• Nuclear Engineering and Technology
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• v.50 no.7
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2733-2742
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• 2020

This paper presents an assessment of applicability of the multigroup cross sections generated with Monte Carlo tools to the fast reactor analysis based on transport calculations. 33-group cross section sets were generated for simple one- (1-D) and two-dimensional (2-D) sodium-cooled fast reactor problems using the SERPENT code and applied to deterministic steady-state and depletion calculations. Relative to the reference continuous-energy SERPENT results, with the transport corrected P₀ scattering cross section, the k-eff value was overestimated by 506 and 588 pcm for 1-D and 2-D problems, respectively, since anisotropic scattering is important in fast reactors. When the scattering order was increased to P₅, the 1-D and 2-D problem errors were increased to 577 and 643 pcm, respectively. A sensitivity and uncertainty analysis with the PERSENT code indicated that these large k-eff errors cannot be attributed to the statistical uncertainties of cross sections and they are likely due to the approximate anisotropic scattering matrices determined by scalar flux weighting. The anisotropic scattering cross sections were alternatively generated using the MC²-3 code and merged with the SERPENT cross sections. The mixed cross section set consistently reduced the errors in k-eff, assembly powers, and nuclide densities. For example, in the 2-D calculation with P3 scattering order, the k-eff error was reduced from 634 pcm to -223 pcm. The maximum error in assembly power was reduced from 2.8% to 0.8% and the RMS error was reduced from 1.4% to 0.4%. The maximum error in the nuclide densities at the end of 12-month depletion that occurred in ²³⁷Np was reduced from 3.4% to 1.5%. The errors of the other nuclides are also reduced consistently, for example, from 1.1% to 0.1% for ²³⁵U, from 2.2% to 0.7% for ²³⁸Pu, and from 1.6% to 0.2% for ²⁴¹Pu. These results indicate that the scalar flux weighted anisotropic scattering cross sections of SERPENT may not be adequate for application to fast reactors where anisotropic scattering is important.

• Nuclear Engineering and Technology
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• v.41 no.8
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• pp.1115-1126
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• 2009

As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.520-545
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• 2010

This paper presents the neutronics modeling capabilities of the fast reactor simulation system SHARP, which ANL is developing as part of the U.S. DOE's NEAMS program. We discuss the three transport solvers (PN2ND, SN2ND, and MOCFE) implemented in the UNIC code along with the multigroup cross section generation code $MC^2$-3. We describe the solution methods and modeling capabilities, and discuss the improvement needs for each solver, focusing on massively parallel computation. We present the performance test results against various benchmark problems and ZPR-6 and ZPPR critical experiments. We also discuss weak and strong scalability results for the SN2ND solver on the ZPR-6 critical assembly benchmarks.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3517-3527
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• 2021

New 80-MW (electric) ultra-long-life sodium cooled fast reactor core having inherent safety characteristics is designed with heterogeneous fuel assemblies comprised of driver and blanket fuel rods. Several options using upper sodium plenum and SSFZ (Special Sodium Flowing Zone) for reducing sodium void reactivity are neutronically analyzed in this core concept in order to improve the inherent safety of the core. The SSFZ allowing the coolant flow from the peripheral fuel assemblies increases the neutron leakage under coolant expansion or voiding. The Monte Carlo calculations were used to design the cores and analyze their physics characteristics with heterogeneous models. The results of the design and analyses show that the final core design option has a small burnup reactivity swing of 618 pcm over ~54 EFPYs cycle length and a very small sodium void worth of ~35pcm at EOC (End of Cycle), which leads to the satisfaction of all the conditions for inherent safety with large margin based on the quasi-static reactivity balance analysis under ATWS (Anticipated Transient Without Scram).

• Nuclear Engineering and Technology
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• v.55 no.8
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• pp.3071-3079
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• 2023

Helium production in the nuclear fuel matrix during irradiation plays a critical role in the design and performance of Gen-IV reactor fuel, as it represents a life-limiting factor for the operation of fuel pins. In this work, a surrogate model for the helium production rate in fast reactor MOX fuels is developed, targeting its inclusion in engineering tools such as fuel performance codes. This surrogate model is based on synthetic datasets obtained via the SCIANTIX burnup module. Such datasets are generated using Latin hypercube sampling to cover the range of input parameters (e.g., fuel initial composition, fission rate density, and irradiation time) and exploiting the low computation requirement of the burnup module itself. The surrogate model is verified against the SCIANTIX burnup module results for helium production with satisfactory performance.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.253-259
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• 2015

