.61
^244Cm	−2.29	2.04	−4.25	1.99	4	−0.29	4.21	−2.09	2.78
^245Cm	5.44	2.17	−5.88	1.94	4	13.34	7.30	2.67	7.21
^246Cm	−16.03	2.34	−21.02	2.20	4	−16.52	5.59	−21.79	4.01

Table 6. C/E − 1's and uncertainties (%) in C/E − 1's for the PWR GGM MOX sample, and averages and standard deviations (%) of C/E − 1's for the BWR GRM MOX samples on fission product nuclides for the single-assembly (Single A) and multiple-assembly (Multi. A) models.

	PWR				BWR
	Single A		Multi. A			Single A		Multi. A
Nuclide	C/E−1	Uncertainty	C/E−1	Uncertainty	Number of samples	Average	Standard deviation	Average	Standard deviation
^90Sr	−15.88	4.47	−15.88	4.47	1	−12.46 ^a	5.15 ^b	−12.28 ^a	5.16 ^b
^95Mo	10.45	4.75	10.64	4.76	2	2.31	1.98	2.45	1.94
^99Tc	21.63	5.16	22.42	5.20	2	16.34	1.80	17.00	1.62
^101Ru	30.70	5.40	31.16	5.41	2	32.34	3.90	32.86	3.99
^106Ru	21.72	1.46	36.20	1.63	1	26.43 ^a	2.14 ^b	38.28 ^a	2.34 ^b
^103Rh	20.00	4.75	19.17	4.72	2	9.53	7.70	8.84	7.51
^109Ag	−7.13	5.80	−6.02	5.87	2	−11.65	4.05	−9.86	3.81
^125Sb	−26.17	1.35	−22.46	1.41	1	−8.29 ^a	1.81 ^b	−3.25 ^a	1.91 ^b
^129I	−21.43	4.46	−21.77	4.45	1	−8.01 ^a	5.48 ^b	−8.33 ^a	5.46 ^b
^133Cs	0.21	2.45	1.14	2.47	4	−2.83	1.50	−1.96	1.56
^134Cs	0.32	1.79	0.98	1.80	4	2.11	2.36	2.17	1.51
^135Cs	1.76	2.12	−2.06	2.04	4	2.74	1.20	−0.56	1.19
^137Cs	−1.41	1.23	−0.99	1.24	4	−3.91	1.05	−3.22	1.11
^144Ce	−3.86	2.37	9.14	2.69	4	−0.77	0.38	6.82	0.95
^142Nd	3.37	2.06	0.41	2.00	4	−5.36	10.11	−7.33	8.63
^143Nd	2.57	0.36	1.35	0.36	4	3.20	0.76	2.09	0.32
^144Nd	−2.30	0.33	−1.78	0.34	4	−5.66	1.70	−4.91	1.53
^145Nd	−0.49	0.35	−0.14	0.36	4	−2.21	0.95	−1.87	1.02
^146Nd	0.33	0.36	−0.53	0.36	4	−1.51	1.23	−2.00	1.06
^148Nd	0.54	0.39	0.39	0.39	4	−1.47	1.38	−1.42	1.34
^150Nd	0.90	0.39	0.86	0.39	4	−0.87	1.92	−0.76	1.87
^147Pm	−7.59	4.40	−1.41	4.69	3	−3.25	2.27	1.79	2.00
^147Sm	−1.04	0.93	0.25	0.94	4	−4.49	2.09	−3.53	2.05
^148Sm	−5.27	0.88	−9.33	0.84	4	−1.83	3.01	−5.67	2.12
^149Sm	17.90	2.03	8.85	1.88	4	22.72	11.44	14.48	9.66
^150Sm	−12.24	0.80	−11.15	0.81	4	−11.84	1.83	−10.49	1.83
^151Sm	27.30	1.15	14.92	1.04	4	29.50	11.23	19.81	11.32
^152Sm	4.78	1.06	8.67	1.10	4	2.34	4.23	5.22	4.45
^154Sm	−4.61	0.84	−4.43	0.84	4	−3.09	2.60	−2.54	2.64
^153Eu	5.20	0.58	5.89	0.58	4	4.07	1.60	5.08	1.60
^154Eu	14.93	1.77	7.47	1.65	4	15.75	3.46	10.07	4.18
^155Eu	−0.40	2.37	−3.05	2.31	4	−10.75	1.56	−12.77	1.51
^155Gd	12.20	0.48	7.72	0.46	4	−3.91	7.87	−6.68	7.36

^a C/E−1 (%) of one sample. ^b Uncertainty (%) of one sample.

Figure 5. Effect of a gap between PWR fuel assemblies on C/E−1's of ²³⁵U for the data sets of Takahama-3 (PWR).

Figure 6. Effect of a gap between PWR fuel assemblies on C/E−1's of ²³⁹Pu for the data sets of Takahama-3 (PWR).

Figure 7. Effect of a gap between PWR fuel assemblies on C/E−1's of ¹⁴⁸Nd for the data sets of Takahama-3 (PWR).

Figure 8. Radial positions of the fuel rods from which fuel samples were taken for the 9×9-9FP and 9×9-9Ac data sets of Fukushima-daini-1 (BWR) [14].

Figure 9. C/E − 1's of ²³⁵U vs. the axial positions of the fuel samples for the 9×9-9FP and 9×9-9Ac data sets of Fukushima-daini-1 (BWR).

Figure 10. C/E − 1's of ²³⁹Pu vs. the axial positions of the fuel samples for the 9×9-9FP and 9×9-9Ac data sets of Fukushima-daini-1 (BWR).

Figure 11. C/E − 1's of ¹⁴⁸Nd vs. the axial positions of the fuel sampl es for the 9×9-9FP and 9×9-9Ac data sets.

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