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Benchmark experiment on copper with D-T neutrons for verification of secondary gamma-ray data in JENDL-3.1
ELSEVIER
Fusion Engineering and Design 28 (1995) 753-761
Fusion Engineering and Design
Benchmark experiment on copper with D - T neutrons for verification of secondary gamma-ray data in JENDL-3.1 F. Maekawa, Y. Oyama, C. Konno, Y. Ikeda, H. Maekawa Department of Reactor Engineering, Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken, 319-11, Japan
Abstract
Copper is a very important material for fusion devices because it is used as material for the first wall and the superconducting magnet. Nuclear heating of these components is a critical issue in the design of ITER. In order to verify secondary gamma-ray data in evaluated nuclear data libraries, a clean benchmark experiment on copper with D T neutrons was carried out at the FNS facility. Gamma-ray spectra and heating rates were measured at several points in a experimental assembly of 629 mm in diameter and 608 mm in thickness along with neutron spectra and reaction rates. The experiment was analyzed by using MCNP-4 and DOT-3.5 with cross-section libraries based on JENDL-3.1. It is pointed out from the results that the secondary gamma-ray data for threshold reactions is good. The probability of direct transitions in JENDL-3.1 from the capture state to the ground state for keV neutron capture should be increased. It is also found that energy balance of secondary gamma-ray data for the neutron capture reaction is inconsistent in an incident neutron energy range between 0.01 and 10 keV. Excess of released gamma-ray energy in JENDL-3.1 results in larger gamma-ray heating rates.
1. Introduction
Copper is extensively used in fusion devices because of its superior electrical and thermal conductivities. In a design of the International Thermonuclear Experimental Reactor (ITER) copper is used as material for the first wall and the superconducting magnet (SCM). Since the gamma-ray heating rate in the SCM and nuclear heating of the first wall are crucial in their shielding and cooling design, highly accurate data in evaluated nuclear data libraries are required for neutron transport and secondary gamma-ray production data. Hence the overall accuracy of the secondary
gamma-ray data should be verified by calculation on benchmark experiments performed under D - T neutron fields. Two integral experiments for copper with D - T neutrons have been performed so far for the purpose of verification of secondary gamma-ray data. One of the experiments [ 1,2] was conducted as a part of the pulsed sphere program at LLNL. In the experiment, a pulse height spectrum of gamma-rays leaking from a copper sphere of 80 mm in diameter were measured with an NE213 liquid organic scintillator. Another experiment [3,4] was performed at OKTAVIAN in Osaka University as one of the series of experiments with pulsed
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spheres. Gamma-ray spectrum leaking from a copper spherical shell of 275 mm in thickness and 6.23 g cm-3 in apparent density was measured with a NaI(T1) scintillation detector. Since a combination of pulsed neutrons and the time-of-flight method was adopted to separate gamma-rays from neutrons in these experiments, the spectra were not measured in the steady state but with time cutoffs. Therefore gamma-rays produced by threshold reactions mainly with D - T neutrons were clearly seen in the measured spectra. However, only a small portion of the gamma-rays produced by radiative capture reactions, mostly with slow neutrons, were contained in the spectra. In order to verify cross-section data in evaluated nuclear data libraries, a clean benchmark experiment on copper with D - T neutrons was carried out at Fusion Neutronics Source, FNS [5]. Since gamma-rays were measured in the steady state, gamma-rays both from threshold reactions and radiative capture reactions were measured in the present experiment. In this paper, the experiment with gamma-rays and verification of secondary gamma-ray data are the focus. The experiment and verification for neutrons are presented in a separate paper [6].
2. Experimental details 2.1. General description o f the experiment The intense D - T neutron source FNS in JAERI was utilized for the experiment. The experimental assembly was constructed in a quasi-cylindrical shape by piling up oxygen-free copper blocks more than 99.99% purity. The dimension of the blocks were 50.7 x 50.7 mm 2 square with length of 203.1, 101.5 or 50.7 mm. The equivalent diameter and thickness of the assembly were 629 and 608 mm, respectively. The assembly was fastened into aluminum frames, and mounted on a deck made of steel. It was placed at 200 mm from the D - T source position locating its central axis of the cylinder on the extended line of deuteron beam. In order to insert several detectors into the assembly, four experimental channels were equipped at depths of 76, 228, 380 and 532 mm from the front surface of the assembly. The experimental channels for insertion of detectors were set from the side of the assembly to prevent radiation streaming through a cavity made by the detector itself. All channels were made by thin sheaths and drawers of 0.2 mm thick stainless steel. Copper blocks of 49.2 mm 2 square, slightly smaller than the
standard size blocks, were loaded in these experimental channels when the channels were not used. A positive deuteron beam of 350 keV energy was bombarded to a tritium-titanium metal target which contained tritiums of about 3.7 × 1011 Bq to generate D - T neutrons. The beam current was controlled between 20 nA and 2 mA according to the measurement conditions. A typical neutron yield was about 1.5 x 1014 neutrons C-1 for whole solid angle. The absolute neutron yield was determined with two systems of associated alpha-particle monitors [7,8] with accuracy of _+2%. Detailed explanations and numerical values of the measured data are presented in Ref. [9].
