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Measurement and calculation of the neutron spectrum and gamma dosimetry of the fast neutron device at Xi’an Pulse Reactor
ARTICLE IN PRESS Nuclear Instruments and Methods in Physics Research A 593 (2008) 361– 366
Contents lists available at ScienceDirect
Nuclear Instruments and Methods in Physics Research A journal homepage: www.elsevier.com/locate/nima
Measurement and calculation of the neutron spectrum and gamma dosimetry of the fast neutron device at Xi’an Pulse Reactor Shu-Huan Liu , Xin-Biao Jiang, Nan-Nan Liu, Guang-Ning Zhu, Da Li, Ji-Hong Zhang, Qing-Yu Yu Northwest Institute of Nuclear Technology, Xi’an 710613, PR China
a r t i c l e in fo
abstract
Article history: Received 5 September 2006 Received in revised form 22 April 2008 Accepted 1 May 2008 Available online 27 May 2008
The neutron spectrum and neutron fluence of a fast neutron device at Xi’an Pulse Reactor were measured with multiple foil activation method and calculated with Monte Carlo technique in the paper. Thermo-luminescence detector (TLD) (7LiF(Mg,Ti)-1007M) is used to measure gamma dosimetry of the device. The experimental result proves that the goal of the design is achieved. & 2008 Elsevier B.V. All rights reserved.
Keywords: Fast neutron facility Neutron spectrum Gamma dosimetry TLD
1. Introduction Xi’an Pulse Reactor (Fig. 1) being a type of swimming pool reactor is affiliated with many experimental facilities and can be used in a variety of research activities, such as NAA, neutron detector calibration, biologic radiation effects and neutron photography, etc. It has two operation modes, one is steady-state (up to 2 MW), the other is pulsed mode (28.9 MJ in an 8.8 ms FWHM pulse that yields approximately 2 1013 n/cm2 in the irradiation chamber). Its largest experimental facility—irradiation chamber—is originally designed for the test of electronic chips and systems; however, the small ratio of its neutron fluence to gamma dose cannot satisfy fast neutron experiment conditions, so a device is needed to help finish this kind of research work using method of theory simulation and taking into account of the reactor’s feature, we designed such a device together with the typical neutron and gamma parameters in the device being calculated and measured.
2. Designed model for fast neutron device As its front aluminum shell of the irradiation chamber which reached out into the reactor water pool cannot burden any heavy apparatus, the experiment sample platform is hanged over shielding door. As the weight limit of the platform is 75 kg, the Corresponding author.
E-mail address: [email protected] (S.-H. Liu). 0168-9002/$ - see front matter & 2008 Elsevier B.V. All rights reserved. doi:10.1016/j.nima.2008.05.028
weight index for the device is set to about 50 kg. The radiation field parameters required in the device are as follows:
The rate of fast neutron fluence (En40.1 MeV) to thermal neutron fluence (Eno0.4 eV) is superior to 100.
The ratio of fast neutron fluence to gamma dose is no less than 5 1011 n/(cm2 Gy(Si)).
The fast neutron fluence should reach a level of 5 1013 n/cm2 after a proper irradiation period. Several types of structures were designed referring to the required values described above, and the box-type shielding structure was decided [1,4] on the basis of simulation results. Lead or bismuth is chosen as shielding material for the reactor gamma rays and powder state B4C or BN for thermal neutron. The powder state B4C or BN would be encapsulated with anticorrosive aluminum. The thickness of B4C layer is 8 mm, the density of B4C power is 1.0 g/cm3, the lead thickness of the front and behind walls are about 60 and 50 mm respectively, the size of the inner air cavity is 80 80 80 mm3 as showed in Fig. 2. The photograph of the fast neutron device is shown in Fig. 3.
3. Theory calculations and experiment measurements 3.1. Theory calculations With respect to the characteristics of the reactor core and geometry of irradiation channels, the technique of Monte Carlo
ARTICLE IN PRESS 362
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Fig. 1. The section of Xi’an Pulse Reactor: (a) transverse section and (b) longitudinal section.
advanced couple-sampling is introduced for the first time into the simulation of the deep penetration process of reactor particles [1], which lead to decrease of the variance of calculating results. The Boundary Source Transform Code (BSTC) and the channel-shielding calculation software named DOT-BSTP-DOT based on MCNP/ 4B were designed. In comparison with the benchmark, the calculation velocity and accuracy is greatly improved on the use of biasing technique. As a result, the theoretical design for the fast neutron device is reliable, and the relative biases for the calculated neutron spectrum and the gamma dosimetry are less than 10%. 3.2. Neutron spectrum measurements
Fig. 2. The boxed-shielding structure sketch for the fast neutron device.
