scholarly journals International Benchmark Activity in the Field of Sodium Fast Reactors

2021 ◽  
Author(s):  
Domenico De Luca ◽  
Simone Di Pasquale ◽  
Marco Cherubini ◽  
Alessandro Petruzzi ◽  
Gianni Bruna
Keyword(s):  
Experimental Data ◽  
Fast Reactor ◽  
Construction Projects ◽  
Fuel Cycle ◽  
Heat Removal ◽  
Fast Reactors ◽  
Benchmark Analysis ◽  
Start Up ◽  
International Projects ◽  
Computer Codes

Global interest in fast reactors has been growing since their inception in 1960 because they can provide efficient, safe, and sustainable energy. Their closed fuel cycle can support long-term nuclear power development as part of the world’s future energy mix and decrease the burden of nuclear waste. In addition to current fast reactors construction projects, several countries are engaged in intense R&D and innovation programs for the development of innovative, or Generation IV, fast reactor concepts. Within this framework, NINE is very actively participating in various Coordinated Research Projects (CRPs) organized by the IAEA, aimed at improving Member States’ fast reactor analytical simulation capabilities and international qualification through code-to-code comparison, as well as experimental validation on mock-up experiment results of codes currently employed in the field of fast reactors. The first CRP was focused on the benchmark analysis of Experimental Breeder Reactor II (EBR-II) Shutdown Heat Removal Test (SHRT-17), protected loss-of-flow transient, which ended in the 2017 with the publication of the IAEA-TECDOC-1819. In the framework of this project, the NINE Validation Process– developed in the framework of NEMM (NINE Evaluation Model Methodology) – has been proposed and adopted by most of the organizations to support the interpretation of the results calculated by the CRP participants and the understanding of the reasons for differences between the participants’ simulation results and the experimental data. A second project regards the CRP focused on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the Fast Flux Test Facility (FFTF), the Loss of Flow Without Scram (LOFWOS) Test #13, started in 2018. A detailed nodalization has been developed by NINE following its nodalization techniques and the NINE validation procedure has been adopted to validate the Simulation Model (SM) against the experimental data of the selected test. The third activity deals with the neutronics benchmark of China Experimental Fast Reactor (CEFR) Start-Up Tests, a CRP proposed by the China Institute of Atomic Energy (CIAE) launched in 2018 the main objective of which is to improve the understanding of the start-up of a SFR and to validate the fast reactor analysis computer codes against CEFR experimental data. A series of start-up tests have been analyzed in this benchmark and NINE also proposed and organized a further work package focused on the sensitivity and uncertainty analysis of the first criticality test. The present chapter intends to summarize the results achieved using the codes currently employed in the field of fast reactor in the framework of international projects and benchmarks in which NINE was involved and emphasize how the application of developed procedures allows to validate the SM results and validate the computer codes against experimental data.

S-CO2 Turbine Design for Decay Heat Removal System of Sodium Cooled Fast Reactor

2016 ◽  
Author(s):  
Seong Kuk Cho ◽  
Jekyoung Lee ◽  
Jeong Ik Lee ◽  
Jae Eun Cha
Keyword(s):  
Experimental Data ◽  
Fast Reactor ◽  
Design Methodology ◽  
Nuclear Reactors ◽  
Heat Removal ◽  
Design Code ◽  
Decay Heat ◽  
Turbine Design ◽  
Heat Removal System ◽  
Removal System

A Sodium-cooled Fast Reactor (SFR) has receiving attention as one of the promising next generation nuclear reactors because it can recycle the spent nuclear fuel produced from the current commercial nuclear reactors and accomplish higher thermal efficiency than the current commercial nuclear reactors. However, after shutdown of the nuclear reactor core, the accumulated fission products of the SFR also decay and release heat via radiation within the reactor. To remove this residual heat, a decay heat removal system (DHRS) with supercritical CO2 (S-CO2) as the working fluid is suggested with a turbocharger system which achieves passive operational capability. However, for designing this system an improved S-CO2 turbine design methodology should be suggested because the existing methodology for designing the S-CO2 Brayton cycle has focused only on the compressor design near the critical point. To develop a S-CO2 turbine design methodology, the non-dimensional number based design and the 1D mean line design method were modified and suggested. The design methodology was implemented into the developed code and the code results were compared with existing turbine experimental data. The data were collected under air and S-CO2 environment. The developed code in this research showed a reasonable agreement with the experimental data. Finally using the design code, the turbocharger design for the suggested DHRS and prediction of the off design performance were carried out. As further works, more effort will be put it to expand the S-CO2 turbine test data for validating the design code and methodology.

scholarly journals On Capabilities and Limitations of Current Fast Neutron Flux Monitoring Instrumentation for the Demo LFR ALFRED

