CH0100355 THE NEUTRON EMISSION METHOD FOR DETERMINATION OF FISSILE MATERIALS WITHIN THE SPENT FUEL -EQUIPMENT OPTIMIZATION.
ATTYAA.ABOU-ZAID Reactors Department, Nuclear Research Center Atomic Energy Authority, 13759- Cairo, Egypt (currently,Ph.D student at Atomic Energy Institute, POLAND)
and K. PYTEL Atomic Energy Institute, Research Reactor Center 05-400 Otwock- Swierk, POLAND
ABSTRACT A nondestructive assay method using neutron technique for determination the fissile isotopes content along the irradiated fuel rods of MARIA reactor is presented. This method is based on detection of the fission neutrons emitted from external neutron source and multiplied by the fissile isotopes U-235, Pu-239, and Pu-241 within the fuel rod. Neutrons emitted from the spent fuel originate mainly from induced fission in the fissile material and source neutrons penetrating the fuel rod without interaction. Additionally, the neutrons from (a, n) reaction and spontaneous fission of actinide isotopes contribute in the total population of emitted ones. The method gives a chance to perform an experimental calibration of the equipment using two points: fresh fuel rod (maximum signal plus background) and its mock-up (background). The Monte Carlo code has been used for the geometrical simulation and optimization of the measuring equipment: neutron source, moderating container, collimator, and the neutron detector. The results of calculation show that the moderating container of 30 cm length and 32 cm diameter and a collimator of 26 cm length, 6.8 cm width, and 2 cm height are the optimal configuration. With respect to the fission chamber position, the number of neutrons has been calculated as a function of distance from the fuel rod surface in the case of fresh fuel and its mock-up. The distance at which the ratio of the signal to background has maximum has been found at 4.5 cm far from the outer surface of the fuel.
1. Introduction
Neutron measurement method is preferred because the spent fuel contains a large amount of fissile isotopes that guarantees effective multiplication in the fuel and the emitted neutrons can be easily detected. This method requires the external neutron source which induce fission in the fissile isotopes U- 235, Pu-239, and Pu-241 within the fuel rod. These fission neutrons are detected by neutron detector positioned at the opposite side to the neutron source. The second main contributor to the detector signal comes from source neutrons penetrating the fuel rod without interaction (elastically scattered neutrons belong to this group). Another neutron source, so called inherent ones having relatively low importance: Thus, the total number of neutrons counted by fission chamber, CRt, is equal to [1,2,3]: CRt^CRf+CRd + CR; (1) where: CRf = neutrons due to fission in the fissile isotopes. CRa = neutrons coming directly from neutron source. CR; = neutrons due to the inherent neutron emission. The most dominant sources of inherent neutron emission are the spontaneous fission and the (a, n) reaction in case of oxide fuels. As an example, the neutron emission rates from spontaneous fission per 190 one MARIA fuel element are shown in table (1). The results have been obtained by ORIGEN [4] Code for MARIA spent fuel element with maximum burn-up of 110 MWd/element [5]. Among the components of the signal (equation (1)) the only CRa is not specific for fissile materials. This value should be subtracted from the total signal, CR*. to obtain the net signal proportional to the amount of fissile isotopes within investigated fuel. The subtraction can be done either numerically (see section 4) or experimentally using mock-up of a given fuel (section 5). TABLE (1) NEUTRONS FROM SPONTANEOUS FISSION. Isotope Half-life (Yr) spontaneous fission Pu-238 8.78E01 7.96E01 Pu-240 6.55E03 5.40E01 Pu-242 3.76E05 2.41E00 Cm-242 4.46E-01 9.37E01 Cm-244 1.81E01 4.94E01 2. Burn-up measurements of MARIA fuel element The fuel element of MARIA reactor (case study) contains six tubes with uranium enriched to 80 % of U-235 [5]. The burn-up of MARIA reactor fuel is measured by means of energy balance. This is possible for individual fuel elements because of the special design of MARIA reactor; each fuel element has its own cooling channel with individual measurement of water flow and temperatures, thus, the burn-up B of a given fuel element is calculated from the formula
B = J m Cp AT dt where:
m : cooling mass flow rate;
Cp : coolant specific heat; AT : outlet-inlet temperature difference and integration is over the operation time. Typically, burn-up of MARIA fuel is expressed in MWd/element. This method allows the evaluation of burn-up of the whole fuel element without information about the longitudinal burn-up distribution. The MARIA fuel has been chosen as a test case because it offers possibility of verification neutron emission method measurements by means of independent method. 