2.2. Gamma-ray spectrum measurement 2.2.1. Measurement procedure A liquid organic scintillation counter NE213 [10] was used for the measurement. The NE213 scintillator had the ability to separate of gamma-rays from neutrons, and it emitted hardly any gamma-rays itself because of its small gamma-ray emission cross-sections. A sectional view of the counter is shown in Fig. 1. The scintillator was contained in a spherical boric-silicic glass cell of 40 mm in diameter and had an isotropic sensitivity. Light pulses of constant intensity were injected into the photo-multiplier tube through a optical fiber in order to stabilize gain drift due to change of counting rate and temperature. In order to expand dynamic ranges of the spectra, two main amplifiers were employed with ten times different gains. Gammaray events were separated from neutron signals by using rise time to pulse height convertors. Prompt gamma-ray spectra were measured in the four experimental channels. The center of the NE213 scintillator was set to be on the central axis of the assembly. The accelerator was operated in an arc-pulse mode with pulse width of 0.75 ms and repetition rate of 1.98 ms. Deuteron beam current was adjusted to between 20 nA and 10 gA depending on the measurement positions to keep the counting rates constant about 2000 cps. Energy calibration of measured pulse height spectra was performed with the Compton edge ofl.275 MeV gamma-rays of sodium 22. Signals from a high precision research pulser were fed into the pre-amplifier to determine zero pulse height. 2.2.2. Rejection o f decay gamma-rays For the measurement of prompt gamma-rays, it is important to clearly reject decay gamma-rays which are regarded as the background. In a case of copper in D T neutron fields, decay gamma-rays of 0.511 MeV
F. Maekawa et al. I Fusion Engineering and Design 28 (1995) 753-761
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froln 62Cu generated by the 63Cu(n,2n) reactions are dominant because of its large cross-section of 570 mb, high gamma-ray emission probability of 195.6% and short half life of 9.73 rain. As shown in Fig. 2(a), the intensity of decay gamma-rays changes continuously depending on the history of neutron irradiation for the experimental assembly. If the decay gamma-rays are measured before or after the measurement of prompt gamma-rays, it is very difficult to determine the averaged amount of decay gamma-rays during the measurement of prompt gamma-rays. Hence the pulsed neutron method is applied to accurately subtract the background. The period between the time when D - T neutrons are injected into the experimental assembly and when neutrons disappear from the assembly due to captures and leakages, is almost several hundred microseconds, which corresponds to the slowing down time of D - T neutrons to thermal neutrons. Prompt g a m m a - r a y s
following neutron reactions are emitted several hundreds microseconds after injection of D - T neutrons. Pulsed neutrons of several hundreds microseconds in width are generated and the periods of measurement for foreground and background gamma-rays are determined as shown in Fig. 2(b). The intensity of decay gamma-rays emitted by activated nuclei, whose halflives are usually longer than one second, can be regarded as constant during the measurement period of foreground and background. Pulse height spectra for the foreground and background periods can be measured in the same measurement with different time-gate. A pulse height spectrum without decay gamma-rays is accurately derived by subtracting the background from the foreground spectrum. 2.2.3. Data reduction After subtraction of the decay gamma-ray background, measured pulse height spectra were unfolded
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F. Maekawa et al. / Fusion Engineering and Design 28 (1995) 753-761
using the FORIST code [11] to derive energy spectra higher than 0.25 MeV. The response matrix used in the unfolding process was calculated by using the M A R T H A code [12] after modification for the NE213 scintillator. The boric-silicic glass of the container of the NE213 scintillator contains boron-10 atoms. The organic scintillator naturally involves hydrogen-1 atoms. The boron-10 and hydrogen-1 atoms emit parasitic gammarays following the two reactions. l°B(n,c07Li*; 7Li* ~ 7 L i + 7(0.478 1H(n,y)2H*;
2H*--+2H +~/(2.225
MeV) MeV)
These parasitic gamma-rays were also included in the measured spectra, because they could not be separated. The main error sources additional to the parasitic gamma-ray emission described above were estimated as follows: response functions ( + 0-8%), perturbation of radiation field by the NE213 counter ( + 0 - 1 0 % ) and statistics error ( _ 4-10%). As a whole, the measured spectra are observed roughly 0-15% larger than the true spectra. 2.3. Measurement o f gamma-ray heating rate 2.3.1. Measurement procedure In order to measure the gamma-ray heating rate, a method which utilized plural kinds of thermoluminescence dosimeter (TLD) was proposed by Tanaka et al. [ 13]. Some examples of the application of the method in D - T neutron fields are described in Refs. [14,15]. Here, brief explanations are given. Gamma-ray heating rate in a medium measured by plural kinds of TLD monotonously increase as a function of effective atomic numbers of the TLDs. Hence the gamma-ray heating rate of the medium can be derived by interpolating those measured by several kinds of TLD. Since the atomic number of copper was 29, three kinds of TLD which effective atomic numbers (Zeer) were around 29 were selected; MSO (Mg2SiO4, Zeer=ll.1), SSO (Sr2SiO 4, Zefr=32.5 ) and BSO (Ba2SiO4, Zeer= 49.9). All TLDs were powder and are sealed in glass capsules of 2 mm in diameter and 12 mm in length. All TLDs were calibrated at a cobalt 60 standard gamma-ray field. Four samples of each kind of TLDs were packed in a polyethylene bag and set in the four experimental channels and on the front and the rear surfaces of the assembly. The TLDs were located at - 1, 58, 2t0, 362, 514 and 609 mm from the front surface of the assembly.
The irradiation was carried out for 16 minutes with deuteron beam current of 1.5 mA, and the total neutron yield was 2.31 x 1014. About 15 rain after the irradiation, TLDs were taken out from the experimental assembly and kept in a dark place. One week later from the irradiation, thermoluminescences (TL) were read out by a TLD reader (KYOKKO 2500). 2.3.2. Data reduction The background TL was subtracted, and the obtained TL was converted to the unit of exposure dose of 6°Co equivalence. Since TLDs were sensitive to neutrons as well as gamma-rays, neutron contribution on each TLD was subtracted from the total TL response. The neutron contribution was calculated by integrating the products of neutron response functions and neutron spectra. The neutron response function of each TLD were calculated with a code [16] developed by Hashikura et al. The neutron spectra were calculated with MCNP-4 [17] as described in the next section. Subsequently, since the atomic number of copper was 29, gamma-ray heating rate of copper was derived by interpolation of absorbed dose for MSO and SSO. In the obtained gamma-ray heating rates, some parasitic contributions were included; (i) gamma-rays generated at the target; (ii) beta- and gamma-rays associated with decay of 62Cu. These contributions were estimated based on neutron-photon-electron coupled transport calculations with MCNP-4 and the measured data were corrected for the contributions. The maximum correction was 25.9% at the front surface of the experimental assembly, and the corrections were less than 1% and could be negligible at positions more than 362 mm in depth. The overall errors ranged between 11 and 25% depending on measurement points.
3. Calculation procedure Calculations on the experiment was conducted with two kinds of transport code. One was the continuous energy Monte Carlo code MCNP-4 with the cross-section library FSXLIB-J3 [18,19] based on JENDL-3.1 [20]. Another was the two-dimensional discrete ordinate code DOT-3.5 [21] with P5-S16 approximation. The G R T U N C L code [21] was used to calculate the first collision sources. Two cross-section libraries, the JSSTDL [22,23] and FUSION-J3 [24], were employed for the DOT calculations. Both libraries were based on JENDL-3.1 and had the same group structure of 125 and 40 groups for neutron and gamma-ray, respectively, but the self shielding correction factors were
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F. Maekawa et al. / Fusion Engineering and Design 28 (1995) 753- 761
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Fig. 3. Gamma-ray spectra in the copper assembly at four positions measured and calculated by MCNP. considered only in JSSTDL. Source neutron spectrum calculated by the MORSE-DD code [25] simulating the real target structure as precise as possible was used for the source. Statistical errors for the MCNP calculations were reduced by adopting the source direction biasing method and the geometry splitting with the Russian roulette method changing cell importance double at every 150 mm of depth. The calculated gamma-ray spectra by M C N P and DOT were smeared with the energy resolution functions of the measured spectra. Gamma-ray heating rates were calculated by energy-integrating products of gamma-ray spectra and kerma factors which were derived from DLC-99 [26]. Gamma-ray production reactions of natural copper in JENDL-3.1 are gathered into two components; the non-elastic reaction (MT3) and the neutron capture reaction (MT102). Gamma-ray production reaction be-
low 0.2 MeV is presented in the MT102 while all the gamma-ray production reactions above 0.2 MeV are presented in the MT3 including the neutron capture reaction. The MCNP-4 code was modified so as to classify the calculated gamma-ray spectra into the two components, MT3 and MT102, according to the gamma-ray production reactions. The modification made it possible to easily identify inadequate component of spectra and reactions sensitive to the spectra.
