Years Of Full Power Operation Research Articles

Inverted fuel geometry implementation for a fast reactor: Potential improvements in neutronic and thermalhydraulic performance

This paper presents a detailed procedure for implementing the inverted fuel geometry to a fast reactor to improve its safety system and economy. The study is starting from the fuel unit cell, fuel assembly, reactor core, and burnup analysis. A proposed multivariable graph (v, ΔP, TFmax–Dco, VF) introduced at the fuel unit cell level provides comprehensive thermalhydraulic and neutronic parameters in a single graph, allowing for an efficient optimization process. The fuel unit cell study reveals that the inverted fuel design has a higher fuel volume fraction and lower core pressure drop than conventional pin-typed fuel. This is beneficial for the reactor economy and enhances the reliability of the safety system. With the inverted fuel design, the primary loop can save pumping power by up to 40 % and provides an excess driving force for natural circulation. The male–female axial grid structure separating the fuel assembly potentially eliminates coolant flow path restriction and fretting issues. The core is named the Inverted Core Fast Reactor (IC-FR), an LBE-cooled fast SMR designed to generate 60 MWth for at least 40 years of full-power operation without refueling and fuel shuffling. IC-FR is a transportable reactor and has load following capability that can be deployed for many applications, including marine and land-based applications, and stand alone or mixing power grid. The burnup study of IC-FR reveals that the balance of neutron leakage and fissile inventory yields a small reactivity swing (<1$) for 40 years. This study extensively utilizes Monte Carlo (MC) code MCS for neutronic calculation. Owing to the high computational expense of MC code, the approaches to optimize the MC usage are also presented.

Impact of fusion reactor neutronics modeling for transmutation and thermal feedback *

Fusion neutronics calculations provide important metrics pertinent to fusion device operations, such as tritium breeding ratios (TBRs) and data on heat deposition, material activation, and damage. Because of the high computational burden required to generate a high-fidelity Monte Carlo simulation of a 3D fusion device, various assumptions are made to reduce computational time by simplifying the reactor model or the calculation iteration. This paper explores the impact of fusion neutronics metrics such as the TBR and decay heat of structural materials based on assumptions of material composition in the fusion reactor and temperature modeling of materials. Results show that for compact tokamaks with high power and long operational cycles, the transmutation of structural materials is significant enough to cause a substantial change in the flux spectrum and decrease the TBR by 1.68% after 2 years of full power operation. Additionally, assuming a constant temperature and material density can impact the TBR calculations up to 3%.

Evaluation of neutron irradiation-induced displacement damage in heat pipe reactor

• Neutron irradiation-induced displacement damage in a 5 MW HPR is evaluated. • 91 maximum DPA is obtained for the SS316 monolith after a reactivity-determined cycle length. • The more than 4% swelling of the monolith leads the refueling to be difficult. • The maximum neutron irradiation dose in the Al 2 O 3 reflector is beyond available experimental data. • Two factors are proposed to convert the fast neutron fluences into DPA for Al 2 O 3 . The present work evaluates the neutron irradiation-induced displacement damage in a 5 MWth Heat Pipe Reactor (HPR) based on the latest released ENDF/B-VIII.0 nuclear reaction data library. The evaluations are performed for the SS316 monolith in the reactor core and the inner surface of the radial Al 2 O 3 reflector, the two individual parts suffering the highest neutron irradiation. At the end of the fuel cycle length, i.e., after 90 years of full-power operation, the global maximum neutron irradiation damage in the SS316 monolith is about 91 Displacements per Atom (DPA). The corresponding more than 4% volume swelling makes the refueling difficult, even though the allowed 200 DPA in other fast neutron reactors implies two potential cycles for the HPR by considering the mechanical behaviors of SS316. In the Al 2 O 3 reflector, the maximum neutron irradiation dose ( Φ E > 0.1 M e V = 3.3 × 10 22 c m - 2 and Φ E > 1 M e V = 6.4 × 10 21 c m - 2 fast neutron fluence) is beyond the available observed data. Further experiments should be performed to study the behavior of Al 2 O 3 in the HPR reflector after more than 63 years of full-power operation. For this purpose, two factors are proposed to convert the fast neutron fluences into DPA with certain specific values for the threshold displacement energy of Al and O in Al 2 O 3 because they are still not well recognized.

