Journal of Nuclear and Radiochemical Sciences, Vol. 4, No. 1, pp. 9-13, 2003

ArticlesAnalysis of Irradiated 237Np in the Experimental Fast Reactor JOYO for the Evaluation of
its Transmutation Behavior in a Fast Reactor
Masahiko Osaka,*, a Shin-ichi Koyama,a and Toshiaki MitsugashirabaAlpha Gamma Section, O-arai Engineering Center, Japan Nuclear Cycle Development Institute,
Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1393, Japanb
The Oarai-branch, Institute for Materials Research, Tohoku University,
Oarai-machi, Higashiibaraki-gun, Ibaraki 311-1313, Japan
Received: April 21, 2003; In Final Form: June 27, 2003
237

Np oxide targets irradiated in the experimental fast reactor JOYO were analyzed by applying sophisticatedradiochemical technique. Pu and Np were isolated from the irradiated targets by anion exchange chromatography.
Total fission events of each target were estimated by 137Cs method. 237Np and 236Pu were determined by alpha
spectrometry. The isotopic ratio of Pu and the content of Pu were determined by thermal ionization mass spectroscopy and isotope dilution mass spectrometry, respectively. From these analytical results, the transmutation
behavior of 237Np in a fast reactor was discussed. Up to about 528 EFPD (Effective Full Power Days) irradiation,
the transformation through capture reaction was much larger than the incineration through fission reaction,although the latter increased comparatively in case of irradiation in hard neutron spectrum. 236Pu was clearlyobserved in all irradiated 237Np samples but its quantity was much smaller than the estimated value on the basis ofpresent nuclear data library.

Nuclides formed by(n, γ) reaction

1. Introduction
The development of future nuclear fuel cycle systememphasized on fast reactor strategy is now under way in JNC(Japan Nuclear Cycle Development Institute).1 In this nuclear
cycle system, 237Np is one of the target materials to be recovered from spent nuclear fuel and transmuted in a fast reactor
core to reduce environmental impact caused by its long andhigh radiotoxity. MOX (U and Pu mixed oxide) fuel containing
several percent of Np, which often includes other MA (minor
actinides) such as Am and some FP (fission products), is
considered as a promising candidate for the recycled fuel infast reactors.2, 3 This MA contained type MOX fuel is now
developed in JNC and will be irradiated in the experimental
fast reactor JOYO to evaluate the irradiation and transmutation
behavior.4 For the evaluation of transmutation behavior, acalculation code must be prepared on the basis of precise
nuclear data with high accuracy. Pure 237Np samples were
fabricated and irradiated in JOYO partly for this purpose,5
since the pure samples could eliminate the influence of 237Np
that was formed during irradiation from precedents such as
238
U and thus make the direct determination of MA transmutedfrom 237Np possible.
Nuclear transformation chain started from 237Np in a reactor
is shown in Figure 1.6, 7 The main neutron capture reaction of237
Np leads to the formation of many Pu isotopes and the fastneutron induced (n,2n) reaction leads to the formation of longlived nuclides, 236Np, 236U, and 236Pu. The thermal fission crosssection of 237Np is only 2.15 × 10−2 b (10−28 m2). Although thethermal fission cross section of 236Np has not been measured so
accurately, this nuclide is known for its very large thermal
fission cross section of about 2500 b.6 Thus, the determinationof 236Np in an irradiated Np sample is quite important, because
several tens ppm of 236Np in 237Np affect greatly the fission
property of a nuclear system including such Np.
The main objective of this study is the analysis of Pu and*Corresponding author. E-mail: osaka@oec.jnc.go.jp. FAX: +8129-266-3714.

236 Pu2.851 yαβ −(12.5%)

238Pu
87.7 yα−β (48%)
22.5− hβ(n, 2n)β +, EC (52%)

236 U
2.3 × 10 7 yα

1.5 ×10 5 y
β + , EC

240 Pu6563 yα(n, γ)

β−236mNp236 Np239 Pu(n, γ)
2.4 ×10 4 yα(n, γ)

237Np(n, γ)
2.1 ×10 6 yαβ +, EC (87.3%)

Nuclides formed by(n, 2n) reaction

238Np
2.117 dβ−(n, f)

137Cs
30.1 yβ−148NdStableNuclides formed by(n, f) reaction

Figure 1. Nuclear reaction chain started from 237Np.