4. Results and discussions
4.1. Gamma-ray spectrum Measured gamma-ray spectra at the four positions are shown in Fig. 3 with those calculated by MCNP. The MT3 component is dominant at 76 mm, and it
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F. Maekawa et al.
/Fusion Engineering and Design 28 (1995) 753-761
implies that most of gamma-rays are generated by threshold reactions with 14 MeV neutrons. The MT3 component decreases relatively to the MT102 component as increase of depth. At the deepest measurement point of 532 mm, the spectrum is dominated by the neutron capture reactions because of the remarkable MT102 component. Hence secondary gamma-ray data for non-threshold reactions, in addition to threshold reactions, can be tested by the measurements at different positions. This is one of the advantages of the present experiment because gamma-rays from non-threshold reactions have not been measured so far in the previous experiments. As for gamma-rays produced by threshold reactions mainly with 14 MeV neutrons, secondary gamma-ray data is usually verified only by comparing calculated gamma-rays with experiments because neutron transport calculations of copper is highly accurate at 14 MeV as proved in Ref. [6]. However, the accuracy of calculation for low energy neutrons, especially in a resonance region, is not always satisfactory. Since gamma-rays from non-threshold reactions are mainly produced by multiply scattered low energy neutrons, both neutron and gamma-ray should be examined to verify the secondary gamma-ray data. A s for gamma-rays from neutron capture reactions, intensity of gamma-rays is determined by the calculated neutron flux, while their spectral shapes are not so sensitive to those. t n Fig. 3, the calculated spectra in the energy range below 6 MeV at all measurement points agree with the experiment within about 20 %, except at around 0.5 and 2.2 MeV because of the experimental problem. In the energy range above 6 MeV, the calculated spectra are smaller than the experiment. Peaks around 7 - 8 MeV observed: in the experiment are not obtained by the calculations especially at 76, 228 and 380 mm. The peaks around 7 - 8 MeV are arisen from neutron capture reactions. Because Q-values of the capture reactions are 7.07 and 7.92 MeV for copper 63 and copper 65, respectively, and gamma-ray energies are equal to the Q-values when a nucleus on the capture state directly transits to the ground state. The neutron capture peaks are more obvious in the deeper positions where fractions of low energy neutrons to 14 MeV neutrons are more. Fig. 4 shows secondary gamma-ray emission probabilities (SGEPs) for the neutron capture reaction taken from JENDL-3.1. The SGEPs are based on the existing measurements for thermal energy and on the calculation for incident neutron energies of 10, 100 and 200 keV. A remarkable peak at 7 - 8 MeV is seen only for thermal energy. Such a big differences in SGEP between thermal energy and keV energies might not be reasonable.
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Fig. 4. Secondary gamma-ray emission probabilities for the neutron capture reaction taken from JENDL-3.1. Comparisons of the calculated spectra with the measured ones and treatment of the SGEPs in JENDL:3.1 lead us to the following interpretation. For the SGEPs in JENDL-3.1 in the incident neutron energies of 10, 100 and 200 keV, direct transition probability from the capture state to the ground state is too small and the gamma-ray spectra are too soft. The peaks associated with the direct transitions are smaller in the calculation and agreement between the calculation and the experiment at the peak portion is worse. In deeper positions, since the energy of neutrons which cause capture reactions becomes lower, the SGEP for thermal neutron incidence comes to be adopted more frequently. Hence the disagreement at the peak is improved in the deeper measurement points. 2.0
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Fig. 5. Calculated to experimental values of gamma-ray heating rate of copper along the central axis of the assembly.
F. Maekawa et al. / Fusion Engineering and Design 28 (1995) 753 76I 4.2. Gamma-ray heating rate Fig. 5 shows calculated to experimental values (C/Es) of gamma-ray heating rate of copper. The C/Es of MCNP and DOT + JSSTDL are very close one another. The figure illustrates that the gamma-ray heating rates by the two calculations agree with the experiment almost within the experimental error ranges of 11 25%. On the other hand, gamma-ray heating rates by the DOT ÷ FUSION-J3 calculation are much larger than the experiment at positions deeper than 200 mm; C/Es are about 1.8 at 210 and 362 mm and 1.3 at 514 mm. This fact can be easily explained. If the self-shielding correction of cross-section is omitted as the DOT + FUSION-J3 calculation, calculated gamma-ray heating rates involve large ambiguity. If the self-shielding correction is taken into account, gamma-ray heating rate calculations by DOT with group constants have almost the same accuracy with continuous energy calculations even for gamma-rays from resonance capture reactions. At the front surface and 58 mm, C/Es for three calculations are almost the same. This fact implies that gamma-rays near the D - T source are mostly generated by threshold reactions with D - T neutrons. In the measurement positions deeper than 200 mm, differences of C/Es between two DOT calculations with and without the self-shielding correction are large. These differences prove t h a t gamma-rays in the positions are mostly produced by neutron capture reactions of which selfshielding correction is important. 4.3: Energy balance o f (n,7) reaction The energy sum of gamma-rays emitted from (n,7) reactions should be equal to the Q-value of the reaction when the energies of the incident neutron and recoil atom are negligible. This relation in nuclear data files can be tested by the measurement of gamma-ray heating rate and (n,7) reaction rate. The gamma-ray heating rate, i.e. absorbed energy in a media, is measured by the TLD technique. The heating rate is almost proportion to the energy sum of gamma-rays at the measurement point, so that the sum of released energy from the (n,?) reaction is estimated from the gamma-ray heating rate. On the other hand, the Cu(n,7) reaction rate at the deepest measurement position of 514 mm is estimated as follows. Since a fraction of the 65Cu(n,2n) reaction to the nateu(n, x)64Cu reaction is about 1% according to the MCNP calculation, the measured reaction rate, natCu(n,x)64Cu, a t the position is almost equal to
759
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..................................L...........".......................L.....................