Neutronics Analysis of Small Compact Prismatic Nuclear Reactors for Space Crafts

Due to the advantages of small volume, light weight, and long-time running, nuclear reactor can provide an ideal energy source for space crafts. In this paper, two small compact prismatic nuclear reactors with different core block materials are presented, which have a thermal power of 5 MW for 10 years of equivalent full power operation. These two reactors use Mo-14%Re alloy or nuclear grade graphite IG110 as core block material, loaded with 50% and 39.5% enriched uranium nitride (UN) fuel and cooled by helium, whose inlet/outlet temperature of the reactor and operational pressure are 850/1300 K and 2 MPa, respectively. High temperature helium flowing out of the reactor can be used as the working medium for closed Brayton cycle power conversion with high efficiency (more than 20%). Neutronics analyses of reactors for the preliminary design in this paper are performed using reactor Monte Carlo (RMC) code developed by Tsinghua University. Both the reactors have enough initial excess reactivity to ensure 10 years of full power operation without refueling, have safety margin for reactor shutdown with one control drum failed, and remain subcritical in the submersion accident. Finally, the two reactors are compared in aspect of the 235U mass and the total reactor mass.

The Energy Multiplier Module (EM2): Status of Conceptual Design

The Energy Multiplier Module (EM2) is a helium-cooled fast reactor with a core outlet temperature of 850°C. It is designed as a modular, grid-capable power source with a net unit output of 265 MWe. The reactor employs a convert-and-burn core design that converts fertile isotopes to fissile and burns them in situ over a 30-year core life. The reactor is sited in a below-grade sealed containment. It uses passive safety methods for heat removal and reactivity control to protect the integrity of the fuel, reactor vessel, and containment. The plant also incorporates a below-grade, passively cooled spent fuel storage facility with capacity for 60 years of full-power operation. EM2 employs a direct closed-cycle gas turbine power conversion unit (PCU) with an organic Rankine bottoming cycle for 53% net power conversion efficiency assuming evaporative cooling. The high-power conversion efficiency and long-burn fuel cycle reduce the electricity cost by 35% when compared with the conventional light water reactor.The conceptual design has been conducted for the EM2 plant with focus on the reactor, fuel, and safety system designs. A detailed model of the passive direct reactor auxiliary cooling system was created to demonstrate functionality for selected design-basis accidents. The bench-scale fuel development campaign demonstrated high-quality uranium carbide pellet fabrication as well as β-SiC composite cladding and SiC-joining technologies. Irradiation tests of reactor materials are also being conducted. The PCU variable-speed generator mechanical design was validated with operational testing of its novel rotor at speeds >13 000 rpm. The design of the turbo-compressor with active magnetic bearings continues. A large cost database and financial model have been constructed for use as a key driver for the design to be economically competitive with competing generating technologies after 2030.

Measurement of mechanical properties of a reactor operated Zr–2.5Nb pressure tube using an in situ cyclic ball indentation system

Periodic measurement of mechanical properties of pressure tubes of Indian Pressurised Heavy Water Reactors is required for assessment of their fitness for continued operation. Removal of pressure tube from the core for preparation of specimens to test for mechanical properties in laboratories consumes large amounts of radiation and hence is to be avoided as far as possible. In the field of in situ estimation of properties of materials, cyclic ball indentation is an emerging technique. Presently, commercial systems are available for doing indentation test either on outside surface of a component at site or on a test piece in a laboratory. However, these systems cannot be used inside a pressure tube for carrying out ball indentation trials under in situ condition. Considering this, a remotely operable hydraulic In situProperty Measurement System (IProMS) based on cyclic ball indentation technique has been designed and developed in house. The tool head of IProMS can be located inside a pressure tube at any axial location under in situ condition and the properties can be estimated from an analysis of the data on load and depth of indentation, recorded during the test. In order to qualify the system, a number of experimental trials have been conducted on spool pieces and specimens prepared from Zr–2.5Nb pressure tube having different mechanical properties. Based on the encouraging results obtained from the qualification trials, IProMS has been used inside a reactor operated pressure tube which has been removed from Kakrapar Atomic Power Station-2 after 12.76 Full Power Years of operation for post irradiation examination studies at Bhabha Atomic Research Centre. This paper highlights the basic theory of measurement, results from the qualification trials on Zr–2.5Nb material having different mechanical properties and reactor operated irradiated pressure tube.