Np isotopes recovered from some 237Np samples irradiated in
JOYO. In this report, the radiochemical separation, the radioanalytical procedure, and its results are briefly described andthe transmutation behavior of 2 3 7 Np in a fast reactor is
discussed on the basis of present analysis.
2. Experimental
2.1. 237Np Target for Irradiation. Figure 2 shows the positions of 237Np samples loaded in JOYO. 237Np dioxide powder
was enclosed in a small (φ = 1.5 mm, length = 8 mm, weight =70 mg) vanadium capsule and set in a central tube of irradiation rig as is shown in Figure 2. The B7 (un-instrumented fuelirradiation subassembly type-B No.7) which contained only
one sample at the axial center of the subassembly was inserted
in a core center region. The SMIR-15 (Structure Materials
Irradiation Rig No.15) that contained three 237 Np samples,
which were set in different axial position, was inserted in astainless steel reflector region. Table 1 shows the specificationand the irradiation condition of each 237 Np sample. Initial
weight of 237Np dioxide in each sample was about 0.2 mg. The
B7 was irradiated for 190.392 EFPD (Effective Full PowerDays) from September 5, 1986 to May 31, 1987. The SMIR15 was irradiated for 527.958 EFPD from May 7, 1985 to© 2003 The Japan Society of Nuclear and Radiochemical Sciences
Published on Web 06/30/2003

10

Osaka

J. Nucl. Radiochem. Sci., Vol. 4, No. 1, 2003
237 Np sample

237Np sample
242 Puspike

Reflector

Tie Rod
Compartment

30mL 8M HNO 3 +HF
100℃, 4hr

Dissolution

Control Rod(center of the rig)

Samplesolution

Fuel
Element

1mL

239 Np tracer

Dry upA

A

Valenceadjustment

WrapperA-A crossTubesection

Irradiation Rig

B7

1.5mL H2 O 2 +8MHNO 3
Np(→IV)，Pu(→IV)

Loading
solution

Fuel Subassembly

SMIR-15

Irradiation
Rig/Subassembly

Muromac 1 × 8, 200–400mesh
2mL

Anion exchange“JOYO” Core
Configuration(for B7)

Figure 2. Loaded position of 237Np sample in JOYO reactor.

Total
fluence(n/m2)

9mL 0.1M NH 4 I
51mL 12M HCl(30CV)FP, V fraction

Pu fraction

30mL 0.1M HCl(15CV)

Pu(→III)TABLE 1: Specification and Irradiation Condition of 237Np
SampleSampleAssembly No.

30mL 8M HNO3
(15CV)

Np fraction

Axial Initial weight Number of
position*of NpO2
initial237(D.F.P.B. mm)(µg)
Np atoms

Figure 3. Flow sheet for chemical separation of irradiated 237 Np
sample.