3. Description of the method water The principle of the method is shown in fig. (1). The measuring installation is submerged under water to provide shielding against intensive gamma radiation emitted from the spent fuel. This method depends on the measurement of neutrons from fission in the fissile isotopes present in the moderating spent fuel. container A neutron source of Pu-Be type[6] has been used and it has a relatively high energy spectrum. Since these energetic neutrons have low fission cross section wi- Fig 1. Schematic diagram of the measuring equipment th the fissile isotopes, their thermalization is necessary. The source neutron is positioned in a center of cylinder filled with water playing a role of neutron moderating material. This cylinder is surrounded by a cadmium having thickness of 1 mm . To make possible measurement of the fissile material content along the spent fuel element, a collimator 191 of rectangular shape has been applied. The function of this collimator is to establish a narrow and wide beam of neutrons. The coUimator is inserted to the moderating container. Also the collimator is shielded with 1 mm cadmium and is filled with air. Collimated neutrons are interacting (fission, absorption, scattering) with the fuel and then detected by the neutron detector. To eliminate the signal background from gamma radiation, the U-235 fission chamber has been chosen. The fission chamber is not adjacent to the fuel surface. Certain distance from the fuel surface guarantees proper thermalization of fission neutrons in the water. The installation also gives a chance to insert a neutron absorbing material besides the fuel element opposite to neutron source. In this case, the amount of non-fission neutrons (e.g. scattered or transmitted from the neutron source) is minimized and the total fissile isotopes content are proportional to the number of fast neutrons resulting from fission. The arrangement and geometry of each component of the equipment i.e. neutron source, moderating container, collimator, and fission chamber play an important role in the measurement. The optimization of the equipment is required and it will be discussed in the next section. 4. Optimization of the components arrangement The components of the equipment are simulated and optimized numerically. The Monte Carlo MCNP4A Code [7] has been used for this purposes. Although Monte Carlo method is time consuming, it is well suited for complicated and complex problems that can not be modeled by computer codes which use deterministic methods. The objective function of the optimization is the maximum integral of thermal neutron current multiplied by energy dependent fission cross section for U-235 escaping from the neutron source, moderating container, and collimator towards the fuel element as well as the maximum neutron flux emitted from the fuel element and counted by fission chamber as in the following equations:* us j(r,E,J-0 dA dp. (2) E A u u c|)(r,E) af{E)dE dA/A (3) E A where, j(r,E,u) dE dA dp.: expected number of neutrons passing through an area dA with energy E in dE, direction u, in du. cJf{E) : the energy dependent microscopic fission cross section for U-235 u : the cosine of the angle between surface normal and neutron trajectory. (|)(r,E) dA dE : expected number of neutrons in dA with energy E about dE. The geometry of the equipment is too complicated to perform the optimization within one computation step. The whole optimization problem has been spilitted into three separate steps: moderating container, collimator and, fuel-detector geometry.
4.1 Moderating Container To get the optimum diameter of the moderating container having the geometry of a cylinder, the current-cross section integral (Eq.(2)) is calculated as a function of the distance from the neutron source surface. These calculations are repeated for neutron moderating Distance from the neutron source centre ,cm Fig 2. The number of neutrons as a function container having the diameters from 8 cm up to 34 cm. of the distance from the neutron source surface. The results are shown in fig. (2) and it is observed that with increasing the diameter of the moderating container, the current-cross section integral increases. The increasing in the current-cross section integral continue up to certain diameter (30 cm ) and further
* we will refer to this quantities as: current-cross section integral and flux-cross section integral 192 extension of the diameter is not very effective. In this study a diameter of 32 cm is chosen which gives a maximum of current-cross section integral at about 3 cm far from neutron source surface. 4.2 The Collimator It is assumed that the geometry of the collimator is rectangular of 6.8 cm width (the effective diameter of the O1: fuel rod). With respect to the collimator height, a three < A colllmatlon ratio = 2.OIL cases were studied , one slide of 2 cm height, two slides £ 9 coBlmstlon ratio = 1.0/L • colllmation ratio = C5IL separating with neutron absorbing material each one of 1 £ cm height, and four slides each of them of 0.5 cm height ~ °01 (also these slides are separated from each other by neutron S absorbing material). For every case the length was ^ assumed from 12 cm up to 26 cm and the collimator face °* at the maximum point in the moderating container i.e. 3 cm far from the neutron source surface. The Monte Carlo calculations have been performed and the 16 20 24 results are shown in fig.(3). The figure shows that with collimator length {L\ cm increasing collimator height, the current-cross section Fig. 3. The integration as a function of collimator length and collimation ratio integral increases. Also it is seen that with increasing collimator length, the current-cross section integral decreases. Starting from certain length (24 cm) the current-cross section integral tends to be approximately constant. In this study a collimator of 26 cm length and 2 cm height is chosen. 4.3 Fission Chamber With respect to the fission chamber position, the flux-cross section integral (Eg. (3)) is calculated as a function of distance from the surface of the fuel element. The calculations have been performed for the fresh fuel and mock-up of the fuel as shown in fig.(4). The difference in value between the two curves in fig.(4) is due to the fission neutrons and the inherent neutron emission.The fission chamber is positioned at the distance in which the ratio of number of neutrons emitted from fresh fuel over the same number from the mock-up of the fuel (signal to back-ground ratio) has maximum and it is found on bout 4.5 cm. from the fuel rod surface (see fig.(5)).