Shielding door
Fast neutron device
The neutron spectrum of the designed fast neutron device was measured with multiple foil activation method. The foils were placed at positions of interest and irradiated for a given time. The quantity of some produced radio-nuclides during radiation was evaluated by counting beta or gamma ray. By selecting some foil detectors that are sensitive to the measurement neutron energy range, an estimate of the neutron spectrum and fluence could be obtained according to the relation between the foils activities and the neutron spectrum or neutron fluence. In the steady state case, the activated foil detectors obey activation Eq. (1): Z 1 Ai ¼ si ðEÞfðEÞ dE; i ¼ 1; 2; . . . ; I (1) 0
Irradiation
chamber Sample platform
Fig. 3. The photograph of the experimental fast neutron device.
where I is the number of the foil detectors; Ai is the experiment determined activity per target nucleus of the ith foil detector at the end of the neutron irradiation, Bq; si(E) is the energydependent activation cross-section, barn; and f(E) is the neutron difference flux density (n/cm2/eV/s). In order to measure the neutron spectrum of the whole energy range in the device, 19 kinds of foil detectors (Table 1) were chosen. After radiation, the activities of the foils were obtained by measuring the net counts of the different energy gamma peaks with HpGe gamma spectrometer. The experimental activity is
ARTICLE IN PRESS S.-H. Liu et al. / Nuclear Instruments and Methods in Physics Research A 593 (2008) 361–366
363
Table 1 The characteristic parameters of the foil detectors Nuclear reaction
Abundance (%)
Half live
Gamma energy, Eg (keV)
Relative intensity (%)
Cross-section library
Na23(n,g) Na24 Al27(n,a) Na24 Al27(n,p) Mg27 In115(n,g) In116m Ti46(n,p) Sc46 Ti47(n,p) Sc47 Ti48(n,p) Sc48 Zn64(n,p) Cu64 Cu63(n,g) Cu64 Co59(n,g) Co60 Ni58(n,p) Co58 Mg24(n,p) Na24 Mn55(n,g) Mn56 Fe54(n,p) Mn54 Fe56(n,p) Mn56 Au197(n,g) Au198 Mo98(n,g) Mo99 Lu176(n,g) Lu177 Dy164(n,g) Dy165
100.0 100.0 100.0 95.70 8.00 7.30 73.80 48.6 69.17 100.0 68.08 78.99 100.0 5.90 91.72 100.0 24.13 2.59 28.2
14.96 h 14.96 h 9.46 min 54.15 min 83.82 d 3.42 d 43.7 h 12.7 h 12.7 h 5.271 y 70.80 d 14.96 h 2.5785 h 312.1 d 2.58 h 2.695 d 66.02 h 161.04 h 2.334 h
1368.6 1368.6 843.73 1293.5 1120.5 159.40 1312.0 511.0 511.0 1332.50 810.8 1368.6 846.80 834.8 846.8 411.8 739.5 208.0 361.7
100.0 100 71.4 84.40 99.98 68.30 100.0 38.60 38.60 99.98 99.50 100.0 98.9 99.98 98.9 95.56 12.14 11.0 0.84
CENDL-2 CENDL-2 CENDL-2 JENDL-3 JENDL-3 JENDL-3 JENDL-3 JEF-2 ENDF/B-6 ENDF/B-6 ENDF/B-6 JENDL-3 ENDF/B-6 ENDF/B-6 ENDF/B-6 CENDL-2 ENDF/B-6 ENDF/B-6 ENDF/B-6
Table 2 Comparison of measured activities to calculated activities Nuclear reaction
Experiment foil activities, Am(i) (Bq)
Standard uncertainties of Am(i) (%)
Theory calculation foil activities, A(i) (%)
Am(i)/A(i)
Biasa (%)
Response energy range of foil detectors (eV)
Na23(n,g) Na24 Al27(n,a) Na24 Al27(n,p) Mg27 In115(n,g) In116m Ti46(n,p) Sc46 Ti47(n,p) Sc47 Ti48(n,p) Sc48 Zn64(n,p) Cu64 Cu63(n,g) Cu64 Co59(n,g) Co60 Ni58(n,p) Co58 Mg24(n,p) Na24 Mn55(n,g) Mn56 Fe54(n,p) Mn54 Fe56(n,p) Mn56 Au197(n,g) Au198 Mo98(n,g) Mo99 Lu176(n,g) Lu177 Dy164(n,g) Dy165
9.65e16 5.34e17 2.77e16 1.24e12 7.90e16 1.37e15 2.22e17 2.36e15 2.86e14 1.87e13 7.23e15 1.08e16 6.47e14 5.58e15 8.26e17 1.97e12 4.21e14 2.62e12 8.69e13
3.58 4.21 4.56 3.86 5.10 4.96 4.88 4.02 3.87 3.63 4.69 4.19 3.95 4.52 4.60 3.01 3.24 3.43 3.51