2.3. Experimental Procedure
2.3.1. Chemical Separation. In order to avoid the loss ofanalytical samples and to prepare the sample stock solution, avanadium capsule together with the irradiated 237Np oxide wasdissolved directly in an 8 M (mol/dm3) nitric acid solution byadding small quantity of hydrofluoric acid. A 0.1 mL aliquotof the sample stock solution was taken for gamma-ray spectrometry to determine 137Cs. The chemical separation procedure is shown in Figure 3, which is principally the same,
except the existence of vanadium, to that successfully applied
to the analysis of Np and Pu in irradiated MOX fuel.10 One
mL of the sample stock solution was taken for actinides
analysis. After the addition of 239Np tracer, which was milked
from 243 Am standard solution, the sample was dried up by
gentle heating. The residue was dissolved in 1.5 mL of 8 M
HNO3 solution that contained a small quantity of H2O2 for thevalence adjustment of Np and Pu to their tetravalent state andvanadium to its pentavalent state. Then, the solution was
loaded into a small polypropylene tube column filled with 2
mL of Muromac 1x8 200~400 mesh (5 cm in column length).
Np(IV) and Pu(IV) were adsorbed at the top of the column andthe major fraction of fission products including rare earth
elements and V(V) was eluted out from the column by
washing with 30 mL of 8 M HNO3. Then, the Pu fraction was
eluted out from the column by washing with 60 mL of a mixed
solution of 0.1 M NH4I and 12 M HCl. Washing with 30 mL
of 0.1 M HCl eluted out the Np fraction finally.
For the isotope dilution analysis of Pu, 242Pu was spiked into
the sample stock solution before the first dry up process of thepresent separation scheme.
In order to eliminate the interference of Pu on alpha spectrometry of Np, the separation of Pu from Np fraction was
repeated twice to get the Np fraction pure enough for alpha
spectrometry.
All chemical procedures were carried out in the airtight
glove box settled in the Alpha-Gamma Facility (AGF) of Oarai Engineering Center of JNC.

SMIR-15

U
C
L

9.3 × 1026
7.9 × 1926
3.2 × 1026

840
573
240

184
186
187

4.12 × 1017
4.16 × 1017
4.18 × 1017

B7C

5.5 × 1026

775

192.5

4.31 × 1017

*Distance from core bottom, Axial center: D.F.P.B.=775mm

February 1, 1989. Fast neutron fluence (E>0.1 MeV) for the
sample at the axial center of the SMIR-15 was reported to be
7.9 × 1026 n/m2.
2.2. Preliminary Estimation of Actinide Nuclides andFission Products Formed in Each Sample. In order to see
the actinide nuclides that were formed by neutron capture reaction and to design the analysis plan, the preliminary estimationof transformation and fission was carried out by using
ORIGEN2 code.8 Results are summarized in Table 2.
As shown in Table 2, each sample was expected to contain aconsiderable amount of 238 Pu, 239 Pu, 241 Pu, and 242 Pu. The
isotopic composition of Pu was determined by thermal ionization mass spectrometry (TIMS) and used for evaluating theneutron capture reaction rate. In order to determine the total
Pu content in an irradiated sample, isotope dilution mass spectroscopic technique was applied by using 242Pu standard spike.
For the evaluation of total fission events in an irradiatedsample, 137Cs and 148Nd monitoring method9 are well established and used widely. 148Nd method seems to be better than137
Cs method in terms of accuracy. However, the amount of148
Nd seems too small (about 80 ng at maximum) to analyze it
by TIMS. Thus, 137Cs method was selected for the estimationof total fission events.
236
Np and 236m Np may become key nuclides for futurenuclear transformation system including 237Np. As describedlater, the direct determination of 236Np was not successful forpresent analytical samples. Thus, 236Pu was determined byalpha spectrometry and the result was used for the evaluationof effective 237Np(n,2n)236mNp cross section.

TABLE 2: Number of Atoms for Each Nuclide in Irradiated 237Np Calculated Preliminarily by ORIGEN2 Code8
Assembly

SampleNo.

SMIR15U
C
L

3.60E+17 9.40E+16 5.50E+15 2.10E+14 9.50E+12 1.60E+11 3.60E+11 1.40E+10 4.50E+12 5.30E+12 1.20E+15 3.40E+14
3.80E+17 8.80E+16 4.60E+15 1.60E+14 6.40E+12 9.70E+10 2.40E+11 8.40E+09 3.90E+12 4.50E+12 1.00E+15 2.70E+14
3.70E+17 1.10E+17 5.30E+15 2.60E+14 1.70E+13 3.30E+11 6.40E+11 2.60E+10 1.50E+12 1.90E+12 4.30E+14 1.20E+14

B7C

4.60E+17 2.90E+16 5.40E+14 3.60E+12 2.00E+10 8.70E+07 1.30E+08 2.20E+06 4.70E+12 3.70E+12 8.10E+14 2.20E+14

237

Np

238

Pu

239

Pu

240

*Summation of isotopes of Am (except 241Am) and Cm

Pu

241

Pu

242

Pu

241

Am

Am,Cm*)

