**1. Introduction**

#### **1.1 History, main features, advantages and future of sodium-cooled fast reactors**

Since the very beginning of its commercial operation, immediately after the end of the Second World War, nuclear energy has been getting a significant and often increasing part in the production of safe, secure, and economic low carbon electricity.

Innovation has always been - and still is nowadays - a powerful engine for progress in the fields of regulation (also including the trans-national aspects of the emergency management), safety (with a specific care and attention paid to severe accident prevention and mitigation, e.g. through inherent and passive safety features), reliability and efficiency in design and operation (including reliability and independence of control systems), incineration of long-life by-products of the fission-conversion-breeding process, non-proliferation (uselessness of fuel materials for weapon production), environmental impact (to the air, the soil and the water, both in normal operation and in emergency), management of high and low activity wastes, and also in the very sensitive domain of the public-awareness and acceptance, which are the key-issues for the civil nuclear future [1, 2].

This trend has been even more reinforced after the Fukushima events, also accounting for the wide stress test campaign conducted worldwide, as well as the large effort for public information, participation, and inclusion carried-out by International Bodies, Governments, Constructors, Operators, Academia and Research Organizations [3].

Safety has undergone a continuous improvement effort and has been a relevant driving force for progress, improvement as well as research and development in different fields of endeavor for the current GEN III (Generation III) and GEN III+ (Generation III plus) reactor designs, but also for the advanced concept-designs both inside the GIF (Generation IV International Forum) framework and outside, and in complement to it, e.g. with the ever growing interest for the SMRs (Small Modular Reactors), small compact elementary modules, generally sizing from 10MWe to 300MWe, which are designed and engineered along with a modular construction approach enabling to combine them and incrementally extend the power capacity of the overall plant thus offering economy of scale and reducing both capital costs and construction time [4]. Designs with power outputs smaller than 10MWe, often designed for semiautonomous operation, have been referred to as Micro Modular Reactors (MMRs).

Today, facing the high investment needs and the ever increasing costs, the large delays in the licensing process and the construction, the highly expensive financing modes as well as the low public acceptance and sometimes even the fierce opposition of a majority of the population, some developed countries (mainly in the Western Europe, even if Europe in a whole lasts hosting the largest nuclear capacity of the world), have decided to either step out or phase out nuclear energy in a short-medium term.

Nevertheless, nuclear energy still enjoys an increasing and dynamic trend. The Year 2018 has even been a hit as for the installed new nuclear capacity, mainly because the interest for nuclear reactors has widely moved from developed to emergent - developing countries. This trend is to continue and even expand as, according

**243**

Japan (see **Figure 1**).

*International Benchmark Activity in the Field of Sodium Fast Reactors*

to the activity of the spent fuel for thousands of years.

to current estimations, the installed nuclear capacity should double in the emergent economies within the next 20years. Relying upon a robust industrial capacity, the Russian Federation is today by far the larger exporter/provider of nuclear technology worldwide, and the People Republic of China is on the way to become a future

In order to allow nuclear power contribute effectively to the solution of the global warming challenge in the future, it shall be necessary: to continuously up-date and improve regulations; to enhance the safety under the guidance of proactive, transparent and independent Safety Authorities; to establish suitable roadmaps providing all the actors in the nuclear field with a medium - long term clear vision, and to reduce overall costs through continuous improvement, harmonization of practices and standardization. But, mostly, it will be worth addressing and providing a longlasting and sustainable solution to the crucial problem of the long-lived wastes.

Today, the installed nuclear capacity is by far from GEN III reactors, only a few of them belonging to the GEN III+ generation (which includes, e.g., French EPR, American AP-1000, Russian VVER-1000 …), and even less to other concepts. During their operation, such reactors produce, as by-products of the fission-conversion-breeding process, a large quantity of long-lived isotopes, quoted as Actinides and/or Minor Actinides, depending on their features and nature, which contribute

The build-up of such by-products turns-out a major challenge both from the non-proliferation and the waste management viewpoint. Their recycling in the reactor fuel as well as their incineration through suitable strategies will contribute to "close the cycle", i.e., at least theoretically, to bring the spent fuel activity back to a level comparable to the natural earth radiation. The acceptance of the public of further installations of nuclear power plants will strongly depend in the future on

Fast reactors closed fuel cycle can efficiently and effectively contribute to the solution of the problem decreasing the burden of nuclear waste and supporting long-term nuclear power development as part of the world's future energy mix [5]. Global interest in fast reactors has been growing since their inception in 1960's because they can provide efficient, safe and sustainable energy. In addition to the current fast reactor construction projects underway, several countries are engaged in intense research and development programs for the development of innovative, Gen IV, fast reactor concepts, as proposed by the GIF. They include three fast neutrons concepts: the SFR - Sodium Fast Reactor -, the LFR - Lead (or Lead-Bismuth) Fast Reactor - and the GFR (Gas Fast Reactor), as well as the MSR (Molten Salt Reactor) which can be declined both in a thermal and a fast neutrons version.