9.66e16 5.46e17 2.91e16 1.25e12 8.02e16 1.52e15 1.99e17 2.59e15 2.86e14 1.87e13 6.61e15 1.19e16 6.47e14 4.91e15 7.843e17 1.96e12 4.21e14 2.64e12 8.64e13
0.99 0.98 0.95 1.00 0.98 0.90 1.11 0.91 1.00 0.99 1.09 0.90 1.00 1.13 1.05 1.00 1.00 0.99 1.00
0.10 1.98 4.84 0.039 1.41 10.5 10.9 9.5 0.039 0.026 8.61 11.03 0.049 11.99 5.17 0.086 0.0072 0.78 0.60
1.00e42.80e5 5.90e61.27e7 3.10e69.0e6 0.403.80e5 3.30e68.70e6 1.30e66.60e6 5.10e61.33e7 2.10e66.90e6 50.001.30e6 2.205.00e3 1.20e66.90e6 5.90e61.21e7 0.882.55e4 1.90e67.00e6 5.00e61.18e7 2.201.15e4 11.505.00e5 0.0245.75e3 7.20e3 4.75e2
a
Notes: Bias ¼ |Am(i)A(i)|/Am(i) 100%.
defined as follows: Ai ¼
c li expðli t w Þ Mi gd kt ½1 expðli t c Þ½1 expðli t irr Þ ymi N0
(2)
where c is the net count of the detected gamma ray energy peak in gamma spectrum; gd is the relative intensity of the detected gamma ray; N0 is Avogadro constant (6.023 1023); tc is the gamma spectrum measuring life time, s; y is the isotope abundance of the ith nuclide; Mi is the atomic mass of the ith foil; li is the decay constant of the ith foil detector, s1; mi is the mass of the ith foil, g; tirr and tw are, respectively, the irradiation and cool time period of the ith foil, s; and kt is a correction factor. The experiment neutron spectrum is unfolded with SAND-II [2], its energy range from 1010 to 18 MeV is divided into 620 groups. When in steady state case with reactor power at 2.0 MW,
the calculated foil activities are obtained from Eq. (1). Comparison of experimental foil activities with theoretical calculated foil activities is shown in Table 2. As shown in Table 2, ratios of experimental activity to theoretical calculated activity for most foils are near to unity, the result proves that the agreements are fairly good. The measured neutron spectrum and the theoretical calculated are displayed in Figs. 4 and 5, respectively. The total neutron fluence jt of the measured neutron spectrum was 2.97 1011 n/(cm2 s). The equivalent 1 MeV neutron fluence, Feq,1 MeV was 1.68 1011 n/(cm2 s). The fast neutron fluence jf with energy En above 0.1 MeV was 2.01 1011 n/(cm2 s). The epithermal neutron fluence jepi with energy between 0.4 eV and 0.1 MeV was 9.75 1010 n/(cm2 s). The thermal neutron fluence jth with energy below 0.4 eV was 4.75 108 n/(cm2 s). The value of jf/jth was above 100.
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It was obvious that most of the thermal neutrons in the chamber were shielded by the device. The standard uncertainty of the experimental neutron spectrum was analyzed and calculated with Monte Carlo method [5,6]. The major factors (including the initial neutron spectrum and the foil detectors selected, the cross-section and the experimental activities uncertainty of the foils, etc.) that have influence on the unfolding neutron spectrum uncertainty were compared and analyzed. The influence of initial spectrum on the iterative neutron spectrum uncertainty is ignored. The uncertainty distribution of measured neutron spectrum induced by both the experimental activities and the neutron cross-section uncertainties were simulated, respectively. The simulation results showed that the measured foil activity uncertainties contribute about 7–20% to the resolved neutron spectrum uncertainty, while the cross-section uncertainties contribute 0.7–5%. The ultimate uncertainty distribution of the experimental spectrum induced by the two factors was from 10% to 30% in this work.