2016 ◽  
Vol 2 (4) ◽  
Author(s):  
Luigi Lepore ◽  
Romolo Remetti ◽  
Mauro Cappelli
Keyword(s):  
Monte Carlo ◽  
Neutron Flux ◽  
Fast Reactor ◽  
Nuclear Power ◽  
Power Plants ◽  
Fuel Cycle ◽  
Signal To Noise Ratio ◽  
Reactor Power ◽  
Fast Reactors ◽  
Flux Monitoring

Among GEN IV projects for future nuclear power plants, lead-cooled fast reactors (LFRs) seem to be a very interesting solution due to their benefits in terms of fuel cycle, coolant safety, and waste management. The novelty of this matter causes some open issues about coolant chemical aspects, structural aspects, monitoring instrumentation, etc. Particularly, hard neutron flux spectra would make traditional neutron instrumentation unfit to all reactor conditions, i.e., source, intermediate, and power range. Identification of new models of nuclear instrumentation specialized for LFR neutron flux monitoring asks for an accurate evaluation of the environment the sensor will work in. In this study, thermal hydraulics and chemical conditions for the LFR core environment will be assumed, as the neutron flux will be studied extensively by the Monte Carlo transport code MCNPX (Monte Carlo N-Particles X-version). The core coolant’s high temperature drastically reduces the candidate instrumentation because only some kinds of fission chambers and self-powered neutron detectors can be operated in such an environment. This work aims at evaluating the capabilities of the available instrumentation (usually designed and tailored for sodium-cooled fast reactors) when exposed to the neutron spectrum derived from the Advanced Lead Fast Reactor European Demonstrator, a pool-type LFR project to demonstrate the feasibility of this technology into the European framework. This paper shows that such a class of instrumentation does follow the power evolution, but is not completely suitable to detect the whole range of reactor power, due to excessive burnup, damages, or gamma interferences. Some improvements are possible to increase the signal-to-noise ratio by optimizing each instrument in the range of reactor power, so to get the best solution. The design of some new detectors is proposed here together with a possible approach for prototyping and testing them by a fast reactor.

Simulation Model of a Passive Decay Heat Removal System for Lead-Cooled Fast Reactors

2012 ◽  
Author(s):  
Lorenzo Damiani ◽  
Alessandro Pini Prato
Keyword(s):  
Fast Reactor ◽  
Heat Removal ◽  
Coolant Temperature ◽  
Feed Water ◽  
Steam Generators ◽  
Fast Reactors ◽  
Primary Coolant ◽  
Reaction Heat ◽  
Generator System ◽  
Decay Heat

The generation IV lead cooled fast reactors are of particular interest for the Italian research: several influential companies (Ansaldo Nucleare, ENEA) are involved in these important European R&D projects. At present, one significant European project in progress is LEADER (Lead cooled European Advanced DEmonstrator Reactor) which includes, among its goals, the construction of a lead-cooled fast reactor demonstrator, ALFRED (Advanced Lead Fast Reactor European Demonstrator). The demonstrator has to include technical solutions that simplify the construction phase and assure full safety in operation; according to the latest guidelines, ALFRED final configuration will be characterized by a secondary loop providing bayonet-tube steam generators. The Authors have addressed the issue of bayonet-tube steam generators proposing the EBBSG (External Boiling Bayonet Steam Generator) system, in which the reaction heat is extracted from the lead by means of coolant under vapor phase. This is possible thanks to an external feed-water boiling, based on the known Loeffler scheme, coupled to the bayonet tube concept. In the present paper, the Authors propose a decay heat removal (DHR) system to match the EBBSG scheme. The DHR system is fully passive, exploiting natural circulation phenomena. The performance of the proposed DHR system is investigated through a Matlab-Simulink model. The results are satisfactory since, according to the simulations, the proposed DHR system is able to keep the primary coolant temperature within a safety range for a sufficient time, avoiding the lead freezing or over-heating.

Simulation Model of a Passive Decay Heat Removal System for Lead-Cooled Fast Reactors

2014 ◽  
Vol 137 (3) ◽  
Author(s):  
Lorenzo Damiani ◽  
Alessandro Pini Prato
Keyword(s):  
Fast Reactor ◽  
Natural Circulation ◽  
Heat Removal ◽  
Coolant Temperature ◽  
Feed Water ◽  
Steam Generators ◽  
Fast Reactors ◽  
Primary Coolant ◽  
Reaction Heat ◽  
Decay Heat

