A Finnish District Heating Reactor: Neutronics Design and Fuel Cycle Simulations

2021 ◽  
Author(s):  
Jaakko Leppänen ◽  
Ville Valtavirta ◽  
Riku Tuominen ◽  
Antti Rintala ◽  
Unna Lauranto
Keyword(s):  
Fuel Cycle ◽  
Natural Circulation ◽  
Reactor Core ◽  
District Heating ◽  
Research Centre ◽  
Computational Framework ◽  
Conceptual Design Phase ◽  
Burnable Absorber ◽  
Control Rods ◽  
Type Fuel

Abstract The development of a small PWR for district heating applications has been started at VTT Technical Research Centre of Finland, and the pre-conceptual design phase was completed by the end of year 2020. The heating plant consists of one or multiple 50 MW reactor modules, operating on natural circulation at around 120°C temperature. This paper presents the neutronics design and fuel cycle simulations carried out using VTT’s Kraken computational framework. The reactor is operated without soluble boron, which together with low operating temperature and pressure brings certain challenges to the use of control rods and burnable absorber. The reactor core is loaded with 37 truncated AP1000-type fuel assemblies with 2.0–3.0% fuel enrichment and erbium burnable absorber. The resulting cycle length is around 900 days. The results show that the criteria set for stability, reactivity control and thermal margins are fulfilled. More importantly, it is concluded that the new Kraken framework is a viable tool for the core design task.

Possibility of Different Fuel Cycles Usage in GT-MHR

2002 ◽  
Author(s):  
N. Kodochigov ◽  
Yu. Sukharev ◽  
E. Marova ◽  
N. Ponomarev-Stepnoy ◽  
E. Glushkov ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Block Design ◽  
Average Power ◽  
Reactor Core ◽  
Fuel Burnup ◽  
Fuel Load ◽  
Fuel Cycles ◽  
Partial Cycle ◽  
Control Rods ◽  
Neutronic Characteristics

The GT-MHR reactor core is characterized by flexibility of neutronic characteristics at the given average power density and fixed geometrical dimensions of reactor core. Such flexibility makes it possible to start the reactor operation with one fuel cycle, and then to turn to another type of core fuel load without changes of main reactor elements: fuel block design, core and reflector size, control rod number etc. Preliminary analysis reindicates the commercial viability of the GT-MHR, part of which is due to the ability to accommodate different fuel types and cycles. This paper presents the results of studies of the neutronic characteristics of reactor cores using different fuel (low- and high-enriched uranium, MOX fuel). Comparison of different fuel cycles is carried out for a three-batch refueling option with respect to following characteristics: discharged fuel burnup, reactivity change during one partial cycle of fuel burnup, consumption of fissile isotopes per unit of produced energy, power distribution, reactivity effects, control rods worth. It is shown, that the considered options of fuel loads provide the three-year fuel campaign (with accounting of capacity factor ∼ 0,8) without change of core design, number and design of control rods at transition from the one fuel type to another.

Development of Safety-Enhanced Fast Reactor by Using Minor Actinide Bearing Internal Blanket

2020 ◽  
Author(s):  
Sho Fuchita ◽  
Satoshi Takeda ◽  
Koji Fujimura ◽  
Toshikazu Takeda ◽  
Kazuhiro Fujimata
Keyword(s):  
Fast Reactor ◽  
Inner Core ◽  
Reactor Core ◽  
Minor Actinide ◽  
Core Damage ◽  
Burnable Absorber ◽  
Sensitivity Studies ◽  
Core Height ◽  
Gas Expansion ◽  
Control Rods

Abstract For a 750MWe sodium-cooled fast reactor core using MOX fuel, safety-enhancement measures have been studied to reduce the risk of core damage under unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents. As passive measures the followings are considered: 1) adoption of the axial heterogeneous core configuration with sodium plenum and Gas Expansion Modules (GEMs) to lower sodium void reactivity for ULOF, and 2) addition of minor actinides (MAs) as burnable absorber and fertile nuclides to the internal blanket in the inner core to reduce burnup reactivity for UTOP. In this study, configurations of the safety-enhanced core were optimized based on sensitivity studies as follows. Firstly, effects of 1) above on the sodium void reactivity were evaluated by changing the inner core height, B-10 content of the upper shield, GEMs, and standby position of the backup control rods, which are the dominant factors of core behavior in the event of ULOF. Secondly, the effects of 2) above on the burnup reactivity were evaluated by changing the MA content in the internal blanket and the burnup period, which are the dominant factors of UTOP. Finally, by utilizing sensitivity analysis results, the safety-enhanced core which satisfies the provisional design goals has been developed. This core has negative sodium void reactivity and burnup reactivity less than 1 $.