3.3. Gamma dose measurement The gamma dosimetry of the fast neutron device was measured using TLD. The 7LiF(Mg,Ti)-1007M detector’s experimental gamma dose linearity response range was from 0.01 mGy to 500 Gy, the non-uniformity is within 73%. 7LiF(Mg,Ti)-1007M detector is often used in measuring the gamma dose of mixed n–g irradiation field due to its low neutron reaction cross-section (Fig. 6) and non-sensitivity to neutron. In the procedure of gamma dose measurement, detectors were annealed with 2000B TLD far-infrared precise annealing furnace at 290 1C for 30 min and then selected by the standard source of 137 Cs, at last the absorbed dose was measured with thermoluminesence dosimetry system of type RGD-3A. The gamma dose–response of 7LiF(Mg,Ti)-1007M was calibrated with the secondary standard gamma source 60Co at Northwest Institute of Nuclear Technology. In terms of the experimental and calculation methods offered by ASTM E688-00, the gamma dose–response of 7 LiF(Mg,Ti)-1007M was shown in Fig. 7.
Fig. 4. The respective distribution of j(E) EE for experiment measured and theory calculated neutron spectrum of the fast neutron experiment device.
Fig. 5. The respective distribution of j(E)E for experiment measured and theory calculated neutron spectrum of the fast neutron experiment device.
ARTICLE IN PRESS S.-H. Liu et al. / Nuclear Instruments and Methods in Physics Research A 593 (2008) 361–366
Fig. 6. Neutron reaction cross-sections of (a) 7Li and (b)
365
19
F.
Table 3 The measured and calculated values of neutron flux and gamma dose rate on the inner front surface of the fast neutron device Typical parameters
jf ( 1011 cm2 s1) jepi ( 1011 cm2 s1) jth ( 108 cm2 s1) jt ( 1011 cm2 s1) _ g (Gy(Si)/s) D jf/jth _ g 1011 (n/ jf = D (cm2 Gy(Si)))
Calculated values
1.84 1.09 0.053 2.94 0.26 34521 7.07
Experimental measured values
Standard uncertainty of the experimental measured results (%)
2.01 0.98 4.76 2.97 0.35
2.01 4.12 2.46 2.21 30.0
422.57 5.74
3.31 20.78
4. Analysis and discussion
Fig. 7. The gamma dose–response curve of 7LiF(Mg,Ti)-1007M.
Considering the dose detection limit of 7LiF(Mg, Ti)-1007M, the TLD detector might be ruined by the high-density gamma rays and neutrons if the reactor runs in high power; however, the gamma dose in high reactor power can still be obtained through extrapolation method in terms of the low reactor power measurement results and the approximately proportional relations between gamma dose and reactor power. The dose rate measured with 7LiF(Mg, Ti)-1007M was about 34.91 rad(Si)/s. Taking into account the influence factors including the TLD uniformity, the energy response of the TLD, the bias of thermoluminescence dosimeter (RGD-3A) and the delayed gamma rays, etc. the uncertainty of the measured gamma dose was estimated about 30%. In short, the calculated values of the typical irradiation parameters [1] and the measured one with techniques described above were summarized in Table 3.
In Table 3, we find that the calculated and the measured results of jf, jepi and jt are consistent with each other, respectively. The _ g and the experimental measured are all calculated value of jf =D above 5 1011 n/(cm2 Gy(Si)), the values of jf/jth are all superior to 100. These results prove that the theory design models are reasonable. The reliability of the calculation models and the experimental method has been demonstrated in Refs. [1,3]. Although the values of the typical parameters both measured and calculated were reliable, the bias between the calculated results and measured results for thermal neutron fluence and gamma dose rate were very large. In one hand, the difference may be introduced by the fast neutron device technology (e.g., the very narrow gap between the fast neutron device shielding walls may result in thermal neutrons leakage) and the reactor delayed gamma rays; on the other hand, the nuclear reaction crosssections adopted in MCNP/4B and SAND-II unfolding spectrum programs are from different cross-section libraries, which could introduce some biases. References [1] X.-B. Jang, S.-H. Liu, Y.-H. Zhong, et al., Nucl. Power Eng. (in Chin.) 27 (1) (2006) 62. [2] S. Berg, W.N. McElory, Technical Report No. AFWL-TR-67-41, September.
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[3] S.-H. Liu, Studies on the measurement neutron spectra at experimental tubes in a reactor using multiple foil activation technique, Master Thesis, Northwest Institute of Nuclear Technology, Xi’an, 2000. [4] X.-B. Jiang, W. Chen, S.-H. Liu, Development and application of fast neutron experiment device at radiation cavity of Xi’an pulsed
reactor[R], GF Report, Northwest Institute of Nuclear Technology, Xi’an, 2004. [5] S.-H. Liu, D. Chen, Nucl. Power Eng. (in Chin.) 24 (3) (2003) 204. [6] W.N. McElroy, J.M. Marr, A Monte Carlo Program for SAND-II Error Analysis, HEDL-TME, 1973, pp. 73–20.