The generation IV lead cooled fast reactors are of particular interest for the Italian research: several influential companies (Ansaldo Nucleare, ENEA) are involved in these important European R&D projects. At present, one significant European project in progress is lead cooled European advanced demonstrator reactor (LEADER) which includes, among its goals, the construction of a lead-cooled fast reactor demonstrator, advanced lead fast reactor European demonstrator (ALFRED). The demonstrator has to include technical solutions that simplify the construction phase and assure full safety in operation; according to the latest guidelines, ALFRED final configuration will be characterized by a secondary loop providing bayonet-tube steam generators. The authors have addressed the issue of bayonet-tube steam generators proposing the external boiling bayonet steam generator (EBBSG) system, in which the reaction heat is extracted from the lead by means of coolant under vapor phase. This is possible thanks to an external feed-water boiling, based on the known Loeffler scheme, coupled to the bayonet tube concept. In the present paper, the authors propose a decay heat removal (DHR) system to match the EBBSG scheme. The DHR system is fully passive, exploiting natural circulation phenomena. The performance of the proposed DHR system is investigated through a Matlab-Simulink model. The results are satisfactory since, according to the simulations, the proposed DHR system is able to keep the primary coolant temperature within a safety range for a sufficient time, avoiding the lead freezing or over-heating.

scholarly journals EXTENDING CTF MODELING CAPABILITIES TO SFRs AND VALIDATION AGAINST SHRT TESTS

2021 ◽  
Vol 247 ◽  
pp. 10034
Author(s):  
Ahmed Aly ◽  
Agustin Abarca ◽  
Maria Avramova ◽  
Kostadin Ivanov
Keyword(s):  
Liquid Metals ◽  
Heat Removal ◽  
Light Water Reactors ◽  
Fuel Cladding ◽  
Fast Reactors ◽  
Mixing Coefficient ◽  
Flow Mixing ◽  
Computer Codes ◽  
Water Reactors ◽  
Transient Phases

The utilization of liquid metals as coolants for fast reactors brings several economical and practical advantages that lead to a sustainable future for nuclear energy. Molten sodium is used as a coolant in Sodium Fast Reactors (SFRs). Sodium is relatively cheaper than other metal coolants. It requires lower pumping power, causes less neutron moderation and it is non-corrosive to the fuel cladding. The SFR hexagonal subassemblies are relatively smaller than Light Water Reactors (LWRs) subassemblies. The differences in the geometrical design of SFRs compared to LWRs lead to different physical behavior of the coolant. Several models and correlations particular to sodium were implemented in thermal-hydraulics (TH) computer codes in order to model the coolant behavior accurately. CTF is a subchannel TH code that was designed and validated for LWRs. In this work, the capabilities of the code were extended to SFRs by incorporating sodium coolant properties and correlations for friction factor, flow mixing coefficient and conduction heat transfer. The code was then validated against selected steady state data from the Experimental Breeder Reactor II Shutdown Heat Removal Tests SHRT-17 and SHRT-45R. CTF was used to simulate the instrumented subassemblies XX09 and XX10. The results demonstrate the capability of CTF to model SFRs. Code validation is currently being extended to the transient phases of the SHRT experiments.

Development and Basic Verification of Decay Heat Removal Analysis Code of Sodium-Cooled Fast Reactor

2018 ◽  
Author(s):  
Ping Song ◽  
Dalin Zhang ◽  
Tangtao Feng ◽  
Shibao Wang ◽  
Yapei Zhang ◽  
...  
Keyword(s):  
Experimental Data ◽  
Fast Reactor ◽  
Heat Removal ◽  
Code Verification ◽  
Design And Optimization ◽  
Decay Heat ◽  
Removal Capacity ◽  
Heat Removal System ◽  
China Experimental Fast Reactor ◽  
Pool Type

As one of the generation IV reactors, pool-type Sodium-cooled Fast Reactors (SFRs) is attracting more and more attention. Loss of flow and heat sink accident is one of the most serious accidents for SFRs. Therefore, the decay heat removal capacity after emergency shutdown is of great importance. This paper has developed a one-dimensional code named Decay heat Removal Analysis Code for Sodium-cooled Fast Reactor (DRAC-SFR) to analyze the decay heat removal capacity. In the code, the decay heat removal system contains the primary loop, the intermediate loop and air circuit. The decay heat is removed out step by step with the above three loops. Many studies have been conducted on code verification. The international benchmark analysis of EBRII reactor is applied in the code verification. The calculation is compared with the experimental data and the results of DRAC-SFR agreed well with the experimental data. The comparison with the steady state of China Experimental Fast Reactor (CEFR) shows a good agreement with the design value. The errors of all the compared parameters are within 2%. What’s more, calculation is performed to analyze the characteristics of the decay heat removal capacity for CEFR. Thus, code verification shows that DRAC-SFR is proper to evaluate the decay heat removal capacity for SFRs and has the ability to provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.