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Neutron Energy [MeV]
Fig. 6. A plot to check the energy balance for the neutron capture reaction in JENDL-3.1. 63Cu(n,7) reaction rate as follows: natCu(n,x)64Cu
= 63Cu(n,~)
-}- 6 S C u ( n , 2 n )
~ 63Cu(n,7)
(1) At any position along the central axis of the experimental assembly through the front to the rear surface, the following equation stands up according to the MCNP calculation: 63Cu(n,'~) natCu(n,.~)
63Cu(n,'y)
63Cu(n,7) ÷ 65Cu(n,?)
= 0.79 _+ 0.02
(2)
This means that the ratio of the Eq. (2) is almost independent from changes of neutron spectrum. Considering the Eqs. (1) and (2), the natCu(n,?') reaction rate at the 514 mm is in proportion to the measured natCu(n,x)64Cu reaction rate: n a t C u ( n 7) ~ n a t C u ( n , x ) 6 4 C u / 0 - 7 9
off n a t e u ( n , x ) 6 4 C u .
(3)
At the 514 mm position, C/Es of gamma-ray heating rate and Cu(n, x)64Cu reaction rate are 1.02 and 0.84, respectively, for the MCNP calculation. Experimental errors for them are 11 and 4%, and statistical errors of the MCNP calculations are 2 and 3%, respectively. The calculated gamma-ray heating rate is in good agreement with the experiment. However this agreement does not mean that the secondary gamma-ray data in JENDL-3.1 are valid, because the calculated Cu(n,y) reaction rate, which is the source reaction for the gamma-ray production, is meaningfully smaller than the experiment. In other words, the underestimation of gamma-ray production reaction and the overestimation of released gammaray energies are cancelled each other so as to give the good agreement in the gamma-ray heating rate.
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F. Maekawa et al. / Fusion Engineering and Design 28 (1995) 753 761
According to the discussions above, it is supposed that the secondary gamma-ray data of JENDL-3.1 is slightly assigned by excess energy of gamma-rays for the neutron capture reaction. This excess energy release can be attributed to the secondary gamma-ray data in JENDL-3.1. Fig. 6 is a plot to check the energy conservation for the neutron capture reaction. In the figure, total gamma-ray energy released by one neutron capture reaction is presented. The thick line is composed from JENDL-3.1 and the two thin lines are calculated based on the law of conservation of energy for 63Cu and 65Cu. Obviously the thick line must be in a range between the two thin lines. However, the released energy of JENDL-3.1 is out of the range around incident neutron energy of 1 keV. An energy of 12 MeV, about 1.5 times higher values to the expected, is released at 1 keV. Consequently it is indicated that the energy balance of gamma-ray emission in JENDL-3.1 is not consistent.
5. Concluding remarks G a m m a - r a y spectra and heating rates were measured in a copper slab assembly bombarded by D - T neutrons. The pulsed neutron method was adopted to reject decay gamma-rays correctly. The experiment presented the first benchmark experiment in which gamma-rays from not only threshold but also non-threshold reactions were measured. Through the experimental analyses, it was found for secondary gamma-ray data of copper in JENDL-3.1 that the gamma-ray data associated with threshold reactions were valid while modifications of spectral shapes arid energy balance were recommended Direct transition probabilities from the capture~excited state to the~ ground state for .non-thermal ne.utron energy should be increased for JENDL-3.1 reev~!uation. F o r applying grollp con~t.ant~ to calculations, self-shielding correction was inevitable for gamma-rays produced by neutron capture reactions in resonance region.
References [1] L.F. Hansen, C. Wong, T.T. Komoto, B.A. Rohl and R.J. Howerton, Nucl. Technol. 51 (1973) 278. [2] E. Goldberg, L.F. Hansen, T.T. Komoto, B.A. Pohl, R.J. Howerton, R.E. Dye, E.F. Plechaty and W.E. Warren, Neutron and gamma-ray spectra from a variety of materials bombarded with 14-MeV neutrons, Nucl. Sci. Eng. 105 (1990) 319.
[3] J. Yamamoto, T. Kanaoka, I. Murata, A. Takahashi and K. Sumita, Gamma-ray emission spectra from spheres with 14 MeV neutron source, Proc. 1988 Seminar on Nucl. Data, JAERI-M 89-026, 1989, p. 232. [4] Sub Working Group of Fusion Reactor Physics Subcommittee (ed.), Collection of Experimental Data for Fusion Neutronics Benchmark, JAERI-M 94-014, 1994, p. 32. [5] T. Nakamura, H. Maekawa, Y. Ikeda and Y. Oyama, A DT neutron source for fusion neutronics experiments at the JAERI, Proc. Int. Ion Eng. Congress - - ISIAT '83/IAPT '83, Kyoto, Japan, Vol. 1, 1983, p. 567. [6] C. Konno, F. Maekawa, Y. Oyama, Y. Ikada and H. Maekawa, Benchmark experiment on copper with D - T neutrons for verification of neutron transport codes and the related nuclear data of JENDL-3.1, Fusion Eng. Des. 28 (1995). [7] H. Maekawa, Y. Ikeda, Y. Oyama, S. Yamaguchi and T. Nakamura, Neutron yield monitors for the Fusion Neutronics Source (FNS) for 80°beam line, JAERI-M 83-219, 1983. [8] S. Yamaguchi, Y. Oyama and H. Maekawa, Calculation of Anisotropy Correction Factor for Determination of D - T Neutron Yield by Associated s-particle Method, JAERI-M 84-109, 1984. [9] F. Maekawa, Y. Oyama, C. Konno, Y. Ikeda, K. Kosako and H. Maekawa, Benchmark Experiment on a Copper Slab Assembly Bombarded by D-T Neutrons, JAERI-M 94-038, 1994. [10] F. Maekawa and Y. Oyama, Characteristics of a 40 mm Diameter NE213 Scintillation Counter for in-system Gamma-ray Spectrum Measurement, JAERI-M 91-138, 1991, p. 97. [11] FORIST Spectrum Unfolding Code, Radiation Shielding Information Center, Oak Ridge National Laboratory, PSR-92, 1975. [12] K. Saito and S. Moriuchi, Monte Carlo calculation of accurate response functions for a NaI(TI) detector for gamma rays, Nucl. Instrum. Meth. 185 (1981) 299. [13] S. Tanaka and N. Sasamoto, Gamma-ray absorbed dose measurements in media with plural thermoluminescent dosimeters having different atomic numbers, J. Nucl. Sci. Technol. 22 (1985) 109. [14] S. Yamaguchi, H. Maekawa, K. Kosako, T. Nakamura and K.G. Porges, Measurements of gamma-ray heating in lithiumoxide, graphite and iron slab assemblies bombarded by D - T neutrons, Fusion Eng. Des. 10 (1989) 163. [15] Y. Oyama, Experiments of Nuclear Heating by Gammarays at FNS, JAERI-M 91-062, 1991, p. 106. [16] H. Hashikura, K. Haikawa, S. Tanaka and S. Kondo, Calculation of neutron response of thermoluminescent dosimeters, J. Faculty Eng., University of Tokyo 39 (1987) 7. [17] J.F. Briesmeister (ed.), MCNP: A General Monte Carlo Code for Neutron and Photon Transport, Version 4, RSIC/CCC-200, 1991.