Contents lists available at ScienceDirect
Nuclear Instruments and Methods in Physics Research A journal homepage: www.elsevier.com/locate/nima
Measurement and calculation of the neutron spectrum and gamma dosimetry of the fast neutron device at Xi’an Pulse Reactor Shu-Huan Liu , Xin-Biao Jiang, Nan-Nan Liu, Guang-Ning Zhu, Da Li, Ji-Hong Zhang, Qing-Yu Yu Northwest Institute of Nuclear Technology, Xi’an 710613, PR China
a r t i c l e in fo
abstract
Article history: Received 5 September 2006 Received in revised form 22 April 2008 Accepted 1 May 2008 Available online 27 May 2008
The neutron spectrum and neutron fluence of a fast neutron device at Xi’an Pulse Reactor were measured with multiple foil activation method and calculated with Monte Carlo technique in the paper. Thermo-luminescence detector (TLD) (7LiF(Mg,Ti)-1007M) is used to measure gamma dosimetry of the device. The experimental result proves that the goal of the design is achieved. & 2008 Elsevier B.V. All rights reserved.
Keywords: Fast neutron facility Neutron spectrum Gamma dosimetry TLD
1. Introduction Xi’an Pulse Reactor (Fig. 1) being a type of swimming pool reactor is affiliated with many experimental facilities and can be used in a variety of research activities, such as NAA, neutron detector calibration, biologic radiation effects and neutron photography, etc. It has two operation modes, one is steady-state (up to 2 MW), the other is pulsed mode (28.9 MJ in an 8.8 ms FWHM pulse that yields approximately 2 1013 n/cm2 in the irradiation chamber). Its largest experimental facility—irradiation chamber—is originally designed for the test of electronic chips and systems; however, the small ratio of its neutron fluence to gamma dose cannot satisfy fast neutron experiment conditions, so a device is needed to help finish this kind of research work using method of theory simulation and taking into account of the reactor’s feature, we designed such a device together with the typical neutron and gamma parameters in the device being calculated and measured.
2. Designed model for fast neutron device As its front aluminum shell of the irradiation chamber which reached out into the reactor water pool cannot burden any heavy apparatus, the experiment sample platform is hanged over shielding door. As the weight limit of the platform is 75 kg, the Corresponding author.
E-mail address: [email protected] (S.-H. Liu). 0168-9002/$ - see front matter & 2008 Elsevier B.V. All rights reserved. doi:10.1016/j.nima.2008.05.028
weight index for the device is set to about 50 kg. The radiation field parameters required in the device are as follows:
The rate of fast neutron fluence (En40.1 MeV) to thermal neutron fluence (Eno0.4 eV) is superior to 100.
The ratio of fast neutron fluence to gamma dose is no less than 5 1011 n/(cm2 Gy(Si)).
The fast neutron fluence should reach a level of 5 1013 n/cm2 after a proper irradiation period. Several types of structures were designed referring to the required values described above, and the box-type shielding structure was decided [1,4] on the basis of simulation results. Lead or bismuth is chosen as shielding material for the reactor gamma rays and powder state B4C or BN for thermal neutron. The powder state B4C or BN would be encapsulated with anticorrosive aluminum. The thickness of B4C layer is 8 mm, the density of B4C power is 1.0 g/cm3, the lead thickness of the front and behind walls are about 60 and 50 mm respectively, the size of the inner air cavity is 80 80 80 mm3 as showed in Fig. 2. The photograph of the fast neutron device is shown in Fig. 3.
3. Theory calculations and experiment measurements 3.1. Theory calculations With respect to the characteristics of the reactor core and geometry of irradiation channels, the technique of Monte Carlo
ARTICLE IN PRESS 362
S.-H. Liu et al. / Nuclear Instruments and Methods in Physics Research A 593 (2008) 361–366
Fig. 1. The section of Xi’an Pulse Reactor: (a) transverse section and (b) longitudinal section.
advanced couple-sampling is introduced for the first time into the simulation of the deep penetration process of reactor particles [1], which lead to decrease of the variance of calculating results. The Boundary Source Transform Code (BSTC) and the channel-shielding calculation software named DOT-BSTP-DOT based on MCNP/ 4B were designed. In comparison with the benchmark, the calculation velocity and accuracy is greatly improved on the use of biasing technique. As a result, the theoretical design for the fast neutron device is reliable, and the relative biases for the calculated neutron spectrum and the gamma dosimetry are less than 10%. 3.2. Neutron spectrum measurements
Fig. 2. The boxed-shielding structure sketch for the fast neutron device.
