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Nuclear Physics

Assessment of fissionable material behaviour in fission chambers

Profile image of MARÍA NATIVIDAD HERRANZ ALFAROMARÍA NATIVIDAD HERRANZ ALFARO

2010, Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment

Abstract

A comprehensive study is performed in order to assess the pertinence of fission chambers coated with different fissile materials for high neutrón flux detection. Three neutrón scenarios are proposed to study the fast component of a high neutrón flux: (i) high neutrón flux with a significant thermal contribution such as BR2, (ii) DEMO magnetic fusión reactor, and (iii) IFMIF high flux test module. In this study, the inventory code ACAB is used to analyze the following questions: (i) impact of different deposits in fission chambers; (ii) effect of the irradiation time/burn-up on the concentration; (iii) impact of activation cross-section uncertainties on the composition of the deposit for all the range of burn-up/irradiation neutrón fluences of interest. The complete set of nuclear data (decay, fission yield, activation cross-sections, and uncertainties) provided in the EAF2007 data library are used for this evaluation.

Figures (18)
Fig. 1. Neutron spectra of different neutron environments in a VITAMIN-J+ (211) multi-group structure.
Fig. 1. Neutron spectra of different neutron environments in a VITAMIN-J+ (211) multi-group structure.
Isotopic composition of the potential fissile deposit or solutions. Table 1 in a neutron field may be written in matrix notation as The sensitivity to the neutron spectrum is assessed as well. Then, in Section 3, the influence of different parameters on the FC behaviour is analyzed: radioactivity, xenon inventory, temperature, and deposit thickness. In Section 4, the impact of the uncertainties in activation cross-sections on the total fission rates and on the sensitivities to the neutron spectrum are performed, by using the Monte Carlo technique. The most contributing isotopes to the error in fission rates for deposits of Pu242#2 and Np237 are shown. Finally, in Section 5, conclusions for the different realistic fissile deposits analyzed for a long-term performance are drawn.
Isotopic composition of the potential fissile deposit or solutions. Table 1 in a neutron field may be written in matrix notation as The sensitivity to the neutron spectrum is assessed as well. Then, in Section 3, the influence of different parameters on the FC behaviour is analyzed: radioactivity, xenon inventory, temperature, and deposit thickness. In Section 4, the impact of the uncertainties in activation cross-sections on the total fission rates and on the sensitivities to the neutron spectrum are performed, by using the Monte Carlo technique. The most contributing isotopes to the error in fission rates for deposits of Pu242#2 and Np237 are shown. Finally, in Section 5, conclusions for the different realistic fissile deposits analyzed for a long-term performance are drawn.
Collapsed one-group fission cross-section (barns), fraction of fissions for neutron energies above 1 MeV (in %) and fission/capture ratio
Collapsed one-group fission cross-section (barns), fraction of fissions for neutron energies above 1 MeV (in %) and fission/capture ratio
Total neutron flux and fraction of neutrons in four energy ranges for the three neutron scenarios
Total neutron flux and fraction of neutrons in four energy ranges for the three neutron scenarios
Fig. 2. Fission rates (fission/s) for the different pure and solution deposits in a typical high flux thermal neutron environment (BR2)
Fig. 2. Fission rates (fission/s) for the different pure and solution deposits in a typical high flux thermal neutron environment (BR2)
Fig. 3. Sensitivity to fast neutrons for initially different pure and solution deposits in a typical high flux thermal neutron environment (BR2)
Fig. 3. Sensitivity to fast neutrons for initially different pure and solution deposits in a typical high flux thermal neutron environment (BR2)
Fig. 4. Total activity evolution (mCi) induced in the fission chamber for initially different pure (Np237 and Pu242) and solution (U235 and U238) deposits in typical high flux neutron environments.
Fig. 4. Total activity evolution (mCi) induced in the fission chamber for initially different pure (Np237 and Pu242) and solution (U235 and U238) deposits in typical high flux neutron environments.
Fig. 5. Contamination of the filling gas by xenon (atoms) induced in the fission chamber for initially different pure (Np237 and Pu242) and solution (U235 and U238 deposits in typical high neutron flux environments. First, to analyze the self-shielding phenomenon (that is, less neutron absorption in the inner layers of the deposit), we predict In this section, we assess the temperature effect on the total fission rates for different deposits irradiated in BR2 [13]. In this reactor, the temperature dependence can be explained by the fact that the thermal neutron flux, described by the Maxwell distribution, becomes harder with increasing temperature, and consequently the one-group cross-sections will change.
Fig. 5. Contamination of the filling gas by xenon (atoms) induced in the fission chamber for initially different pure (Np237 and Pu242) and solution (U235 and U238 deposits in typical high neutron flux environments. First, to analyze the self-shielding phenomenon (that is, less neutron absorption in the inner layers of the deposit), we predict In this section, we assess the temperature effect on the total fission rates for different deposits irradiated in BR2 [13]. In this reactor, the temperature dependence can be explained by the fact that the thermal neutron flux, described by the Maxwell distribution, becomes harder with increasing temperature, and consequently the one-group cross-sections will change.
Fig. 7. Predicted maximum thickness assuming an interaction probability (¢) of 0.001. Fig. 6. Temperature dependence of the fission rates for different deposits. Branching cases show the difference in % between a reference irradiation case at 325 K and an instantaneous change in the temperature of +25 K. In the spectral history cases, we compare the nominal case at 325K with a modified irradiation case at 350K.