Thermal-Hydraulic Analysis of the EBR-II SHRT-45R Unprotected Loss of Flow Experiment With Modified RELAP5

2017 ◽  
Author(s):  
Tang Simiao ◽  
Song Jian ◽  
Zhang Dalin ◽  
Wang Chenglong ◽  
Qiu Suizheng ◽  
...  
Keyword(s):  
Fast Reactor ◽  
Plant Protection ◽  
Rapid Development ◽  
National Laboratory ◽  
Argonne National Laboratory ◽  
Heat Removal ◽  
Safety Performance ◽  
Metallic Fuel ◽  
Fast Reactors ◽  
Negative Reactivity

Sodium-cooled fast reactor (SFR) is one of most promising Generation IV reactor technology and has a rapid development in recent years. Experimental Breeder Reactor II (EBR-II) designed by Argonne National Laboratory (ANL) is a typical sodium-cooled fast reactor with a sodium-bonded metallic fuel core, featured with reactor negative reactivity feedback. In order to verify and validate (V&V) the inherent safety performance of sodium-cooled fast reactor, the International Atomic Energy Agency (IAEA) established the coordinated research project (CRP) in which Xi’an Jiaotong University has participated. The “Benchmark Analysis of EBR-II shutdown heat removal tests” was conducted by ANL. SHRT-45R, namely Unprotected Loss of Flow (ULOF), is one of the reactor tests among many transients. Argonne National Laboratory has prepared a detailed benchmark specification and has provided the enough benchmark data for SHRT-45R. At the beginning of SHRT-45R both primary main pumps and the intermediate loop pump were synchronously tripped to simulate the unprotected loss of flow accident. During the test, the plant protection system (PPS) was disabled to initiate a control rod scram. The SHRT-45R test demonstrated that EBR-II could keep in safe during the potentially adverse consequences of unprotected accidents. This paper introduces the models for predicting SHRT-45R in detail and presents the results of the analysis of the Unprotected Loss of Flow (ULOF) test SHRT-45R performed in the EBR-II reactor. The thermal-hydraulic calculations are performed with Modified RELAP5 in which the thermodynamic and transport properties of liquid and vapor state sodium have been supplemented, as well as the specific heat transfer correlations. The numerical results show that the loss of forced coolant flow causes the coolant temperatures in the instrumented subassemblies XX09 and XX10 to increase to a peak point but keep at an acceptable level about 930K and 850K at the early state of accidents, and then the reactor can shut down by itself due to the negative reactivity feedback including Doppler reactivity feedback and density reactivity feedback after 70s. The variation of the key thermal-hydraulic parameters including the coolant temperatures and fuel cladding temperature in the instrumented subassembly has a good agreement with the experimental data in general. The results could not only verify and validate the inhere safety performance of sodium-cooled fast reactors during potentially unprotected accidents but also demonstrate the analysis capability of the modified RELAP5 for sodium-cooled fast reactors.

Current Status of the Development of the Fast Reactor Fuel Cycle in India

2009 ◽  
pp. 120-126
Author(s):  
K.V. Govindan Kutty ◽  
P.R. Vasudeva Rao ◽  
Baldev Raj
Keyword(s):  
Fast Reactor ◽  
Fuel Cycle ◽  
Reactor Fuel ◽  
Current Status ◽  
Fast Reactor Fuel

Nuclear Non-Proliferation Security on Exportation of Fast Reactors with a Closed Fuel Cycle

Atomic Energy ◽  
2021 ◽  
Author(s):  
N. V. Gorin ◽  
N. P. Voloshin ◽  
Yu. I. Churikov ◽  
A. N. Chebeskov ◽  
V. P. Kuchinov ◽  
...  
Keyword(s):  
Fuel Cycle ◽  
Fast Reactors ◽  
Closed Fuel Cycle

Prospects for fast reactors

1987 ◽  
Vol 92 (1-2) ◽  
pp. 57-71
Author(s):  
C. W. Blumfield
Keyword(s):  
Fast Reactor ◽  
High Energy ◽  
High Energy Density ◽  
Fast Reactors ◽  
Advantages And Disadvantages ◽  
Reactor Technology ◽  
European Collaboration ◽  
Fast Reactor Fuel ◽  
Fuel Reprocessing ◽  
Planned Activities

SynopsisThe background to recent major advances in the construction and operation of fast reactors is outlined with particular reference to the Dounreay Prototype Fast Reactor. The advantages and disadvantages of sodium as a coolant of the high energy density assembly are discussed and an account given of the consequences of a leak and the precautions taken against this eventuality. Attention is drawn to the safety aspects of the system. The economics of the plans for fuel reprocessing are explained and an account given of the progress in the fabrication of fast reactor fuel pins. Finally the environmental impact of present and planned activities on the Dounreay site is presented in the context of participation in the European Collaboration on Fast Reactor Technology and attention drawn to the importance of the planning inquiry in progress at Dounreay.

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