F. Maekawa et al. /Fusion Engineering and Design 28 (1995) 753-761 [18] K. Kosako, Y. Oyama and H. Maekawa, FSXLIBJ3: MCNP Continuous Energy Cross-Section Library Based on JENDL-3, JAERI-M 91-187, 1991, in Japanese. [19] K. Kosako, Y. Oyama and H. Maekawa, MCNP continuous cross-section library based on JENDL-3, Proc. Topical Meeting on New Horizons in Radiation Protection and Shielding, Pasco, WA, 1992, p. 357. [20] K. Shibata, T. Nakagawa, T. Asami, T. Fukahori, T. Narita, S. Chiba, M. Mizumoto, A. Hasegawa, Y. Kikuchi, Y. Nakajima and Sin-iti Igarashi, Japanese Evaluated Nuclear Data Library, Version-3, JENDL-3, JAERI 1319, 1990. [21] W.A. Rhoades and F.R. Mynatt, The DOT III Two-dimensional Discrete Ordinates Transport Code, ORNL/ TM-4280, 1973. [22] A. Hasegawa, JSSTDL-295n-104g: a common nuclear group cross-section library based on JENDL-3 nuclear
[23]
[24]
[25]
[26]
761
data file, Proc. Second Specialists' Meeting on Nuclear Data for Fusion Reactor, JAERI-M 91-062, 1991, p. 15. A. Hasegawa, Development of a common nuclear group constants library system: JSSTDL-295n-104g based on JENDL-3 nuclear data library, Proc. Int. Conf. on Nucl. Data, Jfilich, Germany, 1991, p. 232. K. Maki, K. Kosako, Y. Seki and H. Kawasaki, Nuclear Group Constant Set FUSION-J3 for Fusion Reactor Nuclear Calculations Based on JENDL-3, JAERI-M 91072, 1991, in Japanese. M. Nakagawa and T. Mori, MORSE-DD: A Monte Carlo Code Using Multi-Group Double Differential Form Cross-section, JAERI-M 84-126, Japan Atomic Energy Research Institute, 1984. R.W. Roussin, J.R. Knight, J.H. Hubbell and R.J. Howerton, Description of the DLC-99/HUGO Package of Photon Interaction Data in ENDF/B-V format, ORNL/ RSIC-46, ENDF-335.
Fusion Engineering and Design 28 (1995) 753-761
Fusion Engineering and Design
Benchmark experiment on copper with D - T neutrons for verification of secondary gamma-ray data in JENDL-3.1 F. Maekawa, Y. Oyama, C. Konno, Y. Ikeda, H. Maekawa Department of Reactor Engineering, Japan Atomic Energy Research Institute, Tokai-mura, Naka-gun, Ibaraki-ken, 319-11, Japan
Abstract
Copper is a very important material for fusion devices because it is used as material for the first wall and the superconducting magnet. Nuclear heating of these components is a critical issue in the design of ITER. In order to verify secondary gamma-ray data in evaluated nuclear data libraries, a clean benchmark experiment on copper with D T neutrons was carried out at the FNS facility. Gamma-ray spectra and heating rates were measured at several points in a experimental assembly of 629 mm in diameter and 608 mm in thickness along with neutron spectra and reaction rates. The experiment was analyzed by using MCNP-4 and DOT-3.5 with cross-section libraries based on JENDL-3.1. It is pointed out from the results that the secondary gamma-ray data for threshold reactions is good. The probability of direct transitions in JENDL-3.1 from the capture state to the ground state for keV neutron capture should be increased. It is also found that energy balance of secondary gamma-ray data for the neutron capture reaction is inconsistent in an incident neutron energy range between 0.01 and 10 keV. Excess of released gamma-ray energy in JENDL-3.1 results in larger gamma-ray heating rates.
1. Introduction
Copper is extensively used in fusion devices because of its superior electrical and thermal conductivities. In a design of the International Thermonuclear Experimental Reactor (ITER) copper is used as material for the first wall and the superconducting magnet (SCM). Since the gamma-ray heating rate in the SCM and nuclear heating of the first wall are crucial in their shielding and cooling design, highly accurate data in evaluated nuclear data libraries are required for neutron transport and secondary gamma-ray production data. Hence the overall accuracy of the secondary
gamma-ray data should be verified by calculation on benchmark experiments performed under D - T neutron fields. Two integral experiments for copper with D - T neutrons have been performed so far for the purpose of verification of secondary gamma-ray data. One of the experiments [ 1,2] was conducted as a part of the pulsed sphere program at LLNL. In the experiment, a pulse height spectrum of gamma-rays leaking from a copper sphere of 80 mm in diameter were measured with an NE213 liquid organic scintillator. Another experiment [3,4] was performed at OKTAVIAN in Osaka University as one of the series of experiments with pulsed
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spheres. Gamma-ray spectrum leaking from a copper spherical shell of 275 mm in thickness and 6.23 g cm-3 in apparent density was measured with a NaI(T1) scintillation detector. Since a combination of pulsed neutrons and the time-of-flight method was adopted to separate gamma-rays from neutrons in these experiments, the spectra were not measured in the steady state but with time cutoffs. Therefore gamma-rays produced by threshold reactions mainly with D - T neutrons were clearly seen in the measured spectra. However, only a small portion of the gamma-rays produced by radiative capture reactions, mostly with slow neutrons, were contained in the spectra. In order to verify cross-section data in evaluated nuclear data libraries, a clean benchmark experiment on copper with D - T neutrons was carried out at Fusion Neutronics Source, FNS [5]. Since gamma-rays were measured in the steady state, gamma-rays both from threshold reactions and radiative capture reactions were measured in the present experiment. In this paper, the experiment with gamma-rays and verification of secondary gamma-ray data are the focus. The experiment and verification for neutrons are presented in a separate paper [6].
