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Neutronic Characterization and Decay Heat Calculations in the In-Vessel Fuel Storage Facilities for MYRRHA/FASTEF ⇑ S

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Neutronic Characterization and Decay Heat Calculations in the In-Vessel Fuel Storage Facilities for MYRRHA/FASTEF ⇑ S Energy Conversion and Management 64 (2012) 522–529 Contents lists available at SciVerse ScienceDirect Energy Conversion and Management journal homepage: www.elsevier.com/locate/enconman Neutronic characterization and decay heat calculations in the in-vessel fuel storage facilities for MYRRHA/FASTEF ⇑ S. Di Maria a, , M. Ottolini b, E. Malambu Mbala c, M. Sarotto d, D. Castelliti c a Instituto Tecnológico e Nuclear, Instituto Superior Técnico, Universidade Técnica de Lisboa, EN 10, 2686-953 Sacavém, Portugal b Ansaldo Nucleare S.p.A., c.so Perrone 25, 16161 Genova, Italy c SCK–CEN, Boeretang 200, Mol B-2400, Belgium d ENEA, via Martiri di Monte Sole 4, 40129 Bologna, Italy article info abstract Article history: The main objective of the Central Design Team (CDT) project is to establish an engineering design of a Received 19 October 2011 Fast Spectrum Transmutation Experimental Facility (FASTEF) that is the pilot plant of an experimental- Received in revised form 3 May 2012 scale of both an Accelerator Driven System (ADS) and a Lead Fast Reactor (LFR), based on the MYRRHA Accepted 3 May 2012 reactor concept, planned to be built during the next decade. The MYRRHA reactor concept is devoted Available online 16 August 2012 to be a multi-purpose irradiation facility aimed at demonstrating the efficient transmutation of long-lived and high radiotoxicity minor actinides, fission products and the associated technology. An important Keywords: issue regarding the reactor design of the MYRRHA/FASTEF experiment is the In-Vessel Fuel Storage Facil- Fast reactor ities (IVFSFs), both for fresh and spent fuel, as it might have an impact on the criticality of the overall sys- Neutronic design Decay heat tem that must be quantified. In this work, the neutronic analysis of the in-vessel fuel storage facility and its coupling with the critical core was performed, using the state of the art Monte Carlo program MCNPX 2.6.0 and ORIGEN 2.2 computer code system for calculating the buildup and decay heat of spent fuel. Sev- eral parameters were analyzed, like the criticality behavior (namely the Keff), the neutron fluxes and their variations, the fission power production and the radiation damage (the displacements per atom). Finally, also the heat power generated by the fission products decay in the spent fuel was assessed. Ó 2012 Elsevier Ltd. All rights reserved. 1. Introduction reactor, and at the present status of the design the core consists of MOX fuel pins with an active length of 60 cm and 33% of pluto- The aim of the CDT project [1,2] is to further develop the engi- nium content [1]. The choice of the MOX fuel was made taking into neering design of a first-step experimental device based on the account the large experience in the MOX production in Europe and resulting MYRRHA/XT-ADS facility of the FP6 EUROTRANS project its neutronic properties for a fast spectrum. that may serve both as a test-bed for transmutation and as a fast An important issue regarding the MYRRHA/FASTEF nuclear spectrum irradiation facility, operating as a subcritical [3] in Accel- reactor design is the In-Vessel Fuel Storage Facility (IVFSF), both erator Driven System (ADS) mode [4–6], and/or as a critical reactor. for fresh and spent fuel: in the current design, in order to avoid The CDT project defines also the new R&D activities needed to aid excessive delays between two operation cycles, it was decided to the detailed design and the construction of such facility that is place the IVFSF at the reactor periphery. This is a peculiar feature planned to be built during the next decade under the supervision of MYRRHA/FASTEF, since most of the present working reactors ex- of the SCKÁCEN Belgian Nuclear Research Centre and in collabora- ploit fuel storages not inserted in the reactor vessel. For this reason, tion with different national and international partners. MYRRHA/ it is crucial to assess the neutronic interaction between the central FASTEF, in the subcritical operational mode, has a proton accelera- core and the Fuel Assemblies (FAs) in the storage zones during the tor with energy of 600 MeV and a maximal current intensity of reactor operations, both in terms of criticality estimation and in 4 mA, coupled to a liquid lead bismuth eutectic (LBE) spallation terms of material damage to the overall structure. source. The spallation target is located at the center of the subcrit- As for the fresh fuel, it is important to know the neutron cou- ical core (see Fig. 1). The LBE