2011 International Nuclear Atlantic Conference - INAC 2011 Belo Horizonte,MG, Brazil, October 24-28, 2011 ASSOCIAÇÃO BRASILEIRA DE ENERGIA NUCLEAR - ABEN ISBN: 978-85-99141-04-5 MEASURED AND CALCULATED EFFECTIVE DELAYED NEUTRON FRACTION OF THE IPR-R1 TRIGA REACTOR Rose Mary G. P. Souza, Hugo M. Dalle, Daniel A. M. Campolina Centro de Desenvolvimento da Tecnologia Nuclear (CDTN) Comissão Nacional de Energia Nuclear (CNEN) Campus da UFMG - Pampulha – Caixa Postal 941 30123-970, Belo Horizonte, MG, Brazil. [email protected], [email protected]; [email protected] ABSTRACT The effective delayed neutron fraction, beff , one of the most important parameter in reactor kinetics, was measured for the 100 kW IPR-R1 TRIGA Mark I research reactor, located at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil. The current reactor core has 63 fuel elements, containing about 8.5% and 8% by weight of uranium enriched to 20% in U235. The core has cylindrical configuration with an annular graphite reflector. Since the first criticality of the reactor in November 1960, the core configuration and the number of fuel elements have been changed several times. At that time, the reactor power was 30 kW, there were 56 fuel elements in the core, and the beff value for the reactor recommended by General Atomic (manufacturer of TRIGA) was 790 pcm. The current beff parameter was determined from experimental methods based on inhour equation and on the control rod drops. The estimated values obtained were (774 ± 38) pcm and (744 ± 20) pcm, respectively. The beff was calculated by Monte Carlo transport code MCNP5 and it was obtained 747 pcm. The calculated and measured values are in good agreement, and the relative percentage error is -3.6% for the first case, and 0.4% for the second one. 1. INTRODUCTION The nominal thermal power of the IPR-R1 TRIGA reactor, at the Nuclear Technology Development Center - CDTN, Belo Horizonte, Brazil, is 100 kW. The work presented in this paper shows the effective fraction of delayed neutrons (beff) calculated by Monte Carlo transport code MCNP5, and estimated experimentally by two methodologies: the inhour equation and the rod drop. One of the most important aspects of the fission process from the point of view of controlling the reactor is the presence of delayed neutrons. In general, delayed neutrons are more effective to control the reactor than prompt neutrons because they are born at lower energy compared to prompt (fission) neutrons. Thus, they have a better chance to survive leakage and resonance absorption. A delayed neutron is a neutron emitted by an excited fission product nucleus during beta disintegration some appreciable time after the fission. There are six decay chains which are of significance in the emission of delayed neutrons. Correspondingly, delayed neutrons are commonly discussed as being in six groups. Each of these groups, (i), are characterized by a fractional yield bi and a decay constant li. The amount bi is defined as the fraction of the total number of fission neutrons that are emitted by the decay of fission products in the group i. The effective total delayed neutron fraction is designated beff. The value of this parameter for a given fuel will vary over the lifetime of the reactor core, and also with the average energy of the neutrons producing fission. βeff depends on the core size and presents higher values for smaller cores. The effective fraction of delayed neutrons recommended by the General Atomic Technologies inc. (manufacturer of TRIGA) is 790 pcm [1, 2]. 2. BRIEF DESCRIPTION OF THE IPR-R1 REACTOR The IPR-R1 TRIGA reactor is a 100 kW facility. The reactor reached its first criticality on November 1960, and since then its core configuration and the number of fuel elements have been changed several times. As calculated in [3] about 4% of the U-235 in the core was burned during its 50 years. The fuel is a homogeneous mixture of uranium-zirconium hydride containing about 8.5% and 8% by weight of uranium, for stainless-steel and aluminum clad elements, respectively, enriched to 20% in 235U [1, 2]. The reactor core has 91 positions of which 63 are filled with fuel elements (59 Al-clad and 4 SS-clad elements) arranged in five concentric rings, inside a water tank. In the F ring there are 23 graphite dummy elements. The power level of the reactor is controlled by three control rods: Regulating, Shim, and Safety. The reactor is equipped with three irradiation facilities: rotary specimen rack, pneumatic transfer tube, and central thimble. A complete description of the IPR-R1 reactor can be found in [4]. Figure 1 shows the current core configuration which was used in the experiments. Figure 1. IPR-R1 TRIGA reactor core. INAC 2011, Belo Horizonte, MG, Brazil. 