2. Experimental details 2.1. General description o f the experiment The intense D - T neutron source FNS in JAERI was utilized for the experiment. The experimental assembly was constructed in a quasi-cylindrical shape by piling up oxygen-free copper blocks more than 99.99% purity. The dimension of the blocks were 50.7 x 50.7 mm 2 square with length of 203.1, 101.5 or 50.7 mm. The equivalent diameter and thickness of the assembly were 629 and 608 mm, respectively. The assembly was fastened into aluminum frames, and mounted on a deck made of steel. It was placed at 200 mm from the D - T source position locating its central axis of the cylinder on the extended line of deuteron beam. In order to insert several detectors into the assembly, four experimental channels were equipped at depths of 76, 228, 380 and 532 mm from the front surface of the assembly. The experimental channels for insertion of detectors were set from the side of the assembly to prevent radiation streaming through a cavity made by the detector itself. All channels were made by thin sheaths and drawers of 0.2 mm thick stainless steel. Copper blocks of 49.2 mm 2 square, slightly smaller than the
standard size blocks, were loaded in these experimental channels when the channels were not used. A positive deuteron beam of 350 keV energy was bombarded to a tritium-titanium metal target which contained tritiums of about 3.7 × 1011 Bq to generate D - T neutrons. The beam current was controlled between 20 nA and 2 mA according to the measurement conditions. A typical neutron yield was about 1.5 x 1014 neutrons C-1 for whole solid angle. The absolute neutron yield was determined with two systems of associated alpha-particle monitors [7,8] with accuracy of _+2%. Detailed explanations and numerical values of the measured data are presented in Ref. [9].
2.2. Gamma-ray spectrum measurement 2.2.1. Measurement procedure A liquid organic scintillation counter NE213 [10] was used for the measurement. The NE213 scintillator had the ability to separate of gamma-rays from neutrons, and it emitted hardly any gamma-rays itself because of its small gamma-ray emission cross-sections. A sectional view of the counter is shown in Fig. 1. The scintillator was contained in a spherical boric-silicic glass cell of 40 mm in diameter and had an isotropic sensitivity. Light pulses of constant intensity were injected into the photo-multiplier tube through a optical fiber in order to stabilize gain drift due to change of counting rate and temperature. In order to expand dynamic ranges of the spectra, two main amplifiers were employed with ten times different gains. Gammaray events were separated from neutron signals by using rise time to pulse height convertors. Prompt gamma-ray spectra were measured in the four experimental channels. The center of the NE213 scintillator was set to be on the central axis of the assembly. The accelerator was operated in an arc-pulse mode with pulse width of 0.75 ms and repetition rate of 1.98 ms. Deuteron beam current was adjusted to between 20 nA and 10 gA depending on the measurement positions to keep the counting rates constant about 2000 cps. Energy calibration of measured pulse height spectra was performed with the Compton edge ofl.275 MeV gamma-rays of sodium 22. Signals from a high precision research pulser were fed into the pre-amplifier to determine zero pulse height. 2.2.2. Rejection o f decay gamma-rays For the measurement of prompt gamma-rays, it is important to clearly reject decay gamma-rays which are regarded as the background. In a case of copper in D T neutron fields, decay gamma-rays of 0.511 MeV
F. Maekawa et al. I Fusion Engineering and Design 28 (1995) 753-761
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froln 62Cu generated by the 63Cu(n,2n) reactions are dominant because of its large cross-section of 570 mb, high gamma-ray emission probability of 195.6% and short half life of 9.73 rain. As shown in Fig. 2(a), the intensity of decay gamma-rays changes continuously depending on the history of neutron irradiation for the experimental assembly. If the decay gamma-rays are measured before or after the measurement of prompt gamma-rays, it is very difficult to determine the averaged amount of decay gamma-rays during the measurement of prompt gamma-rays. Hence the pulsed neutron method is applied to accurately subtract the background. The period between the time when D - T neutrons are injected into the experimental assembly and when neutrons disappear from the assembly due to captures and leakages, is almost several hundred microseconds, which corresponds to the slowing down time of D - T neutrons to thermal neutrons. Prompt g a m m a - r a y s
following neutron reactions are emitted several hundreds microseconds after injection of D - T neutrons. Pulsed neutrons of several hundreds microseconds in width are generated and the periods of measurement for foreground and background gamma-rays are determined as shown in Fig. 2(b). The intensity of decay gamma-rays emitted by activated nuclei, whose halflives are usually longer than one second, can be regarded as constant during the measurement period of foreground and background. Pulse height spectra for the foreground and background periods can be measured in the same measurement with different time-gate. A pulse height spectrum without decay gamma-rays is accurately derived by subtracting the background from the foreground spectrum. 2.2.3. Data reduction After subtraction of the decay gamma-ray background, measured pulse height spectra were unfolded
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F. Maekawa et al. / Fusion Engineering and Design 28 (1995) 753-761
using the FORIST code [11] to derive energy spectra higher than 0.25 MeV. The response matrix used in the unfolding process was calculated by using the M A R T H A code [12] after modification for the NE213 scintillator. The boric-silicic glass of the container of the NE213 scintillator contains boron-10 atoms. The organic scintillator naturally involves hydrogen-1 atoms. The boron-10 and hydrogen-1 atoms emit parasitic gammarays following the two reactions. l°B(n,c07Li*; 7Li* ~ 7 L i + 7(0.478 1H(n,y)2H*;
2H*--+2H +~/(2.225
MeV) MeV)
These parasitic gamma-rays were also included in the measured spectra, because they could not be separated. The main error sources additional to the parasitic gamma-ray emission described above were estimated as follows: response functions ( + 0-8%), perturbation of radiation field by the NE213 counter ( + 0 - 1 0 % ) and statistics error ( _ 4-10%). As a whole, the measured spectra are observed roughly 0-15% larger than the true spectra. 2.3. Measurement o f gamma-ray heating rate 2.3.1. Measurement procedure In order to measure the gamma-ray heating rate, a method which utilized plural kinds of thermoluminescence dosimeter (TLD) was proposed by Tanaka et al. [ 13]. Some examples of the application of the method in D - T neutron fields are described in Refs. [14,15]. Here, brief explanations are given. Gamma-ray heating rate in a medium measured by plural kinds of TLD monotonously increase as a function of effective atomic numbers of the TLDs. Hence the gamma-ray heating rate of the medium can be derived by interpolating those measured by several kinds of TLD. Since the atomic number of copper was 29, three kinds of TLD which effective atomic numbers (Zeer) were around 29 were selected; MSO (Mg2SiO4, Zeer=ll.1), SSO (Sr2SiO 4, Zefr=32.5 ) and BSO (Ba2SiO4, Zeer= 49.9). All TLDs were powder and are sealed in glass capsules of 2 mm in diameter and 12 mm in length. All TLDs were calibrated at a cobalt 60 standard gamma-ray field. Four samples of each kind of TLDs were packed in a polyethylene bag and set in the four experimental channels and on the front and the rear surfaces of the assembly. The TLDs were located at - 1, 58, 2t0, 362, 514 and 609 mm from the front surface of the assembly.