Shielding door
Fast neutron device
The neutron spectrum of the designed fast neutron device was measured with multiple foil activation method. The foils were placed at positions of interest and irradiated for a given time. The quantity of some produced radio-nuclides during radiation was evaluated by counting beta or gamma ray. By selecting some foil detectors that are sensitive to the measurement neutron energy range, an estimate of the neutron spectrum and fluence could be obtained according to the relation between the foils activities and the neutron spectrum or neutron fluence. In the steady state case, the activated foil detectors obey activation Eq. (1): Z 1 Ai ¼ si ðEÞfðEÞ dE; i ¼ 1; 2; . . . ; I (1) 0
Irradiation
chamber Sample platform
Fig. 3. The photograph of the experimental fast neutron device.
where I is the number of the foil detectors; Ai is the experiment determined activity per target nucleus of the ith foil detector at the end of the neutron irradiation, Bq; si(E) is the energydependent activation cross-section, barn; and f(E) is the neutron difference flux density (n/cm2/eV/s). In order to measure the neutron spectrum of the whole energy range in the device, 19 kinds of foil detectors (Table 1) were chosen. After radiation, the activities of the foils were obtained by measuring the net counts of the different energy gamma peaks with HpGe gamma spectrometer. The experimental activity is
ARTICLE IN PRESS S.-H. Liu et al. / Nuclear Instruments and Methods in Physics Research A 593 (2008) 361–366
363
Table 1 The characteristic parameters of the foil detectors Nuclear reaction
Abundance (%)
Half live
Gamma energy, Eg (keV)
Relative intensity (%)
Cross-section library
Na23(n,g) Na24 Al27(n,a) Na24 Al27(n,p) Mg27 In115(n,g) In116m Ti46(n,p) Sc46 Ti47(n,p) Sc47 Ti48(n,p) Sc48 Zn64(n,p) Cu64 Cu63(n,g) Cu64 Co59(n,g) Co60 Ni58(n,p) Co58 Mg24(n,p) Na24 Mn55(n,g) Mn56 Fe54(n,p) Mn54 Fe56(n,p) Mn56 Au197(n,g) Au198 Mo98(n,g) Mo99 Lu176(n,g) Lu177 Dy164(n,g) Dy165
100.0 100.0 100.0 95.70 8.00 7.30 73.80 48.6 69.17 100.0 68.08 78.99 100.0 5.90 91.72 100.0 24.13 2.59 28.2
14.96 h 14.96 h 9.46 min 54.15 min 83.82 d 3.42 d 43.7 h 12.7 h 12.7 h 5.271 y 70.80 d 14.96 h 2.5785 h 312.1 d 2.58 h 2.695 d 66.02 h 161.04 h 2.334 h
1368.6 1368.6 843.73 1293.5 1120.5 159.40 1312.0 511.0 511.0 1332.50 810.8 1368.6 846.80 834.8 846.8 411.8 739.5 208.0 361.7
100.0 100 71.4 84.40 99.98 68.30 100.0 38.60 38.60 99.98 99.50 100.0 98.9 99.98 98.9 95.56 12.14 11.0 0.84
CENDL-2 CENDL-2 CENDL-2 JENDL-3 JENDL-3 JENDL-3 JENDL-3 JEF-2 ENDF/B-6 ENDF/B-6 ENDF/B-6 JENDL-3 ENDF/B-6 ENDF/B-6 ENDF/B-6 CENDL-2 ENDF/B-6 ENDF/B-6 ENDF/B-6
Table 2 Comparison of measured activities to calculated activities Nuclear reaction
Experiment foil activities, Am(i) (Bq)
Standard uncertainties of Am(i) (%)
Theory calculation foil activities, A(i) (%)
Am(i)/A(i)
Biasa (%)
Response energy range of foil detectors (eV)
Na23(n,g) Na24 Al27(n,a) Na24 Al27(n,p) Mg27 In115(n,g) In116m Ti46(n,p) Sc46 Ti47(n,p) Sc47 Ti48(n,p) Sc48 Zn64(n,p) Cu64 Cu63(n,g) Cu64 Co59(n,g) Co60 Ni58(n,p) Co58 Mg24(n,p) Na24 Mn55(n,g) Mn56 Fe54(n,p) Mn54 Fe56(n,p) Mn56 Au197(n,g) Au198 Mo98(n,g) Mo99 Lu176(n,g) Lu177 Dy164(n,g) Dy165
9.65e16 5.34e17 2.77e16 1.24e12 7.90e16 1.37e15 2.22e17 2.36e15 2.86e14 1.87e13 7.23e15 1.08e16 6.47e14 5.58e15 8.26e17 1.97e12 4.21e14 2.62e12 8.69e13
3.58 4.21 4.56 3.86 5.10 4.96 4.88 4.02 3.87 3.63 4.69 4.19 3.95 4.52 4.60 3.01 3.24 3.43 3.51