coolant adopted in all the primary pling between the critical core and the storage vessels, the power system allows a fast neutron spectrum. The reactor is a pool-type generated in the storage vessels due to the fission induced by the neutron flux coming from the core and the structural damage of the vessel materials. Since a large amount of heat continue to be ⇑ Corresponding author. developed even after a reactor has been shutdown, it is essential E-mail address: [email protected] (S. Di Maria). to know the exact heat power released (i.e. by fission product 0196-8904/$ - see front matter Ó 2012 Elsevier Ltd. All rights reserved. http://dx.doi.org/10.1016/j.enconman.2012.05.001 S. Di Maria et al. / Energy Conversion and Management 64 (2012) 522–529 523 Fig. 1. Section view of the MYRRHA/FASTEF design operating both in critical and subcritical mode. decay) in this phase and remove it in a safe way, otherwise the fuel tions (IPSs) locations foreseen for irradiation applications [8].As elements may suffer damage (e.g. melting). shown in Fig. 2, the core design envisages a total of 151 positions In this paper the state of the art of the IVFSF for the MYRRHA/ and among them 37 penetrations (in the legend indicated as FASTEF design is described. Several parameters will be analyzed, Multi-Functional Channels) are planned to host indifferently IPS, as the criticality behavior (Keff), the fission power production and Control Rod (CR), Scram Rod (SR), FA and Dummy Assemblies the radiation damage, namely the displacement per atom (dpa). Fi- (DAs) [8]. nally, also the decay heat power of spent fuel generated for differ- The central penetration is used for the beam tube/target in sub- ent operational reactor periods will be assessed. critical mode and can be used for an IPS (or CR/SR) in critical mode. The calculations reported here were done by considering the It was decided to include four IVFSF, each one containing 76 reactor operating in critical mode. assembly positions that will be able to store a maximum of two full core loads. There are two IVFSF for each In-Vessel Fuel Handling 2. Materials and methods Machines (IVFHMs). The FA are inserted in the pipes of the IVFSF and remain there until their residual heat has sufficiently decayed. The core model used for our calculations is the LBE-filled box The positions of the IVFSF are also locations of fresh FA for refuel- dummy and the layout core design is shown in Fig. 2 [7]. The core ing and DA. The two IVFHM are positioned at opposite sides of the is composed of 68 fuel assemblies with a total power of 100 MW in core, each capable of accessing half of the core and half of the stor- the critical operational mode (in subcritical mode the operational age positions, thus allowing to minimize the diameter of the reac- power will be in the range 65–100 MW). The design goal of the tor vessel (see Fig. 3). The IVFSF are placed at the same level of the core is to achieve the maximum of the flexibility in the In-Pile Sec- core and far from it in order to avoid any neutron coupling [9]. Fig. 2. Left: Preliminary design of the 100 MW (68 FA) critical core; in the legend, the different positions for fuel and dummy assemblies, B4C rods and IPS are indicated. In particular the last two parenthetical positions indicate alternative positions (see text for details). Right: MCNPX critical core model used for calculations. 524 S. Di Maria et al. / Energy Conversion and Management 64 (2012) 522–529 Fig. 3. Overall top view of the MYRRHA/FASTEF Diaphragm design with all the parts included in the reactor vessel (see text for details). The core radius is 72 cm and the distance between the core cen- In our case Efiss is about 207 MeV/fission (taking into account ter and IVFSF center is 258 cm. For neutron calculations the both prompt and delayed energy release). Since the power of the MCNPX 2.6.0 Monte Carlo code [10] was used together with the core is known, we can derive the Normalization Factor (NF): JEFF 3.1 nuclear data libraries [11].InFig. 4 the MCNPX MYR- PðWÞ RHA/FASTEF design is shown: in the MCNPX model, at this stage, NF ¼ ð1Þ À13 J MeV Nfiss only the four IVFSF, critical core and diaphragm (part of the reactor 1:6022 Â 10 MeV 207 fiss s that divides the hot and cold LBE zones), were taken into account to perform criticality calculations. where Nfiss is the number of fissions per second in the core. Know- It is not essential to include other components of the system, ing the partial Nfiss of the IVFSF, it is straightforward to obtain the such as the Primary Pump or the Primary Heat Exchanger, since power released in the storage vessel. they are situated far from the active core and they would not influ- Many radiation-induced effects may cause material damage, as ence the criticality.
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