3. MONTE CARLO CALCULATION METHOD AND MODEL The general purpose, continuous energy, generalized-geometry Monte Carlo transport code MCNP5 was used in the calculations. In MCNP it is possible to model the reactor in 3-D geometry in details. A very detailed description of the MCNP reactor core model can be found in [3]. In this model the fuel burnup since the first criticality of the IPR-R1 was simulated taking into account the power history as recorded in the reactor log books. ENDF/B-VII pointwise cross sections data libraries were used in the simulations. When such data were not available then ENDF/B-VI data was used, instead. Three core configurations were simulated. The first two are the Beginning Of Life (BOL) core, as it was in 1960, considering one configuration with graphite elements filling the positions into the F ring and the other with water filling the positions in F ring. The third core is the current core, as shown in Fig. 1. The effective delayed neutron fraction, βeff, was calculated using the following equation [5, 6]: kP b eff = 1 - (1) keff where kp is the effective neutron multiplication factor only for prompt neutrons and keff is the effective neutron multiplication factor for both, prompt and delayed neutrons. The MCNP calculations were run with 4.0x105 histories per cycle and 2500 active cycles of neutrons, leading to a standard deviation within 0.00008 (8 pcm). The results of calculated βeff for all the three core configurations are presented in the Table 1. One can notice that the calculation result of the BOL core without any dummy elements surrounding the core agrees very well with the nominal value of 0,00790 for βeff informed by the reactor manufacturer, GA. This may be an indirect indication in favor of the adopted calculation model. Furthermore, it can be observed that when graphite dummy elements are filled into the F ring, then the βeff value reduces, as expected because the graphite elements work as a reflector, diminishing the leakage of the high energy prompt neutrons and thus improving the contribution of these neutrons in the balance. The current core is larger than the BOL, thus, the additional reduction in the βeff is also expected as the effective delayed neutron fraction depends on the core size and presents higher values for smaller cores. Such results agree with other authors [5, 6] as the delayed neutrons are born with lower energies and are more effective in inducing fission in system with larger leakage. INAC 2011, Belo Horizonte, MG, Brazil. Table 1. Reactor core configurations and calculated βeff Number of Date of the Element in Error β Fuel Factor Mean eff Configuration the F Ring (pcm) (pcm) Elements keff 0.99576 8 06/11/1960 56 H2O 789 kp 0.98790 8 keff 1.02254 8 06/11/1960 56 Graphite 750 kp 1.01487 8 keff 1.02183 8 06/11/2004 63 Graphite 747 kp 1.01420 8 4. EXPERIMENTAL DETERMINATION OF DELAYED NEUTRON FRACTION A brief discussion of the methodologies that were adopted to obtain beff and the results of the experiments performed for such purposes are given in the following items. 4.1. Measurement Results Based on Inhour Equation Theoretically, the reactivity of a control rod can be given approximately by [7, 8]: é Z 1 æ 2p Z öù r (Z ) = r (H ) ê - sen ç ÷ú (2) ë H 2p è H øû where r (Z) and r (H) refer to the control rod reactivity worth inserted the distance Z and fully inserted, respectively. Knowing the control rod worth r (H), the value of the reactivity at different Z positions can be calculated. The integral curve of the Regulating control rod is shown in Fig. 2, and of the Shim and Safety rods in Fig. 3. These curves were obtained by the Eq. (2). The total reactivity worth of the Regulating, Safety and Shim rods are 379 pcm, 2244 pcm and 2536 pcm, respectively [9]. INAC 2011, Belo Horizonte, MG, Brazil. 400 350 300 250 200 150 Reactivity (pcm) 100 50 0 0 200 400 600 800 1000 Regulating Rod Position Figure 2. Integral curve of the Regulating control rod. 3000 2500 2000 Safety 1500 Shim 1000 Reactivity (pcm) 500 0 0 200 400 600 800 1000 Rod Position Figure 3. Integral curves of the Shim and Safety control rods. The relationship between a step change in reactivity (r) and the resulting reactor period (T) is given by the inhour equation: w m b w r = l + å i (3) k 1 w + li In which l is the prompt neutron generation time; m is the number of delayed neutron groups; w are the (m+1) solutions, or roots, of the inhour equation; and k is the effective neutron multiplication factor. The stable period of the reactor is defined by the Eq.