236

Pu

236

Np

137

Cs

148

Nd

Analysis of Irradiated 237Np in the Experimental Fast Reactor JOYO

2.3.2. Alpha and Gamma-Ray Spectrometry. A multichannel pulse height analysis system, ORTEC Model 78008B2, was used to accumulate and analyze alpha or gamma-ray
spectra. A silicon detector, ORTEC Model BA-017-200-300,
and a planar high purity germanium detector, ORTEC Model
GLP-16195/10-P, were used to take alpha and gamma-ray
spectra, respectively.
2.3.3. Mass Spectrometry. A surface ionization mass spectrometer, Finigan Mat model MAT262 equipped with photon
multiplier type detector and a six-channel Faraday cup detector
was used for isotope analysis of Pu.
3. Results and Evaluation of Transmutation Behavior
Figure 4 shows an example of alpha spectrum of an isolated
Np fraction. The alpha peaks of 237Np are well separated fromthose of Pu isotopes. Figure 5 illustrates the observation of236
Pu in an alpha spectrum of Pu fraction. Table 3 summarized
the present analytical results expressed as the number of atoms
for each nuclide in all samples. The figures in parentheses are
relative errors (%) of 1σ criteria. Attempt to observe 236Np bymass spectrometry was not successful. Thus, the data for 236Np
given in Table 3 were derived from the detection limit of mass
spectrometry.
700

Accumulation time :100,000 s
600

237

Np

Counts/channel

500
400
238

Pu

300
200
100

239

Pu

0
4.2

4.4

4.6

4.8

5.0

5.2

5.4

J. Nucl. Radiochem. Sci., Vol. 4, No. 1, 2003

11

For the evaluation of transmutation behavior, the effective
cross section data set is needed for capture and fission reactionof every actinide nuclides tabulated in Table 2. But, as isshown in Figure 2, the present irradiation assembly was quite
complicated and heterogeneous. We have only limited information about the nuclides that undergo capture and fission
reaction, namely 237Np, 238Pu, 239Pu, 240Pu, 241Pu, and 242Pu.
Thus, we selected rough proximate approach to evaluate thetransmutation behavior from the present experimental data.
Total fission events were estimated by Nf = 0.5 × NCs137/
EFY, where NCs137 is the number of 137Cs atoms and EFY iseffective fission yields. The effective fission yields are calculated by EFY = ΣN i × FY i /ΣN i , where N i is the number ofatoms for nuclide i at the end of irradiation and FY i is thefission yield of 137 mass from nuclide i. This assumption does
not introduce so much error to the evaluated total fission events,
because the fission yield values for mass 137 from 237Np, 238Pu,
239
Pu, 240Pu, 241Pu, and 242Pu do not change so significantly.
The contribution of capture reaction was evaluated from thesummation of all Pu atoms. In order to estimate the contribution of fast neutron reaction 237Np(n,2n)236mNp, it was assumed
that all the 236mNp decayed to 236Pu without any reaction loss
and the 236Pu was produced uniquely through this pass, i.e.,
the production from 236Np was ignored. These assumptions
for the production of 236Pu are appropriate because the life of
236m
Np is very short (T1/2 = 22.5 h) and that of 236Np is quite
long (T1/2 = 1500 y). Figure 6 shows the procedure for theevaluation of effective cross section of 237Np(n,2n)236mNp. The
reaction yield of 236Pu (C ) was estimated by using trial effective 1-group cross section for 237Np(n,2n)236mNp reaction. The
first trial 1-group cross section was calculated using 7-group ofenergy-dependent neutron flux and effective cross sections
used in the MAGI code developed for the operation andmanagement of JOYO.11 The estimated yield of 236Pu was

5.6

E α / MeV

Trial Effective (n,2n) cross section
Using 7-group of flux and effective
cross sections from MAGI code

Figure 4. An example of alpha-spectra of isolated Np fraction.
1000000

Accumulation time: 300,000 s
Calculation of yield of 236Pu(C )Counts/channel

10000010000

Measured yieldof 236Pu (E )

238

Pu

1000
100

Adjustment of
Effective (n,2n)cross section

236

Pu

10
1
4.4

4.6

4.8

5.0

5.2

5.4

5.6

∆|C – E | ⁄ E< 1 × 10−5

No

Yes

5.8

Yield ofEα / MeV

Figure 5. Alpha-spectrum of Pu fraction.