The irradiation was carried out for 16 minutes with deuteron beam current of 1.5 mA, and the total neutron yield was 2.31 x 1014. About 15 rain after the irradiation, TLDs were taken out from the experimental assembly and kept in a dark place. One week later from the irradiation, thermoluminescences (TL) were read out by a TLD reader (KYOKKO 2500). 2.3.2. Data reduction The background TL was subtracted, and the obtained TL was converted to the unit of exposure dose of 6°Co equivalence. Since TLDs were sensitive to neutrons as well as gamma-rays, neutron contribution on each TLD was subtracted from the total TL response. The neutron contribution was calculated by integrating the products of neutron response functions and neutron spectra. The neutron response function of each TLD were calculated with a code [16] developed by Hashikura et al. The neutron spectra were calculated with MCNP-4 [17] as described in the next section. Subsequently, since the atomic number of copper was 29, gamma-ray heating rate of copper was derived by interpolation of absorbed dose for MSO and SSO. In the obtained gamma-ray heating rates, some parasitic contributions were included; (i) gamma-rays generated at the target; (ii) beta- and gamma-rays associated with decay of 62Cu. These contributions were estimated based on neutron-photon-electron coupled transport calculations with MCNP-4 and the measured data were corrected for the contributions. The maximum correction was 25.9% at the front surface of the experimental assembly, and the corrections were less than 1% and could be negligible at positions more than 362 mm in depth. The overall errors ranged between 11 and 25% depending on measurement points.
3. Calculation procedure Calculations on the experiment was conducted with two kinds of transport code. One was the continuous energy Monte Carlo code MCNP-4 with the cross-section library FSXLIB-J3 [18,19] based on JENDL-3.1 [20]. Another was the two-dimensional discrete ordinate code DOT-3.5 [21] with P5-S16 approximation. The G R T U N C L code [21] was used to calculate the first collision sources. Two cross-section libraries, the JSSTDL [22,23] and FUSION-J3 [24], were employed for the DOT calculations. Both libraries were based on JENDL-3.1 and had the same group structure of 125 and 40 groups for neutron and gamma-ray, respectively, but the self shielding correction factors were
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F. Maekawa et al. / Fusion Engineering and Design 28 (1995) 753- 761
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Fig. 3. Gamma-ray spectra in the copper assembly at four positions measured and calculated by MCNP. considered only in JSSTDL. Source neutron spectrum calculated by the MORSE-DD code [25] simulating the real target structure as precise as possible was used for the source. Statistical errors for the MCNP calculations were reduced by adopting the source direction biasing method and the geometry splitting with the Russian roulette method changing cell importance double at every 150 mm of depth. The calculated gamma-ray spectra by M C N P and DOT were smeared with the energy resolution functions of the measured spectra. Gamma-ray heating rates were calculated by energy-integrating products of gamma-ray spectra and kerma factors which were derived from DLC-99 [26]. Gamma-ray production reactions of natural copper in JENDL-3.1 are gathered into two components; the non-elastic reaction (MT3) and the neutron capture reaction (MT102). Gamma-ray production reaction be-
low 0.2 MeV is presented in the MT102 while all the gamma-ray production reactions above 0.2 MeV are presented in the MT3 including the neutron capture reaction. The MCNP-4 code was modified so as to classify the calculated gamma-ray spectra into the two components, MT3 and MT102, according to the gamma-ray production reactions. The modification made it possible to easily identify inadequate component of spectra and reactions sensitive to the spectra.
4. Results and discussions
4.1. Gamma-ray spectrum Measured gamma-ray spectra at the four positions are shown in Fig. 3 with those calculated by MCNP. The MT3 component is dominant at 76 mm, and it
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F. Maekawa et al.
/Fusion Engineering and Design 28 (1995) 753-761
implies that most of gamma-rays are generated by threshold reactions with 14 MeV neutrons. The MT3 component decreases relatively to the MT102 component as increase of depth. At the deepest measurement point of 532 mm, the spectrum is dominated by the neutron capture reactions because of the remarkable MT102 component. Hence secondary gamma-ray data for non-threshold reactions, in addition to threshold reactions, can be tested by the measurements at different positions. This is one of the advantages of the present experiment because gamma-rays from non-threshold reactions have not been measured so far in the previous experiments. As for gamma-rays produced by threshold reactions mainly with 14 MeV neutrons, secondary gamma-ray data is usually verified only by comparing calculated gamma-rays with experiments because neutron transport calculations of copper is highly accurate at 14 MeV as proved in Ref. [6]. However, the accuracy of calculation for low energy neutrons, especially in a resonance region, is not always satisfactory. Since gamma-rays from non-threshold reactions are mainly produced by multiply scattered low energy neutrons, both neutron and gamma-ray should be examined to verify the secondary gamma-ray data. A s for gamma-rays from neutron capture reactions, intensity of gamma-rays is determined by the calculated neutron flux, while their spectral shapes are not so sensitive to those. t n Fig. 3, the calculated spectra in the energy range below 6 MeV at all measurement points agree with the experiment within about 20 %, except at around 0.5 and 2.2 MeV because of the experimental problem. In the energy range above 6 MeV, the calculated spectra are smaller than the experiment. Peaks around 7 - 8 MeV observed: in the experiment are not obtained by the calculations especially at 76, 228 and 380 mm. The peaks around 7 - 8 MeV are arisen from neutron capture reactions. Because Q-values of the capture reactions are 7.07 and 7.92 MeV for copper 63 and copper 65, respectively, and gamma-ray energies are equal to the Q-values when a nucleus on the capture state directly transits to the ground state. The neutron capture peaks are more obvious in the deeper positions where fractions of low energy neutrons to 14 MeV neutrons are more. Fig. 4 shows secondary gamma-ray emission probabilities (SGEPs) for the neutron capture reaction taken from JENDL-3.1. The SGEPs are based on the existing measurements for thermal energy and on the calculation for incident neutron energies of 10, 100 and 200 keV. A remarkable peak at 7 - 8 MeV is seen only for thermal energy. Such a big differences in SGEP between thermal energy and keV energies might not be reasonable.