Fig. 7. Predicted maximum thickness assuming an interaction probability (¢) of 0.001. Fig. 6. Temperature dependence of the fission rates for different deposits. Branching cases show the difference in % between a reference irradiation case at 325 K and an instantaneous change in the temperature of +25 K. In the spectral history cases, we compare the nominal case at 325K with a modified irradiation case at 350K.
Fig. 8. On the left axis, the roughly total kinetic energy of any fragment with mass number A, mean and relative error kinetic energy (one standard deviation) from fission of 735U(ny,,f) taken from Ref. [16] and the stopping power (MeV/,1m) at this energy calculated with SRIM in uranium for different elements (charge Z) with the same mass number. On the right axis, the effective fission product yield for ?75U collapsed by ACAB code for different neutron spectra.
Fig. 8. On the left axis, the roughly total kinetic energy of any fragment with mass number A, mean and relative error kinetic energy (one standard deviation) from fission of 735U(ny,,f) taken from Ref. [16] and the stopping power (MeV/,1m) at this energy calculated with SRIM in uranium for different elements (charge Z) with the same mass number. On the right axis, the effective fission product yield for ?75U collapsed by ACAB code for different neutron spectra.
Total relative experimental error (in %) for capture and fission cross sections from EAF2007/UN.
Total relative experimental error (in %) for capture and fission cross sections from EAF2007/UN.
Fig. 9. Relative error in fission rates (in %) for different pure and solution deposits in a typical high flux thermal neutron environment (BR2).
Fig. 9. Relative error in fission rates (in %) for different pure and solution deposits in a typical high flux thermal neutron environment (BR2).
4.2. Monte Carlo method where a; is the fraction of isotope j contributing to the total fission rate. And F is the total number of fissile isotopes.
4.2. Monte Carlo method where a; is the fraction of isotope j contributing to the total fission rate. And F is the total number of fissile isotopes.
Fig. 10. Relative error in Sensitivity to fast neutrons (in %) for different pure and solution deposits in a typical high flux thermal neutron environment (BR2) Here, the implemented Monte Carlo (MC) method in ACAB code is applied to FC uncertainty analysis, showing its capability to deal with overall cross-section uncertainty data. The MC method is based on simultaneous random sampling of all the cross-sections involved in the problem, where uncertainties are taken from EAF2007/UN. Authors of the EAF library have noted
Fig. 10. Relative error in Sensitivity to fast neutrons (in %) for different pure and solution deposits in a typical high flux thermal neutron environment (BR2) Here, the implemented Monte Carlo (MC) method in ACAB code is applied to FC uncertainty analysis, showing its capability to deal with overall cross-section uncertainty data. The MC method is based on simultaneous random sampling of all the cross-sections involved in the problem, where uncertainties are taken from EAF2007/UN. Authors of the EAF library have noted
Fig. 11. Fission rates and relative error (in %) for different U235 and U238 deposits in DEMO and HFTM/IFMIF neutron environments
Fig. 11. Fission rates and relative error (in %) for different U235 and U238 deposits in DEMO and HFTM/IFMIF neutron environments
Fig. 12. Contribution and error bars (one standard deviation) of each isotope in the total fission rate for a deposit of Pu242#2 (see Table 1 for initial composition) in a typical high flux thermal neutron environment (BR2). that when evaluating cross-sections, the quantity log(Gexp:/Ocaic) was approximately normally distributed. Consequently, it can be said that for any given cross-section o we can define the random variable log(o/o) that follows a normal distribution with mean 0 and variance Ajyp. This well known distribution is equivalent to a normal distribution when A is small. [18] Using this uncertainty information, and sampling with the adequate probability distribution, we obtain the transition matrix, A. With this A matrix, ACAB computes the vector of nuclides (N) and response quantities (Riot, Sfast, ...) per nuclide. Repeating this sequence, it is possible to get a sample of vectors and, from this sample; we can estimate the mean, variance, etc. of the nuclide
Fig. 12. Contribution and error bars (one standard deviation) of each isotope in the total fission rate for a deposit of Pu242#2 (see Table 1 for initial composition) in a typical high flux thermal neutron environment (BR2). that when evaluating cross-sections, the quantity log(Gexp:/Ocaic) was approximately normally distributed. Consequently, it can be said that for any given cross-section o we can define the random variable log(o/o) that follows a normal distribution with mean 0 and variance Ajyp. This well known distribution is equivalent to a normal distribution when A is small. [18] Using this uncertainty information, and sampling with the adequate probability distribution, we obtain the transition matrix, A. With this A matrix, ACAB computes the vector of nuclides (N) and response quantities (Riot, Sfast, ...) per nuclide. Repeating this sequence, it is possible to get a sample of vectors and, from this sample; we can estimate the mean, variance, etc. of the nuclide
Fig. 13. Total fission rate and error bars (one standard deviation) for initially pure Np237 deposit irradiated in a high flux thermal environment (e.g. BR2). Contributions of! each isotope to the total fission rate and errors are shown.
Fig. 13. Total fission rate and error bars (one standard deviation) for initially pure Np237 deposit irradiated in a high flux thermal environment (e.g. BR2). Contributions of! each isotope to the total fission rate and errors are shown.

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