Nuclear Analysis During Modification of NPP Core

2010 ◽  
Author(s):  
Andrius Slavickas
Keyword(s):  
Reactor Core ◽  
Reactor Power ◽  
Physical Parameters ◽  
Reactor Operation ◽  
Reactor Control ◽  
The Core ◽  
Burnable Absorber ◽  
Reactivity Coefficients ◽  
Control Rods ◽  
Neutron Activity

Reactor power and neutron activity control is the main key for safe reactor operation. Reactivity coefficients and effects are main measures to estimate reactor control and safety. These characteristics outline reactors behavior during usually exploitation and accident events. Reactivity coefficients and effects quantify the effect, which various parameters (e.g. fuel and graphite temperatures, amount of steam) have for the core neutron activity. Many modifications of RBMK-1500 reactor cores in Ignalina NPP were made during their lifetime. Reactor core modifications like load of higher enriched fuel with burnable absorber and new design control rods affected reactivity coefficients and effects. Neutron-physical parameters calculations of reactor core states with variant fuel loads and new design control rods were performed using QUABOC/CUBBOC-HYCA software. The changes of reactivity coefficients and effects were quantified in this paper.

A Finnish District Heating Reactor: Thermal-Hydraulic Design and Transient Analyses

2021 ◽  
Author(s):  
Rebekka Komu ◽  
Seppo Hillberg ◽  
Ville Hovi ◽  
Jaakko Leppänen ◽  
Joona Leskinen
Keyword(s):  
Low Temperature ◽  
Natural Circulation ◽  
District Heating ◽  
Simulation Software ◽  
Research Centre ◽  
Decay Heat ◽  
The Core ◽  
Hydraulic Design ◽  
Station Blackout ◽  
Temperature And Pressure

Abstract Development of a small district heating reactor was started at VTT Technical Research Centre of Finland. The concept features a 50 MW reactor that operates at low temperature and pressure. Traditional LWR technology, passive safety functions and natural circulation are combined in the integrated design. This paper presents the thermal-hydraulic design and transient analyses done with Apros simulation software. The studied cases include station blackout with reactor trip and as an ATWS scenario, and small break LOCA in the lower downcomer. During the station blackout transients, both temperature and pressure remained at safe levels. The innovative containment design functioned as planned and was capable of efficient decay heat removal. The small break LOCA ceased the natural circulation, but the core was not uncovered at any point and the core temperatures remained low. The results from the thermal-hydraulic analyses are promising and show that the reactor design is capable of producing low temperature heat to the district heating network. The analyzed transients posed no risk to reactor safety, and the passive containment function was capable of removing decay heat efficiently. These preliminary analyses give valuable insight to the design work in the future.

Analysis of Flow Blockage of a Single Fuel Assembly in the JRR-3 20MW Research Reactor

2018 ◽  
Author(s):  
Yuchuan Guo ◽  
Guanbo Wang ◽  
Dazhi Qian ◽  
Heng Yu ◽  
Bo Hu
Keyword(s):  
Fuel Assembly ◽  
System Analysis ◽  
Research Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Nucleate Boiling ◽  
Regulation System ◽  
Lower Power ◽  
Emergency Shutdown ◽  
Flow Blockage

The case of flow blockage of a single fuel assembly in the JRR-3 20MW open-pool-type research reactor is investigated without taking into account the effect of the power regulation system. The coolant system and multi-channel reactor core are modeled in detail using thermal hydraulic system analysis code RELAP5/MOD3.4. MDNBR (Minimum Departure From Nucleate Boiling Ratio) and the maximum fuel central temperature are investigated to assess the integrity of fuels. The fuel plates in blocked assembly are not damaged until the blockage ratio exceeds 70%. In addition, the mitigative effect of the assumed 18 MW lower power emergency shutdown operation on the accident is also discussed qualitatively. Results indicate that although the assumed lower power emergency shutdown operation cannot avoid the most severe operating condition, it can obviously mitigate the consequences of the accident. The reactor eventually remains in the long-term safe state when natural circulation is established.