236

Pu and Effective (n,2n) cross section

Figure 6. Procedure for the estimation of effective 237Np(n,2n)236mNp
cross section.

TABLE 3: Analytical Results of Irradiated 237Np Sample
Assembly

Sample No.

Initial 237Np atoms

SMIR-15

U
C
L

4.12E+17
4.16E+17
4.18E+17

3.41E+17 (1.4)
3.23E+17 (1.8)
3.14E+17 (1.2)

5.99E+16 (2.8)
6.17E+16 (2.8)
9.77E+16 (2.8)

2.38E+15 (2.8)
2.57E+15 (2.8)
5.86E+15 (2.8)

6.59E+13 (2.8)
6.85E+13 (2.8)
2.48E+14 (2.8)

B7C

4.31E+17

3.85E+17 (2.2)

2.36E+16 (2.3)

3.68E+14 (2.2)

9.32E+12 (2.2)

Assembly

Sample No.

241

SMIR-15UC
L

2.70E+12 (7.1)
2.20E+12 (8.1)
3.32E+13 (2.9)

5.21E+12 (4.9)
N.D.
–
N.D.
–

7.88E+14 (5.1)
7.31E+14 (5.1)
1.60E+14 (5.2)

4.71E+11 (4.2)
4.07E+11 (5.0)
7.31E+10 (6.1)<3.99E+12<4.99E+12<4.07E+12

B7

C

2.26E+12 (2.2)

2.55E+12 (2.2)

5.04E+14 (5.1)

1.23E+12 (3.8)<2.32E+12

Pu

237

Np

242

Pu

238

Pu

137

Cs

239

Pu

236

Pu

240

Pu

236

Np

12

Osaka

J. Nucl. Radiochem. Sci., Vol. 4, No. 1, 2003

TABLE 4: Calculated Results of Relative Contribution of Each Reaction Mode
Assembly

Sample No.

Atomic balance*1)
Meas./qualified

Contribution of specified reaction mode(%)*2)
FissionCapture (n,2n)SMIR-15UC
L

1.02±0.04
0.97±0.04
1.02±0.03

3.58±0.20
3.46±0.20
0.72±0.04

15.52±0.46
16.67±0.51
25.51±0.73(3.47±0.15)E-04(3.24±0.17)E-04(4.72±0.24)E-05

19.1±0.50
20.13±0.55
26.23±0.74

B7C

0.98±0.06

2.33±0.15

6.1±0.18(6.95±0.16)E-04

8.43±0.23

Total*1) Initial 237Np atoms/(Residual 237Np atoms + total Pu atoms + total fission events)*2) Based on the number of the sum of residual 237Np atoms, total Pu atoms and total fission events

compared with the measured yield (E). The estimation ofeffective 1-group cross section was repeated by adjusting
mainly the energy dependent neutron flux until ∆C–E/E
reached to 10−5. The results of the evaluation of each reaction
mode are summarized in Table 4.
4. Discussion
The summation of residual 237Np atoms, total Pu atoms, andtotal fission events should be approximately equal to the initial
number of 237Np atoms. The ratio of the summation and thenumber of initial 237Np atoms is given also in Table 4. As isclear in Table 4, the ratios are 0.97~1.02. This fact indicatesthat the present radiochemical analysis was carried out quite
accurately and the rough approximation on the estimation forthe contribution of capture reaction does not introduce so large
error, i.e. fission of each Pu nuclide is smaller than its capture
reaction.
Total fission events were found to be 0.7~3.6% of initial
237
Np, while capture products were 6.1~25.5% of the initial
237
Np. The contribution of capture was always larger thanfission in each sample, while the ratio of fission to capture
varied according to the sample loaded position.
Figure 7 shows the dependence of relative contribution offission and capture on the axial irradiation position of SMIR15 rig. The contribution of fission decreases simply from coreto the bottom of fuel pin, while that of capture increases. This
indicates the normal tendency of the neutron spectrum softening toward the both ends of fuel pin by the effect of reflector
and structural materials. To see this effect in detail, fission tocapture ratio is plotted as a function of fast neutron ratio,
which was reported from management crew of JOYO,11 in
Figure 8. We can see clearly that the contribution of fission,
which means incineration of 237Np, increases greatly by theirradiation with hard neutron spectrum in a fast reactor.
Table 5 shows the estimated effective 237Np(n,2n)236mNp
cross section together with that of JOYO cross section
library,12 which was reported as the mean value of each fuelpin. The estimated values are much smaller than the library