Moreover, current developments of SMRs include, among the more than 100 versions under study, development and/or licensing, several fast neutrons concepts, even though the most mature ones are undoubtedly based on LWR (Light Water Reactor) technology. The fast SMRs, in addition to their efficient use of the fuel, are flexible because they can operate either as breeders, to produce fissile material, or as burners of Plutonium and/or long-lived Minor Actinides. Combining this capability with the benefits in terms of power generation flexibility, SMRs could turn-out quite attracting. The SFR is, by far, the fast reactor technology most widely spread-out worldwide. It enjoys an acknowledged maturity due to the numerous constructions and because it underwent many years of operation in several countries, from the late '60 prototypes up to the development and deployment of the industrial French fleet (including RAPSODIE, Phénix and Superphénix - the biggest fast reactor ever built, now decommissioned -, and the project ASTRID, now delayed), and other reactors now either in operation or under construction in Russia, India, China and

*DOI: http://dx.doi.org/10.5772/intechopen.97812*

leader in the nuclear field.

this crucial problem.

#### *International Benchmark Activity in the Field of Sodium Fast Reactors DOI: http://dx.doi.org/10.5772/intechopen.97812*

*Recent Advances in Numerical Simulations*

**1. Introduction**

electricity.

Research Organizations [3].

as Micro Modular Reactors (MMRs).

and emphasize how the application of developed procedures allows to validate the

**1.1 History, main features, advantages and future of sodium-cooled fast reactors**

Since the very beginning of its commercial operation, immediately after the end of the Second World War, nuclear energy has been getting a significant and often increasing part in the production of safe, secure, and economic low carbon

Innovation has always been - and still is nowadays - a powerful engine for progress in the fields of regulation (also including the trans-national aspects of the emergency management), safety (with a specific care and attention paid to severe accident prevention and mitigation, e.g. through inherent and passive safety features), reliability and efficiency in design and operation (including reliability and independence of control systems), incineration of long-life by-products of the fission-conversion-breeding process, non-proliferation (uselessness of fuel materials for weapon production), environmental impact (to the air, the soil and the water, both in normal operation and in emergency), management of high and low activity wastes, and also in the very sensitive domain of the public-awareness and

acceptance, which are the key-issues for the civil nuclear future [1, 2].

This trend has been even more reinforced after the Fukushima events, also accounting for the wide stress test campaign conducted worldwide, as well as the large effort for public information, participation, and inclusion carried-out by International Bodies, Governments, Constructors, Operators, Academia and

Safety has undergone a continuous improvement effort and has been a relevant driving force for progress, improvement as well as research and development in different fields of endeavor for the current GEN III (Generation III) and GEN III+ (Generation III plus) reactor designs, but also for the advanced concept-designs both inside the GIF (Generation IV International Forum) framework and outside, and in complement to it, e.g. with the ever growing interest for the SMRs (Small Modular Reactors), small compact elementary modules, generally sizing from 10MWe to 300MWe, which are designed and engineered along with a modular construction approach enabling to combine them and incrementally extend the power capacity of the overall plant thus offering economy of scale and reducing both capital costs and construction time [4]. Designs with power outputs smaller than 10MWe, often designed for semiautonomous operation, have been referred to

Today, facing the high investment needs and the ever increasing costs, the large delays in the licensing process and the construction, the highly expensive financing modes as well as the low public acceptance and sometimes even the fierce opposition of a majority of the population, some developed countries (mainly in the Western Europe, even if Europe in a whole lasts hosting the largest nuclear capacity of the world), have decided to either step out or phase out nuclear energy in a short-medium term. Nevertheless, nuclear energy still enjoys an increasing and dynamic trend. The Year 2018 has even been a hit as for the installed new nuclear capacity, mainly because the interest for nuclear reactors has widely moved from developed to emergent - developing countries. This trend is to continue and even expand as, according

SM results and validate the computer codes against experimental data.

**Keywords:** SFR, Benchmark, EBR-II, FFTF, CEFR, M&S tools

**242**

to current estimations, the installed nuclear capacity should double in the emergent economies within the next 20years. Relying upon a robust industrial capacity, the Russian Federation is today by far the larger exporter/provider of nuclear technology worldwide, and the People Republic of China is on the way to become a future leader in the nuclear field.