Analysis of Passive Residual Heat Removal System at Primary Side when Station Blackout Occurs

2014 ◽  
Vol 986-987 ◽  
pp. 231-234
Author(s):  
Jun Teng Liu ◽  
Qi Cai ◽  
Xia Xin Cao
Keyword(s):  
Nuclear Reactor ◽  
Natural Circulation ◽  
Reactor Core ◽  
Saturation Temperature ◽  
Heat Removal ◽  
Primary Circuit ◽  
Station Blackout ◽  
Heat Removal System ◽  
Removal System ◽  
Residual Heat

This paper regarded CNP1000 power plant system as the research object, which is the second-generation half Nuclear Reactor System in our country, and tried to set Westinghouse AP1000 passive residual heat removal system to the primary circuit of CNP1000. Then set up a simulation model based on RELAP5/MOD3.2 program to calculate and analyze the response and operating characteristic of passive residual heat removal system on assumption that Station Blackout occurs. The calculation has the following conclusions: natural circulation was quickly established after accident, which removes core residual heat effectively and keep the core safe. The residual heat can be quickly removed, and during this process the actual temperature was lower than saturation temperature in reactor core.

Cross-Comparison of One-Dimensional Thermal-Hydraulic Codes on Natural Circulation Analysis of NACIE Loop Test for Lead-Alloy Cooled Advanced Nuclear Energy Systems (LACANES)

2014 ◽  
Author(s):  
Yong-Hoon Shin ◽  
Il Soon Hwang ◽  
Massimiliano Polidori ◽  
Paride Meloni ◽  
Vincenzo Casamassima ◽  
...  
Keyword(s):  
Phase I ◽  
Phase Ii ◽  
Nuclear Energy ◽  
Natural Circulation ◽  
Energy Systems ◽  
Research Centre ◽  
Hydraulic Loss ◽  
Lead Alloy ◽  
Test Results ◽  
One Dimensional

As one of the Generation-IV reactor concepts, lead-alloy-cooled advanced nuclear energy systems (LACANES) have been studied worldwide in order to utilize the advantages of good heat transfer properties, neutron transparency and chemical inertness with air and water. Since the Fukushima accident, the passive safety aspect of the LACANES is increasingly emphasized due to outstanding natural circulation capability. To investigate the thermal-hydraulic capability of LBE, an international cooperation has been performed under OECD/NEA program, under the guidance of the Nuclear Science Committee by a task force named as Lead Alloy Cooled Advanced Nuclear Energy Systems (LACANES) since 2007. This international collaboration had dealt with computational benchmarking of isothermal LBE forced convection tests in the phase I, and the working group published a guideline for using one-dimensional system codes to simulate LBE forced circulation test results from HELIOS loop. The phase II was started after that, to give an additional guideline in the case of natural circulation. NACIE, one of benchmarking targets for the phase II which is a rectangular-shape loop located at ENEA-Brasimone Research Centre, Italy. NACIE test results were benchmarked by each participant using their one-dimensional thermal-hydraulic codes, and they are to follow the guideline from the LACANES phase I for regions where hydraulic loss occurs. Due to the selection of hydraulic loss coefficient relations by users, the cross-comparison results of international participants showed some discrepancies and the estimated mass flow rates had 13% of maximum error. Also, the future R&D areas are identified.

Assessment of RELAP/SCDAPSIM for Turbine Trip Transient in NuScale-SMR

2018 ◽  
Author(s):  
Katarzyna Skolik ◽  
Anuj Trivedi ◽  
Marina Perez-Ferragut ◽  
Chris Allison
Keyword(s):  
Natural Circulation ◽  
System Modeling ◽  
Thermal Power ◽  
Reactor Core ◽  
Operating Conditions ◽  
Severe Accident ◽  
Circulation System ◽  
Pressurized Water Reactor ◽  
Pressurized Water ◽  
Secondary Steam

