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1. Introduction
In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the temperature distribution in the fuel pin bundle is of central interest. In particular, the use of lead or lead-bismuth eutectic (LBE) as coolant for the new generation fast reactors is one of the most promising choices. Due to the high density and high conductivity of lead or LBE, a detailed analysis of the thermo-fluid dynamic behavior of the heavy liquid metal (HLM) inside the sub-channels of a fuel rod bundle is necessary in order to support the front-end engineering design (FEED) of GEN. IV/ADS prototypes and demonstrators. In this frame, the synergy between numerical analysis by CFD and data coming from large experimental facilities seems to be crucial to assess the feasibility of the components. At ENEA-Brasimone R.C., large experimental facilities exist to study HLM free, forced and mixed convection in loops and pools: e.g. NACIE-UP is a large scale LBE loop for mixed convection experiments.
The present master thesis is devoted to the Computational Fluid Dynamic (CFD) analysis of Heavy Liquid Metal (HLM) cooled Fuel Bundles to be adopted in the Gen-IV nuclear reactors. The thesis was carried out in collaboration with the ENEA Brasimone research center, where large experimental facilities are operated to investigate HLM technology and thermal hydraulics. In particular liquid Lead or Lead-Bismuth Eutectic (LBE) is considered as working fluid. A brief overview is given on Lead technology for Gen-IV nuclear reactors.
1.1. Overview of the Technology
Among the promising reactor technologies being considered by the Generation IV International Forum (GIF) 0 and shown in Table 1-1 , the Lead-cooled Fast Reactor (LFR) was identified as a system with great potential to meet needs for both remote sites and central power stations. LFR promises in meeting the Generation IV goals of Sustainability, Economics, Safety and Reliability, basing either on the inherent features of lead as a coolant or on the specific engineered designs.
1.1.1. Sustainability
Because lead is a coolant with very low neutron absorption and energy moderation, it is possible to maintain a fast neutron flux even with large amount of coolant in the core. The fast neutron spectrum yields excess neutrons that can be utilized efficiently to manage a variety of fuel materials. Reactor designs can readily achieve a Conversion Ratio of 1, a long core life and a high fuel burn-up.
A fast neutron flux significantly reduces waste generation, Pu recycling in a closed cycle being the first condition recognized by Generation IV for waste minimization. The capability of the LFR systems to safely burn recycled minor actinides within the fuel will add to the attractiveness of the LFR and meet another important Generation IV condition 0 .
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Table 1-1 Brief overview of the innovative reactor concepts selected by GIF. Neutron spectrum Coolant Temperature [°C] Pressure* Fuel Size(s) [MWe] Uses Gas-cooled fast
reactors fast Helium 850 High U-238 + 1200
Electricity & Hydrogen
Lead-cooled
fast reactors fast
Lead /
LBE 480 – 800 Low U-238 +
20-180 ** 300-1200 600-1000 Electricity & Hydrogen
Molten salt fast
reactors fast
Fluoride
salts 700 – 800 Low UF in salt 1000
Electricity & Hydrogen Molten salt reactor-Advanced High-Temperature reactors thermal Fluoride salts 750 - 1000 UO2 particles in prism 1000-1500 Hydrogen Sodium-cooled
fast reactors fast Sodium 550 Low
U-238 & MOX 30-150 300-1500 1000-2000 Electricity Supercritical water-cooled fast reactors Thermal or fast Water 510 - 625 Very High UO2 300-700 1000-1500 Electricity Very high temperature gas reactors
thermal Helium 900 - 1000 High UO2 prism/pebbles 250-300 Electricity & Hydrogen * high = 7-15 Mpa + with some U-235 or Pu-239
** 'battery' model with long cassette core life (15-20 yr) or replaceable reactor module. 1.1.2. Economics
The cost advantages of the LFR are expected to include low capital cost, short construction duration and low production cost. Because of the favorable characteristics of molten lead, it will be possible to significantly simplify the LFR systems, and hence to reduce its overnight capital cost, which is a major cost factor for the competitive generation of nuclear electrical energy.
The use of in-vessel energy conversion equipment (i.e., Steam Generator Units) and the consequent elimination of the need for an Intermediate System increase the thermal efficiency provide competitive generation of electrical energy in the LFR.
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In the short term, to use available structural materials and to reduce the potential risk for investors in a nearly new technology, it is planned to operate at low core outlet temperature (around 500 °C) to produce electricity and low temperature process heat.
In the longer term, when new materials resistant to corrosion in lead will be available at industrial level, the very high boiling temperature of lead will also allow to produce process heat at much higher temperature for hydrogen generation or other outputs. The industrial risk is also reduced by a simple design and possibility of replacement of all components inside the reactor.