9.66e16 5.46e17 2.91e16 1.25e12 8.02e16 1.52e15 1.99e17 2.59e15 2.86e14 1.87e13 6.61e15 1.19e16 6.47e14 4.91e15 7.843e17 1.96e12 4.21e14 2.64e12 8.64e13
0.99 0.98 0.95 1.00 0.98 0.90 1.11 0.91 1.00 0.99 1.09 0.90 1.00 1.13 1.05 1.00 1.00 0.99 1.00
0.10 1.98 4.84 0.039 1.41 10.5 10.9 9.5 0.039 0.026 8.61 11.03 0.049 11.99 5.17 0.086 0.0072 0.78 0.60
1.00e42.80e5 5.90e61.27e7 3.10e69.0e6 0.403.80e5 3.30e68.70e6 1.30e66.60e6 5.10e61.33e7 2.10e66.90e6 50.001.30e6 2.205.00e3 1.20e66.90e6 5.90e61.21e7 0.882.55e4 1.90e67.00e6 5.00e61.18e7 2.201.15e4 11.505.00e5 0.0245.75e3 7.20e3 4.75e2
a
Notes: Bias ¼ |Am(i)A(i)|/Am(i) 100%.
defined as follows: Ai ¼
c li expðli t w Þ Mi gd kt ½1 expðli t c Þ½1 expðli t irr Þ ymi N0
(2)
where c is the net count of the detected gamma ray energy peak in gamma spectrum; gd is the relative intensity of the detected gamma ray; N0 is Avogadro constant (6.023 1023); tc is the gamma spectrum measuring life time, s; y is the isotope abundance of the ith nuclide; Mi is the atomic mass of the ith foil; li is the decay constant of the ith foil detector, s1; mi is the mass of the ith foil, g; tirr and tw are, respectively, the irradiation and cool time period of the ith foil, s; and kt is a correction factor. The experiment neutron spectrum is unfolded with SAND-II [2], its energy range from 1010 to 18 MeV is divided into 620 groups. When in steady state case with reactor power at 2.0 MW,
the calculated foil activities are obtained from Eq. (1). Comparison of experimental foil activities with theoretical calculated foil activities is shown in Table 2. As shown in Table 2, ratios of experimental activity to theoretical calculated activity for most foils are near to unity, the result proves that the agreements are fairly good. The measured neutron spectrum and the theoretical calculated are displayed in Figs. 4 and 5, respectively. The total neutron fluence jt of the measured neutron spectrum was 2.97 1011 n/(cm2 s). The equivalent 1 MeV neutron fluence, Feq,1 MeV was 1.68 1011 n/(cm2 s). The fast neutron fluence jf with energy En above 0.1 MeV was 2.01 1011 n/(cm2 s). The epithermal neutron fluence jepi with energy between 0.4 eV and 0.1 MeV was 9.75 1010 n/(cm2 s). The thermal neutron fluence jth with energy below 0.4 eV was 4.75 108 n/(cm2 s). The value of jf/jth was above 100.
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It was obvious that most of the thermal neutrons in the chamber were shielded by the device. The standard uncertainty of the experimental neutron spectrum was analyzed and calculated with Monte Carlo method [5,6]. The major factors (including the initial neutron spectrum and the foil detectors selected, the cross-section and the experimental activities uncertainty of the foils, etc.) that have influence on the unfolding neutron spectrum uncertainty were compared and analyzed. The influence of initial spectrum on the iterative neutron spectrum uncertainty is ignored. The uncertainty distribution of measured neutron spectrum induced by both the experimental activities and the neutron cross-section uncertainties were simulated, respectively. The simulation results showed that the measured foil activity uncertainties contribute about 7–20% to the resolved neutron spectrum uncertainty, while the cross-section uncertainties contribute 0.7–5%. The ultimate uncertainty distribution of the experimental spectrum induced by the two factors was from 10% to 30% in this work.