u.uue> • with fml A without fuel
0.006 —|
r W 0.004 — It
' 1 ' I 0 2 4 6 8 0 2 4 6 distance from thefuel rod surface.cm distance from the fuel rod surface, cm Fig.5. The ratio of the flux-cross section integral Fig.4 The flux-cross section integral as a functom for fresh fuel over the same value of its mock-up of the distance from the fuel rod surface. versus the distance from the surface of the fuel.
5. Method of fissile isotopes measurement in the spent fuel. To measure the amount of fissile isotopes within the spent fuel, two calibration points at least are required. These two points are the count rate from the fresh fuel and from the mock-up of the fuel which 193 corresponds to the nominal content and zero content of fissile isotopes respectively. The relation between the neutron count rate and the number of fissile material is assumed to*be linear as shown in fig(6). The number of fissile isotopes of any spent fuel i, (Ns)j can then be calculated using the following relation
_ (CRs-ACRs)i-r.CRm Nf ~ (CRf-ACRf)~r.CRm where: CRf = count rate of fission chamber from fresh fuel rod due to the presence of neutron source ACRf = count rate of fission chamber from fresh fuel rod without neutron source = count rate of fission chamber from spent fuel rod due to the presence of neutron source. ACRs = count rate of fission chamber from spent fuel rod without neutron source = count rate of fission chamber from fuel rod mock-up due to the presence of neutron source = correction factor due to the difference in neutron transmission between the fuel material and the mock-up material (Aluminum) N, = fissile isotope content in the fresh fuel element (nominal value) In MARIA reactor the quantity of U-238 in the fuel is low and the term ACRf can be neglected and the above equation becomes (nominal value) z
Nf - CRf-r.CR,,,
Thus, the number of fissile isotopes, (Ns); ,of the spent Isotope s fCNsh / 7 fuel element i is equal to : VP „ (CRs-ACRs)i-r.CRm "S Nf quin u (JNs)i = CRf-r.CRm 6. Conclusions / Numerical simulation of a neutron emission method applied to the measurement of fissile isotopes content in Nm (zero value) / r.CRm (CRs-4CRs)i (CRf-ACRf) irradiated fuel proved feasibility of the method. The neutron count rate experimental equipment consisting of: neutron source, Fig 6. The relation between neutron count rate and the number of fissile isotopes moderating container, collimator, fuel rod, and fission chamber has been optimized by Monte Carlo MCNP-4A code. The output signal is dominated by two components: counts due to fission neutrons and neutrons directly transmitted from the source(reported as background). The only first component is proportional to the fissile material content and the experimental method for subtraction of this information from the output signal has been proposed and discussed. Acknowledgment The authors would like to express their thanks to Dr. K. Andrzejewski for his assistance and fruitful discussion in Monte Carlo calculations. References [1] J. R. Philips et al, LA-9002-Ms, Los Alamos National Laboratory, 1981. [2] H. Wurz et al., Nuc. Tech., Vol. 90, May 1990,p.l91. [3] H. Wurz, Nuc. Tech., Vol. 95, August 1991,p. 193. ' [4] M. J. Bell ,ORIGEN - The ORNL Isotope Generation and Depletion Code, ORNL-4628 (May 1973) [5] W. Byszewski, et al., Nucleonika, Vol.31- No 1 l-12/76,p. 1257. [6] L. Stewart, Physical Review, Vol.98, No.3, May l,1955,p.74O. [7] J. F. Briesmeister, Ed., ,MCNP-4A General Monte Carlo N-Particle Transport Code, Version 4A, LA-12625, 1993.
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