1.1.3. Safety and Reliability
Liquid lead offers excellent neutronic performances; it is chemically inert with air and water, and exhibits low vapor pressure with the advantage of allowing operation of the primary system at atmospheric pressure. A low dose to the operators can also be predicted, owing to its low vapor pressure, high capability of trapping fission products and high shielding of gamma radiation.
The void reactivity coefficient is negative for small size reactors and increases with the plant size until reaching positive values which nevertheless cannot be associated to any credible dangerous scenario because of the very high boiling temperatures of lead and Lead-Bismuth Eutectic (LBE) (1737°C and 1670°C respectively). Design provisions intended to eliminate the risk of steam or CO2 ingress into the core by design.
Due to the low moderating capability of lead it is possible to have relatively large spacing among the fuel rods, i.e. high pitch to diameter ratio p/d, with low pressure losses in spite of the high density of lead. In fact, the average velocity in the core sub channels is of the order of 1÷1.5 m/s and it is reduced with respect to the water reactors, leading to a similar dynamic pressure head. Lead allows a high level of natural circulation of the coolant with simplification of control and protection systems. Any leaked lead would solidify without significant chemical reactions affecting the operation or performance of surrounding equipment or structures. With lead as a coolant, fuel dispersion dominates over fuel compaction, reducing the risk of recriticality in the event of partial core melt. In fact lead, with its high density (close to that of oxide fuel and low-density metal fuel) and its natural convection flow, prevents fuel aggregation with subsequent formation of a secondary critical mass in the event of postulated fuel failure.
Lead or LBE coolants, however, have some safety and reliability concerns primarily related to possible corrosion of structural materials and the production of volatile and radioactive Po-210.
The experience gained with LBE cooled reactors in the Soviet nuclear alpha submarine propulsion program [2], however indicates that the technical problems can be overcome with adequate materials selection, chemical control, design and manufacture. The generation of
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210 is mainly an issue with LBE coolants; the lead cooled designs generate about four orders of magnitude lower levels of Po-210.
1.1.4. Physical Protection
The use of a coolant chemically compatible with air and water and operating at ambient pressure greatly enhances Physical Protection. There is reduced need for robust protection against the risk of catastrophic events, initiated by acts of sabotage and there is a little risk of fire propagation. There are no credible scenarios of significant containment pressurization.
1.2. HLM Gen-IV Demonstrator and irradiation facility Several reactor projects are currently under investigation in the community:
The ALFRED [3] reactor should be the Demonstrator for the HLM Gen-IV technology, and it should have the secondary system and electricity production (~300 MWth, ~100 MWe). It is based on some solutions studied in the FP7 project devoted to the conceptual design of the ELSY reactor [4]. Regarding Fuel Assembly, the main difference between the ELSY and ALFRED core is that ELSY adopted an open/square lattice, while ALFRED adopted a closed hexagonal wrapped element. Both FAs have spacer grids and do not adopt the classical solution of the wire-spacing of the sodium reactors.
The MYRRHA irradiation facility design is under Front End Engineering Design stage and is tied to several EC-FP7 projects (i.e. SEARCH, MAXSIMA and others). The machine will be built at SCK-CEN (Belgium) and will not produce electricity. MYRRHA will adopt LBE as coolant and relatively low temperatures in the core (max 400 °C). The Fuel Assembly will be wrapped wire-spaced as in sodium reactors.
1.2.1. The MYRRHA reactor
MYRRHA, a flexible fast spectrum research reactor, is conceived as an accelerator driven system (ADS), able to operate in sub-critical and critical modes. It contains a proton accelerator of 600 MeV, a spallation target and a multiplying medium with MOX fuel, cooled by liquid lead-bismuth (Pb-Bi).
The design characteristics of MYRRHA are determined by the international needs in terms of flexible fast spectrum irradiation capabilities and ADS demonstration and the targeted applications catalogue for this facility.
Table 1-2 MYRRHA specifications.
General characteristics MYRRHA
Core barrel inner diameter 1.480 mm
Reactor vessel, inner diameter 6.030 mm
Height (cover not included) 8.860 mm
Primary coolant LBE
LBE volume 150 m³
Secondary coolant Boiling water
Core inlet temperature 270 °C
Core average outlet temperature 400 °C
Maximum allowed bulk velocity 2.0 m/s
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As a direct consequence of the desired high flux levels, and hence high power density, a compact core is needed and therefore, the central hole in the core which houses the spallation target should be of limited dimensions (~ 10 cm).
SCK•CEN opted for liquid metal as coolant. Lead-bismuth eutectic (LBE) was selected due to its low melting temperature (124.5 °C), allowing the primary systems to function at rather low temperatures.