The NuScale Small Modular Reactor (SMR) is an integrated Pressurized Water Reactor (iPWR) with the coolant flow based on the natural circulation. The reactor core consists of 37 fuel assemblies similar to those used in typical PWRs, but only half of their length to generate 160MW thermal power (50 MWe). Current study involves the development of a NuScale-SMR model based on its Design Certification Application (DCA) data (from NRC) using RELAP/SCDAPSIM. The turbine trip transient (TTT) was simulated and analysed. The objective was to assess this version of the code for natural circulation system modeling capabilities and also to verify the input model against the publicly available TTT results obtained using NRELAP5. This successful benchmark confirms the reliability of the thermal hydraulic model and allows authors to use it for further safety and severe accident analyses. The reactor core channels, pressurizer, riser and downcomer pipes as well as the secondary steam generator tubes and the containment were modeled with RELAP5 components. SCDAP core and control components were used for the fuel elements in the core. The final input deck achieved the steady state with the operating conditions comparable to those reported in the DCA. RELAP/SCDAPSIM predictions are found to be satisfactory and comparable to the reference study. It confirms the code code capabilities for natural circulation system transients.

scholarly journals Modeling minor actinide multiple recycling in a lead-cooled fast reactor to demonstrate a fuel cycle without long-lived nuclear waste

Nukleonika ◽  
2015 ◽  
Vol 60 (3) ◽  
pp. 581-590 ◽  
Author(s):  
Przemysław Stanisz ◽  
Jerzy Cetnar ◽  
Grażyna Domańska
Keyword(s):  
Nuclear Fuel ◽  
Nuclear Waste ◽  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Core ◽  
Nuclear Fuel Cycle ◽  
Minor Actinide ◽  
Closed Nuclear Fuel Cycle ◽  
European Collaboration ◽  
Multiple Recycling

Abstract The concept of closed nuclear fuel cycle seems to be the most promising options for the efficient usage of the nuclear energy resources. However, it can be implemented only in fast breeder reactors of the IVth generation, which are characterized by the fast neutron spectrum. The lead-cooled fast reactor (LFR) was defined and studied on the level of technical design in order to demonstrate its performance and reliability within the European collaboration on ELSY (European Lead-cooled System) and LEADER (Lead-cooled European Advanced Demonstration Reactor) projects. It has been demonstrated that LFR meets the requirements of the closed nuclear fuel cycle, where plutonium and minor actinides (MA) are recycled for reuse, thereby producing no MA waste. In this study, the most promising option was realized when entire Pu + MA material is fully recycled to produce a new batch of fuel without partitioning. This is the concept of a fuel cycle which asymptotically tends to the adiabatic equilibrium, where the concentrations of plutonium and MA at the beginning of the cycle are restored in the subsequent cycle in the combined process of fuel transmutation and cooling, removal of fission products (FPs), and admixture of depleted uranium. In this way, generation of nuclear waste containing radioactive plutonium and MA can be eliminated. The paper shows methodology applied to the LFR equilibrium fuel cycle assessment, which was developed for the Monte Carlo continuous energy burnup (MCB) code, equipped with enhanced modules for material processing and fuel handling. The numerical analysis of the reactor core concerns multiple recycling and recovery of long-lived nuclides and their influence on safety parameters. The paper also presents a general concept of the novel IVth generation breeder reactor with equilibrium fuel and its future role in the management of MA.

The Effect of MA Transmutation in the PWR on Fuel Cycle

2017 ◽  
Author(s):  
Haoyang Yu ◽  
Bin Liu ◽  
Wenxin Zhang ◽  
Jin Cai
Keyword(s):  
Fuel Cycle ◽  
Minor Actinide ◽  
Minor Actinides ◽  
Pressurized Water Reactor ◽  
Spent Fuel ◽  
Pressurized Water ◽  
Mcnp Program ◽  
And Control ◽  
Control Rods ◽  
Effective Multiplication

The minor actinides (MA) is important nuclides in the spent fuel which is bad for human ecological environment. Pressurized water reactor (PWR) is the main reactor type at commercial operation around world. It is important to find the appropriate loading patterns when introducing minor actinides to the PWR core. In this paper, we study the effect of MA transmutation in the PWR on fuel cycle. First, we use the MCNP program to simulate the model of PWR and the effective multiplication factor.Then,the MA is introduced into core in different ways and mass to simulate the effective multiplication factor. In conclusion,without considering chemical skim control and control rods, we change the thickness of the MA, until the keff closes to 1, We find that loading minor actinides to burnable poison rods for transmutation is an optimal minor actinide loading pattern.

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