Predicting tube repair at French nuclear steam generators using statistical modeling

Electricité de France (EDF) currently operates a total of 58 Nuclear Pressurized Water Reactors (PWR) which are composed of 34 units of 900MWe, 20 units of 1300MWe and 4 units of 1450MWe. This report provides an overall status of SG tube bundles on the 1300MWe units. These units are 4 loop reactors using the AREVA 68/19 type SG model which are equipped either with Alloy 600 thermally treated (TT) tubes or Alloy 690 TT tubes. As of 2011, the effective full power years of operation (EFPY) ranges from 13 to 20 and during this time, the main degradation mechanisms observed on SG tubes are primary water stress corrosion cracking (PWSCC) and wear at anti-vibration bars (AVB) level.Statistical models have been developed for each type of degradation in order to predict the growth rate and number of affected tubes. Additional plugging is also performed to prevent other degradations such as tube wear due to foreign objects or high-cycle flow-induced fatigue. The contribution of these degradation mechanisms on the rate of tube plugging is described.The results from the statistical models are then used in predicting the long-term life of the steam generators and therefore providing a useful tool toward their effective life management and possible replacement.

Parameter study of the LIFE engine nuclear design

LLNL is developing the nuclear fusion based Laser Inertial Fusion Energy (LIFE) power plant concept. The baseline design uses a depleted uranium (DU) fission fuel blanket with a flowing molten salt coolant (flibe) that also breeds the tritium needed to sustain the fusion energy source. Indirect drive targets, similar to those that will be demonstrated on the National Ignition Facility (NIF), are ignited at ∼13 Hz providing a 500 MW fusion source. The DU is in the form of a uranium oxycarbide kernel in modified TRISO-like fuel particles distributed in a carbon matrix forming 2-cm-diameter pebbles. The thermal power is held at 2000 MW by continuously varying the 6Li enrichment in the coolants. There are many options to be considered in the engine design including target yield, U-to-C ratio in the fuel, fission blanket thickness, etc. Here we report results of design variations and compare them in terms of various figures of merit such as time to reach a desired burnup, full-power years of operation, time and maximum burnup at power ramp-down and the overall balance of plant utilization.

Power flattening options for the ENHS (encapsulated nuclear heat source) core

The feasibility of power flattening while maintaining a nearly constant k eff over the core life is assessed for the Encapsulated Nuclear Heat Source (ENHS). A couple of approaches are considered — using different fuel dimensions and using different enrichment levels across the core. Three new cores with flattened power distribution are successfully designed: Design-I uses different fuel rod diameters but uniform fuel composition; Design-II uses different fuel enrichment in the radial direction but uniform fuel rod dimensions; Design-III is similar to Design-II but uses enrichment splitting also in the axial direction. Relative to the reference ENHS core, the BOL peak-to-average channel power ratio is reduced from 1.50 to 1.15, 1.22 and 1.15 and the average discharge burnup increases by 8.5%, 27.9% and 41.2% for, respectively, Design-I, -II and -III. The corresponding burnup reactivity swings over 20 years of full power operation are 0.37%, 0.52% and 0.60% relative to 0.22% of the reference design. Design-II and -III have a negative coolant expansion reactivity defect while in the reference design this defect is positive. The radial power flattening increases the reactivity worth of the peripheral absorbers of the three new designs while the central absorber reactivity worth is reduced but their sum is nearly maintained. The newly designed cores have slightly more positive coolant void reactivity worth than the reference ENHS core.