In order to allow nuclear power contribute effectively to the solution of the global warming challenge in the future, it shall be necessary: to continuously up-date and improve regulations; to enhance the safety under the guidance of proactive, transparent and independent Safety Authorities; to establish suitable roadmaps providing all the actors in the nuclear field with a medium - long term clear vision, and to reduce overall costs through continuous improvement, harmonization of practices and standardization. But, mostly, it will be worth addressing and providing a longlasting and sustainable solution to the crucial problem of the long-lived wastes.

Today, the installed nuclear capacity is by far from GEN III reactors, only a few of them belonging to the GEN III+ generation (which includes, e.g., French EPR, American AP-1000, Russian VVER-1000 …), and even less to other concepts. During their operation, such reactors produce, as by-products of the fission-conversion-breeding process, a large quantity of long-lived isotopes, quoted as Actinides and/or Minor Actinides, depending on their features and nature, which contribute to the activity of the spent fuel for thousands of years.

The build-up of such by-products turns-out a major challenge both from the non-proliferation and the waste management viewpoint. Their recycling in the reactor fuel as well as their incineration through suitable strategies will contribute to "close the cycle", i.e., at least theoretically, to bring the spent fuel activity back to a level comparable to the natural earth radiation. The acceptance of the public of further installations of nuclear power plants will strongly depend in the future on this crucial problem.

Fast reactors closed fuel cycle can efficiently and effectively contribute to the solution of the problem decreasing the burden of nuclear waste and supporting long-term nuclear power development as part of the world's future energy mix [5].

Global interest in fast reactors has been growing since their inception in 1960's because they can provide efficient, safe and sustainable energy. In addition to the current fast reactor construction projects underway, several countries are engaged in intense research and development programs for the development of innovative, Gen IV, fast reactor concepts, as proposed by the GIF. They include three fast neutrons concepts: the SFR - Sodium Fast Reactor -, the LFR - Lead (or Lead-Bismuth) Fast Reactor - and the GFR (Gas Fast Reactor), as well as the MSR (Molten Salt Reactor) which can be declined both in a thermal and a fast neutrons version.

Moreover, current developments of SMRs include, among the more than 100 versions under study, development and/or licensing, several fast neutrons concepts, even though the most mature ones are undoubtedly based on LWR (Light Water Reactor) technology. The fast SMRs, in addition to their efficient use of the fuel, are flexible because they can operate either as breeders, to produce fissile material, or as burners of Plutonium and/or long-lived Minor Actinides. Combining this capability with the benefits in terms of power generation flexibility, SMRs could turn-out quite attracting.

The SFR is, by far, the fast reactor technology most widely spread-out worldwide. It enjoys an acknowledged maturity due to the numerous constructions and because it underwent many years of operation in several countries, from the late '60 prototypes up to the development and deployment of the industrial French fleet (including RAPSODIE, Phénix and Superphénix - the biggest fast reactor ever built, now decommissioned -, and the project ASTRID, now delayed), and other reactors now either in operation or under construction in Russia, India, China and Japan (see **Figure 1**).

**Figure 1.** *World Sodium Fast Reactor Status.*

Design and operation of such reactors are demanding extended computation capacity, to assess their safety, security, and economics [6], which justifies the organization under IAEA's umbrella of Coordinated Research Projects (CRPs) 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. NINE is very actively participating in these exercises, and sometimes conducting them.

The present chapter summarizes the results and discusses the main outcomes of the above-mentioned benchmark exercises, in the aim at underlying the wide convergence among the computational tools adopted by the participants, as well as detecting the main discrepancies and seeking for their common origin and trend, whether and whenever existing. That should enable defining a mid-term vision for further development of the computer codes in the field of fast reactors, whatever their features and nature, and identifying new needs for their extended validation against either available or expected experimental data.

#### **1.2 NINE involvement/interest in sodium-cooled fast reactors**

Starting from the considerations above regarding the deployment of fast reactors and the maturity gained by the SFR, NINE joints the effort of International community to assess the actual computational capabilities in modeling SFRs features. Taking advantage of reactor data gathered in full scale reactor demonstrators, NINE participated, and is still doing, in several International benchmarks aiming at demonstrating the applicability of its modeling methodology to Fast Reactor design and, in particular, to SFRs; to evaluate the level of assessment of computer codes available at NINE in respect to SFR specific features; to check the applicability of the NINE Validation Process – which is part of the more general framework of NEMM (NINE Evaluation Model Methodology)1 - with particular focus on the quantification of accuracies of the Thermal–Hydraulic (TH) simulations by means of Fast Fourier Transform Based Method (FFTBM) and finally to perform independent validation of the Serpent code.

All the analysis presented hereafter have been performed following a best estimate approach which requires, among the other things, a high-fidelity Simulation

**245**

*International Benchmark Activity in the Field of Sodium Fast Reactors*

**2. Analysis and validation of EBR-II SHRT-17**

Department of Energy at the Argonne-West site.