Inside the compact core geometry only an effective target diameter of about 88 mm is possible. With the given properties of the proton beam, this value leads to a beam current density of 65 µA/cm2. A windowless target design, i.e. without a physical separation between the accelerator beam line vacuum and the liquid target material, is proposed.
The sub-criticality level of around 0.95 has been considered as an appropriate level for a first of a kind medium-scale ADS.
MOX fast reactor fuel technology has been chosen due to the large experience in Europe and
Figure 1-1 Overall view of the MYRRHA reactor with its main
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in particular in Belgium. A maximum plutonium enrichment of 35 % was considered based on the available manufacturing and qualification experience by Belgonucleaire in the past. The neutronic configuration of the MYRRHA sub-critical core is similar to that of a classical LMFR. The core has 102 hexagonal positions with pitch ("centre-to-centre") of 87 mm. Three central positions are free and occupied by the spallation target. The target is surrounded by the active zone loaded with 45 (or more) fuel assemblies with wire-wrap spacers as in Figure 1-2. In their turn, they are surrounded by the reflector zone. The maximum core radius is 1000 mm, the core height determined by the assembly length is 1844 mm and the active core height is 600 mm. A typical MYRRHA configuration with keff of 0.95 can be achieved by using 45 fuel
assemblies with 30 % MOX. There are 19 positions accessible through the reactor lid capable to house experimental devices.
Within the reactor vessel, the core is held in place by support structure, including a core barrel and core support plate. The core itself consists of 151 positions, with 37 positions are available for multi-functional channels (MFC); see.
Dummy assemblies and mixed oxide (MOX) fuel assemblies are all installed from the bottom of the reactor vessel. Two kinds of dummy assemblies are envisaged: an internal ring surrounding the fissile zone made of eutectic lead bismuth (LBE) dummy to increase neutron reflection, and an external ring with the same structure as the fuel assemblies, but filled with YZrO pellets to shield the core barrel.
Seven central in-pile test sections (IPS), along with six control rods and three scram rods, are loaded from the top of the reactor vessel. The control rods are buoyancy-driven in LBE, while the scram rods are gravity-driven.
To profit from the thermal inertia provided by a large coolant volume, a pool-type system was choosen in which the components of the primary loop (pumps, heat exchangers, fuel handling tools, experimental rigs, etc.) are inserted from the top in penetrations in the cover. The loading of fuel assemblies is foreseen to be from underneath, which is not the classical approach of the sodium fast reactors. The reasons behind the approach are firstly to keep a large flexibility for the experimental devices loading from the top and secondly, from the safety point of view, the fact that all structures including the spallation module are in place before starting the core loading.
The pool vessel, which contains the MYRRHA core the internals, is located in an air-controlled containment environment. Furthermore, several factors lead to the decision to design both operation and maintenance (O&M) and In-Service Inspection & Repair (ISI&R) of MYRRHA with fully-remote handling systems.
1.2.2. Wire wrapped vs. grid spacer solution for the Fuel Assembly
The passage for the coolant flow between the fuel rods is maintained either by grid spacers or by thin wires wound around the rods. The prototype and demonstration plants in UK and
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Germany and the Fermi reactor in US used grid spacers; other plants use the wire-wrap spacer design. A steel wire of circular cross-section is welded at one axial extremity and spirally wound around the pin with a specified axial pitch.
Grid spacers, on the other hand, consist of a steel web structure anchored to the subassembly duct wall at specified axial levels. Less steel volume is occupied to provide a higher breeding ratio than wire wraps. In spite of this, wire wraps are preferred for several advantages. Firstly, they are easy to fabricate and less expensive. Further, mechanical vibration problems and reactivity oscillations are minimized by using wire wraps. This is because for every pitch of wire wrap, contact with adjacent cladding occurs at 6 axial locations.
To provide similar structural stability, grid spacers require several grids, resulting in excessive pressure drop. The wire-wrap design also enables better thermal mixing of coolant due to induction of lateral velocity components and increase in local turbulence level.
1.3. Plan of the thesis
The thesis will be structured in the following way:
Chapter 2 is devoted to a wide literature review on Fuel Assembly thermal-hydraulics and experiments, with a special attention to liquid metals as coolant (Sodium, Lead and LBE) and wire-wrap fuel bundle geometry and TH studies;
Chapter 3 is focused on the description of the NACIE-UP facility, the fuel pin bundle simulator (FPS) and its instrumentation;
Chapter 4 describes the CFD model developed for the NACIE-UP pin bundle simulator with the numerical models and methods adopted;
Chapter 5 is devoted to the ANSYS code assessment for this wire-wrapped geometry; Chapter 6 specifically describes all the sensitivity studies investigated, the test matrix case adopted for the pre-test analysis, the main results obtained and the highlights for the future experimental campaign;