The encapsulated nuclear heat source reactor for low-waste proliferation-resistant nuclear energy

The encapsulated nuclear heat source (ENHS) is a new Pb-Bi cooled modular reactor concept that features a combination of the following useful features that may make nuclear energy more attractive: (1) 20 years of full power operation without refueling. (2) Nearly constant fissile fuel contents and k eff. (3) No on-site refueling and fueling hardware. (4) The ENHS modules are factory manufactured and transported already fueled to the site. (5) No access to neutrons. (6) No mechanical connections between the ENHS module and the energy conversion plant (The ENHS module has the function of a nuclear battery — with 20 years of full power operation at 125 MW th). (7) At end of life, the ENHS module serves as a spent fuel storage cask and, later, as a spent fuel shipping cask. That is, the fuel is locked inside the ENHS from “cradle to grave”. (8) 100% natural circulation resulting in passive load following capability and autonomous control. This combination of features offers a highly safe nuclear energy system that is characterized by low waste, high proliferation resistance and high uranium utilization. The low waste and high uranium ore utilization are achieved by recycling the Pu and MA many times using a proliferation-resistant dry process; only fission products are to be extracted between cycles. Spent LWR fuel can provide for the HM make-up. The high level of proliferation resistance is obtained by restricting access to the fuel and neutrons and by eliminating the economic incentive of the client country to invest in sensitive technologies or infrastructure that can be used for clandestine production of strategic nuclear materials.

Activation Analysis for the D-3He Reactor ARIES-III

Activation analysis has been performed for the D-3He fueled reactor ARIES-III. The activity, decay heat and biological hazard potential (BHP) have been calculated for the low activation steel (modified HT-9) first wall and shield as a function of time following the reactor shutdown. The total activity produced in the reactor at shutdown is 1549 MCi. The total activity produced in the reactor organic coolant following 30 full power years of operation without reprocessing is 458 Ci. The modified HT-9 shield qualifies for shallow land burial as Class A low level waste. The biological dose rate after shutdown at the back of the outboard shield is too high to allow hands-on maintenance. Burning all the tritium in the plasma chamber results in increasing the radioactivity generated in ARIES-III by 65% to 85% at different times following the reactor shutdown.

Materials testing in the ICF test facility SIRIUS-M

The symmetric illumination laser-driven SIRIUS-M test facility provides materials testing in relevant ICF conditions. The test module is placed 2 m away from the target to achieve a goal neutron wall loading of 2 MW/m2. The 2 mm thick graphite liner reduces the peak dpa rate in the module by only 2.4%. Using a lead reflector results in 50% more damage in the test module compared to a stainless steel reflector. Two circular test modules are used in SIRIUS-M. Each module fits between three beam ports. About 1 MW of nuclear heating is removed by the helium coolant from each module. The peak iron dpa rate is 24 dpa/FPY yielding an accumulated damage of 120 dpa after 5 full power years of operation. A total volume-integrated figure of merit of 14200 dpa — l can be achieved. The test matrix and testing schedule are described. It is possible to perform all tests needed for the ICF Demo in the two SIRIUS-M test modules.

Materials testing in TASKA-M

The materials test module design for TASKA-M is presented. A representative 23000 specimen test matrix that utilizes two-thirds of the volume in a single module is described. The remaining test volume can be utilized for high heat flux or surface effects testing as well as dynamic in-situ or component testing. A peak dpa value of 78 can be achieved at the end of the 7.8 full power years of operation. The cumulative volume integrated damage level in the test modules is 4120 dpa·ℓ. TASKA-M, therefore, is most useful for interactive component testing and large specimen qualification testing.

Materials testing in TASKA-M

The materials test module design for TASKA-M is presented. A representative 23000 specimen test matrix that utilizes two-thirds of the volume in a single module is described. The remaining test volume can be utilized for high heat flux or surface effects testing as well as dynamic in-situ or component testing. A peak dpa value of 78 can be achieved at the end of the 7.8 full power years of operation. The cumulative volume integrated damage level in the test modules is 4120 dpa·ℓ. TASKA-M, therefore, is most useful for interactive component testing and large specimen qualification testing.