3.3. Gamma dose measurement The gamma dosimetry of the fast neutron device was measured using TLD. The 7LiF(Mg,Ti)-1007M detector’s experimental gamma dose linearity response range was from 0.01 mGy to 500 Gy, the non-uniformity is within 73%. 7LiF(Mg,Ti)-1007M detector is often used in measuring the gamma dose of mixed n–g irradiation field due to its low neutron reaction cross-section (Fig. 6) and non-sensitivity to neutron. In the procedure of gamma dose measurement, detectors were annealed with 2000B TLD far-infrared precise annealing furnace at 290 1C for 30 min and then selected by the standard source of 137 Cs, at last the absorbed dose was measured with thermoluminesence dosimetry system of type RGD-3A. The gamma dose–response of 7LiF(Mg,Ti)-1007M was calibrated with the secondary standard gamma source 60Co at Northwest Institute of Nuclear Technology. In terms of the experimental and calculation methods offered by ASTM E688-00, the gamma dose–response of 7 LiF(Mg,Ti)-1007M was shown in Fig. 7.
Fig. 4. The respective distribution of j(E) EE for experiment measured and theory calculated neutron spectrum of the fast neutron experiment device.
Fig. 5. The respective distribution of j(E)E for experiment measured and theory calculated neutron spectrum of the fast neutron experiment device.
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Fig. 6. Neutron reaction cross-sections of (a) 7Li and (b)
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F.
Table 3 The measured and calculated values of neutron flux and gamma dose rate on the inner front surface of the fast neutron device Typical parameters
jf ( 1011 cm2 s1) jepi ( 1011 cm2 s1) jth ( 108 cm2 s1) jt ( 1011 cm2 s1) _ g (Gy(Si)/s) D jf/jth _ g 1011 (n/ jf = D (cm2 Gy(Si)))
Calculated values
1.84 1.09 0.053 2.94 0.26 34521 7.07
Experimental measured values
Standard uncertainty of the experimental measured results (%)
2.01 0.98 4.76 2.97 0.35
2.01 4.12 2.46 2.21 30.0
422.57 5.74
3.31 20.78
4. Analysis and discussion
Fig. 7. The gamma dose–response curve of 7LiF(Mg,Ti)-1007M.
Considering the dose detection limit of 7LiF(Mg, Ti)-1007M, the TLD detector might be ruined by the high-density gamma rays and neutrons if the reactor runs in high power; however, the gamma dose in high reactor power can still be obtained through extrapolation method in terms of the low reactor power measurement results and the approximately proportional relations between gamma dose and reactor power. The dose rate measured with 7LiF(Mg, Ti)-1007M was about 34.91 rad(Si)/s. Taking into account the influence factors including the TLD uniformity, the energy response of the TLD, the bias of thermoluminescence dosimeter (RGD-3A) and the delayed gamma rays, etc. the uncertainty of the measured gamma dose was estimated about 30%. In short, the calculated values of the typical irradiation parameters [1] and the measured one with techniques described above were summarized in Table 3.
In Table 3, we find that the calculated and the measured results of jf, jepi and jt are consistent with each other, respectively. The _ g and the experimental measured are all calculated value of jf =D above 5 1011 n/(cm2 Gy(Si)), the values of jf/jth are all superior to 100. These results prove that the theory design models are reasonable. The reliability of the calculation models and the experimental method has been demonstrated in Refs. [1,3]. Although the values of the typical parameters both measured and calculated were reliable, the bias between the calculated results and measured results for thermal neutron fluence and gamma dose rate were very large. In one hand, the difference may be introduced by the fast neutron device technology (e.g., the very narrow gap between the fast neutron device shielding walls may result in thermal neutrons leakage) and the reactor delayed gamma rays; on the other hand, the nuclear reaction crosssections adopted in MCNP/4B and SAND-II unfolding spectrum programs are from different cross-section libraries, which could introduce some biases. References [1] X.-B. Jang, S.-H. Liu, Y.-H. Zhong, et al., Nucl. Power Eng. (in Chin.) 27 (1) (2006) 62. [2] S. Berg, W.N. McElory, Technical Report No. AFWL-TR-67-41, September.
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[3] S.-H. Liu, Studies on the measurement neutron spectra at experimental tubes in a reactor using multiple foil activation technique, Master Thesis, Northwest Institute of Nuclear Technology, Xi’an, 2000. [4] X.-B. Jiang, W. Chen, S.-H. Liu, Development and application of fast neutron experiment device at radiation cavity of Xi’an pulsed
reactor[R], GF Report, Northwest Institute of Nuclear Technology, Xi’an, 2004. [5] S.-H. Liu, D. Chen, Nucl. Power Eng. (in Chin.) 24 (3) (2003) 204. [6] W.N. McElroy, J.M. Marr, A Monte Carlo Program for SAND-II Error Analysis, HEDL-TME, 1973, pp. 73–20.