TABLE 5: Effective One-Group 237Np(n,2n)236mNp CrossSection
Sample No. Effective 1-groupcross section (b) of
237
Np(n,2n)236mNp
SMIR-15 U
SMIR-15 C
SMIR-15 L
B7
C
JOYO *

Estimated atom
ratio of236
Np/237Np

6.15E+11
5.49E+11
9.22E+10
1.08E+12

1.80E-06
1.70E-06
2.93E-07
2.80E-06–* Values of “JOYO” cross-section library.12**Values estimated from effective cross section ratio of
237
Np(n,2n)236mNp/237Np(n,2n)236Np in JOYO ORIGEN2 library.12

value, as shown in Table 5. In spite of this finding, authors
have no strong evidence to recommend presently evaluatedcross section values. But, at least, this result reveals that there
is either over-estimate of 237Np(n,2n)236mNp cross section orunder-estimate of total cross section of 236Pu in the present datalibrary.
The number of atoms of long lived 236Np in Table 5 is calculated from the presently estimated effective cross section for
237
Np(n,2n)236mNp assuming that the JOYO ORIGEN2 librarydata for the ratio of the cross section for 237Np(n,2n)236Np and
237
Np(n,2n)237mNp is correct, i.e. 0.349.12 As shown in Table 5,
the content of long lived 236Np in irradiated Np is supposed to
be of ppm order. The presence of 236Np in this order does notaffect so much the fission characteristics of recovered Np.
But, for the case of under-estimate of total cross section of
236
Pu, the content of 236 Np in the recovered Np sample isexpected to be higher than 10 ppm and, thus, the macroscopic
fission cross section of recovered Np will be affected by thepresence of 236Np.
This is the first experimental report that mentions the
production of 236Np and 236Pu from 237Np in a nuclear reactor.
The authors consider that further experimental work is necessary to ensure the present finding of unexpected small production of 236 Pu on much strong experimental evidence. The
0.7

100.0

Core region

Reflector

Reflector

0.6

SMIR-15
B7

0.5
10.0

Fission to capture ratio

Relative contribution/ %4.56E-05
5.07E-05
1.89E-05
1.39E-04
3.47E-04

Estimated valuesof 236Np**(atoms)Capture reaction
Fission reaction
Total reaction
Capture reaction (calc.)
Fission reaction (calc.)

1.0

200

400

600

800

0.3
0.2
0.1

0.1
0

0.4

1000

1200

1400

Distance from core bottom / mmFigure 7. Change of relative contribution of capture and fission as afunction of axial position for SMIR-15 rig.

0.0
0.2

0.3

0.4

0.5

0.6

0.7

0.8

Fast neutron ratio

Figure 8. Change of fission to capture ratio as a function of fastneutron ratio.

Analysis of Irradiated 237Np in the Experimental Fast Reactor JOYO

isolated Np samples are stored in order to wait for the growthof 236Pu from 236Np in AGF. The result, the radiochemical
determination of 236Np in irradiated 237Np, will be reported innear future.
Acknowledgements. The authors are grateful to Mr.
Mitsuo Sudo for his technical assistance in target analysis andalso thankful to Dr. Masahiko Ito, Deputy Director of the
Irradiation Center of O-arai Engineering Center, JNC, for his
support and encouragement to this research.
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