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600
Fig. 5. Calculated to experimental values of gamma-ray heating rate of copper along the central axis of the assembly.
F. Maekawa et al. / Fusion Engineering and Design 28 (1995) 753 76I 4.2. Gamma-ray heating rate Fig. 5 shows calculated to experimental values (C/Es) of gamma-ray heating rate of copper. The C/Es of MCNP and DOT + JSSTDL are very close one another. The figure illustrates that the gamma-ray heating rates by the two calculations agree with the experiment almost within the experimental error ranges of 11 25%. On the other hand, gamma-ray heating rates by the DOT ÷ FUSION-J3 calculation are much larger than the experiment at positions deeper than 200 mm; C/Es are about 1.8 at 210 and 362 mm and 1.3 at 514 mm. This fact can be easily explained. If the self-shielding correction of cross-section is omitted as the DOT + FUSION-J3 calculation, calculated gamma-ray heating rates involve large ambiguity. If the self-shielding correction is taken into account, gamma-ray heating rate calculations by DOT with group constants have almost the same accuracy with continuous energy calculations even for gamma-rays from resonance capture reactions. At the front surface and 58 mm, C/Es for three calculations are almost the same. This fact implies that gamma-rays near the D - T source are mostly generated by threshold reactions with D - T neutrons. In the measurement positions deeper than 200 mm, differences of C/Es between two DOT calculations with and without the self-shielding correction are large. These differences prove t h a t gamma-rays in the positions are mostly produced by neutron capture reactions of which selfshielding correction is important. 4.3: Energy balance o f (n,7) reaction The energy sum of gamma-rays emitted from (n,7) reactions should be equal to the Q-value of the reaction when the energies of the incident neutron and recoil atom are negligible. This relation in nuclear data files can be tested by the measurement of gamma-ray heating rate and (n,7) reaction rate. The gamma-ray heating rate, i.e. absorbed energy in a media, is measured by the TLD technique. The heating rate is almost proportion to the energy sum of gamma-rays at the measurement point, so that the sum of released energy from the (n,?) reaction is estimated from the gamma-ray heating rate. On the other hand, the Cu(n,7) reaction rate at the deepest measurement position of 514 mm is estimated as follows. Since a fraction of the 65Cu(n,2n) reaction to the nateu(n, x)64Cu reaction is about 1% according to the MCNP calculation, the measured reaction rate, natCu(n,x)64Cu, a t the position is almost equal to
759
14 12
IJ.l~ ,
....................
-~v
4
F
"'! ...............
i
Cu-;
"
i
u-65
..............
- ........... Curves Based on i Ex~,~vted~ , ,. ,. the Principle : ,. of the Conservation of Energy
..................................L...........".......................L.....................
o02 0
, ....... ~ ........ I , ,,,,,,,I i ,,,,,,,I ........ I ........ I ........ I ...... 10-6 10 -7 10 -6 10-6 10 -4 10 -3 10 -2 10 "1
Neutron Energy [MeV]
Fig. 6. A plot to check the energy balance for the neutron capture reaction in JENDL-3.1. 63Cu(n,7) reaction rate as follows: natCu(n,x)64Cu
= 63Cu(n,~)
-}- 6 S C u ( n , 2 n )
~ 63Cu(n,7)
(1) At any position along the central axis of the experimental assembly through the front to the rear surface, the following equation stands up according to the MCNP calculation: 63Cu(n,'~) natCu(n,.~)
63Cu(n,'y)
63Cu(n,7) ÷ 65Cu(n,?)
= 0.79 _+ 0.02
(2)
This means that the ratio of the Eq. (2) is almost independent from changes of neutron spectrum. Considering the Eqs. (1) and (2), the natCu(n,?') reaction rate at the 514 mm is in proportion to the measured natCu(n,x)64Cu reaction rate: n a t C u ( n 7) ~ n a t C u ( n , x ) 6 4 C u / 0 - 7 9
off n a t e u ( n , x ) 6 4 C u .
(3)
At the 514 mm position, C/Es of gamma-ray heating rate and Cu(n, x)64Cu reaction rate are 1.02 and 0.84, respectively, for the MCNP calculation. Experimental errors for them are 11 and 4%, and statistical errors of the MCNP calculations are 2 and 3%, respectively. The calculated gamma-ray heating rate is in good agreement with the experiment. However this agreement does not mean that the secondary gamma-ray data in JENDL-3.1 are valid, because the calculated Cu(n,y) reaction rate, which is the source reaction for the gamma-ray production, is meaningfully smaller than the experiment. In other words, the underestimation of gamma-ray production reaction and the overestimation of released gammaray energies are cancelled each other so as to give the good agreement in the gamma-ray heating rate.
760
F. Maekawa et al. / Fusion Engineering and Design 28 (1995) 753 761
According to the discussions above, it is supposed that the secondary gamma-ray data of JENDL-3.1 is slightly assigned by excess energy of gamma-rays for the neutron capture reaction. This excess energy release can be attributed to the secondary gamma-ray data in JENDL-3.1. Fig. 6 is a plot to check the energy conservation for the neutron capture reaction. In the figure, total gamma-ray energy released by one neutron capture reaction is presented. The thick line is composed from JENDL-3.1 and the two thin lines are calculated based on the law of conservation of energy for 63Cu and 65Cu. Obviously the thick line must be in a range between the two thin lines. However, the released energy of JENDL-3.1 is out of the range around incident neutron energy of 1 keV. An energy of 12 MeV, about 1.5 times higher values to the expected, is released at 1 keV. Consequently it is indicated that the energy balance of gamma-ray emission in JENDL-3.1 is not consistent.
5. Concluding remarks G a m m a - r a y spectra and heating rates were measured in a copper slab assembly bombarded by D - T neutrons. The pulsed neutron method was adopted to reject decay gamma-rays correctly. The experiment presented the first benchmark experiment in which gamma-rays from not only threshold but also non-threshold reactions were measured. Through the experimental analyses, it was found for secondary gamma-ray data of copper in JENDL-3.1 that the gamma-ray data associated with threshold reactions were valid while modifications of spectral shapes arid energy balance were recommended Direct transition probabilities from the capture~excited state to the~ ground state for .non-thermal ne.utron energy should be increased for JENDL-3.1 reev~!uation. F o r applying grollp con~t.ant~ to calculations, self-shielding correction was inevitable for gamma-rays produced by neutron capture reactions in resonance region.
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