The fast neutron reactor BOR-60 is one of the key experimental facilities worldwide to perform large-scale tests of fuel, absorbing, and structural materials for advanced reactors. The BOR-60 reactor was put into operation in December 1969, and by the end of 2014 it had been operating on power for ~265,000 hours. BOR-60 still demonstrates potential capabilities to extend the lifetime of sodium-cooled fast reactors. The BOR-60 lifetime should have expired at the end of 2014. Over the past few years, a great scope of work has been performed to justify the possibility of extending its lifetime. The work included inspection of the equipment conditions, calculations and experimental research on operating parameters and the conditions of nonremovable components, investigation of the structural material samples after their long-term operation under irradiation, etc. Based on the results of the work performed, the residual lifetime was evaluated and the reactor operator made a decision to extend the lifetime period of the BOR-60 reactor. After considering both a set of documents about the reactor conditions and the positive decision of independent experts, the Regulatory Authority of the Russian Federation extended the BOR-60 operating license up to 2020.

• Journal of the Earthquake Engineering Society of Korea
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• v.11 no.3 s.55
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• pp.73-86
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• 2007

In this paper, a simple seismic analysis model of the KALIMER-600 sodium-cooled fast reactor selected to be the candidate of the GEN-IV reactor is developed. By using this model, the seismic time history analysis is carried out to investigate the feasibilities of a seismic isolation design. The developed simple seismic analysis model includes the reactor building, reactor system,, IHTS piping system, steam generator, and seismic isolators. The dynamic characteristics of the simple seismic model are verified with the detailed 3-dimensional finite element analysis for each part of the KALIMER-600 system. By using the developed simple seismic model, the seismic time history analyses for both cases of a seismic isolation and non-isolation design are performed for the artificial time history of a SSE (Safe Shutdown Earthquake) 0.3g. From the comparison of the calculated floor response spectrum, it is verified that the seismically isolated KALIMER-600 reactor building shows a great performance of a seismic isolation and assures a seismic integrity.

• Journal of Korea Society of Waste Management
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• v.34 no.5
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• pp.518-527
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• 2017

The development of renewable energy is currently strongly required to address environmental problems such as global warming. In particular, biomass is highlighted due to its advantages. When using biomass as an energy source, the conversion process is essential. Fast pyrolysis, which is a thermochemical conversion method, is a known method of producing bio-oil. Therefore, various studies were conducted with fast pyrolysis. Most studies were conducted under a lab-scale process. Hence, scaling up is required for commercialization. However, it is difficult to find studies that address the process analysis, even though this is essential for developing a scaled-up plant. Hence, the present study carries out the process analysis of biomass pyrolysis. The fast pyrolysis system includes a biomass feeder, fast pyrolyzer, cyclone, condenser, and electrostatic precipitator (ESP). A two-stage, semi-global reaction mechanism was applied to simulate the fast pyrolysis reaction and a circulating fluidized bed reactor was selected as the fast pyrolyzer. All the equipment in the process was modeled based on heat and mass balance equations. In this study, process analysis was conducted with various reaction temperatures and residence times. The two-stage, semi-global reaction mechanism for circulating fluidized-bed reactor can be applied to simulate a scaled-up plant.

• Nuclear Engineering and Technology
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• v.47 no.3
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• pp.227-239
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• 2015

The thermal, mechanical, and neutronic performance of the metal alloy fast reactor fuel design complements the safety advantages of the liquid metal cooling and the pool-type primary system. Together, these features provide large safety margins in both normal operating modes and for a wide range of postulated accidents. In particular, they maximize the measures of safety associated with inherent reactor response to unprotected, doublefault accidents, and to minimize risk to the public and plant investment. High thermal conductivity and high gap conductance play the most significant role in safety advantages of the metallic fuel, resulting in a flatter radial temperature profile within the pin and much lower normal operation and transient temperatures in comparison to oxide fuel. Despite the big difference in melting point, both oxide and metal fuels have a relatively similar margin to melting during postulated accidents. When the metal fuel cladding fails, it typically occurs below the coolant boiling point and the damaged fuel pins remain coolable. Metal fuel is compatible with sodium coolant, eliminating the potential of energetic fuel-coolant reactions and flow blockages. All these, and the low retained heat leading to a longer grace period for operator action, are significant contributing factors to the inherently benign response of metallic fuel to postulated accidents. This paper summarizes the past analytical and experimental results obtained in past sodium-cooled fast reactor safety programs in the United States, and presents an overview of fuel safety performance as observed in laboratory and in-pile tests.