**2.1 The NEMM validation process**

*2.1.1 Overview of SCCRED methodology*

Best Estimate and Uncertainty applications.

of the performed code calculation;

input development;

Argonne's Experimental Breeder Reactor II (EBR-II) in the 1980's.

Model (SM), i.e., a SM that represents with a high level of detail the hardware subject of the analysis to avoid the introduction of inaccuracies due to rough

The IAEA CRP "Benchmark Analyses of EBR-II Shutdown Heat Removal Tests" was initiated in 2012 [7] with the objective of improving the state-of-the-art SFR codes by extending their validation to include comparisons against whole-plant data recorded during landmark Shutdown Heat Removal Tests (SHRT) conducted at

The EBR-II plant was a uranium metal-alloy-fuelled liquid-metal-cooled fast reactor designed and operated by Argonne National Laboratory (ANL) for the U.S.

Several loss of flow tests were conducted in the facility between 1984 and 1986, as a part of SHRT series [8]. SHRT-17, protected loss of flow transient, was one of the mentioned tests to demonstrate the inherent safety of LMR type reactors. At the beginning of the test, the primary pumps were tripped and at the same time a scram was actuated through a full control rod insertion. The effectiveness of natural circulation cooling capability of the reactor, which makes them inherently safe under described accident conditions, was successfully demonstrated by this test.

A key feature of the activities performed in the field of nuclear reactor safety is the need to demonstrate the validation level of each tool adopted within an assigned process and of each step of the concerned process. Therefore, the validation of best estimate codes, models, "best modeling practices" and uncertainty methods must be considered of great importance to ensure the validity of performed Best Estimate and Uncertainty analysis. A consistent code assessment supported by a qualified experimental database is an important step for developing a solid ground for the uncertainty evaluation in the frame of Best Estimate and Uncertainty approach. Thus, the SCCRED (Standardized and Consolidated Calculated & Reference Experimental Database) methodology [9], embedded in NEMM [10], has been developed to generate a series of documents and tools to set up a qualified experimental and calculated database for Verification and Validation (V&V) purposes of

**Figure 2** depicts the SCCRED diagram: the information contained in the experimental reports together with the code input nodalization are the sources to be elaborated in a systematic way by a qualified database made up of the following documents:

• The Reference Data Set for the selected facility or test (RDS-facility and RDStest) containing the information (geometrical data of the facility and boundary and initial conditions of the selected test, respectively) needed for the code

• The Validation Report (VR), which collects the results of the Validation Process

• The Engineering Handbook (EH), that describes the code input file and provides the engineering justifications of the code-user choices.

*DOI: http://dx.doi.org/10.5772/intechopen.97812*

approximations and assumptions.

<sup>1</sup> Within the NEMM (NINE Evaluation Model Methodology), NINE adopts the conventional and internationally acknowledged process to achieve the validation of computation tools and define the inherent uncertainties. It includes three steps: the analytical compliance test - the verification -, the qualification through code-to-code comparisons and benchmarks, the actual validation – supported by scaling analysis – on experimental data originating from mock-up experiments and the outcomes of the operating experience, including downgraded operation and emergency.

*Recent Advances in Numerical Simulations*

sometimes conducting them.

*World Sodium Fast Reactor Status.*

**Figure 1.**

Model Methodology)1

against either available or expected experimental data.

**1.2 NINE involvement/interest in sodium-cooled fast reactors**

Design and operation of such reactors are demanding extended computation capacity, to assess their safety, security, and economics [6], which justifies the organization under IAEA's umbrella of Coordinated Research Projects (CRPs) 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. NINE is very actively participating in these exercises, and

The present chapter summarizes the results and discusses the main outcomes of the above-mentioned benchmark exercises, in the aim at underlying the wide convergence among the computational tools adopted by the participants, as well as detecting the main discrepancies and seeking for their common origin and trend, whether and whenever existing. That should enable defining a mid-term vision for further development of the computer codes in the field of fast reactors, whatever their features and nature, and identifying new needs for their extended validation

Starting from the considerations above regarding the deployment of fast reactors and the maturity gained by the SFR, NINE joints the effort of International community to assess the actual computational capabilities in modeling SFRs features. Taking advantage of reactor data gathered in full scale reactor demonstrators, NINE participated, and is still doing, in several International benchmarks aiming at demonstrating the applicability of its modeling methodology to Fast Reactor design and, in particular, to SFRs; to evaluate the level of assessment of computer codes available at NINE in respect to SFR specific features; to check the applicability of the NINE Validation Process – which is part of the more general framework of NEMM (NINE Evaluation

the Thermal–Hydraulic (TH) simulations by means of Fast Fourier Transform Based Method (FFTBM) and finally to perform independent validation of the Serpent code. All the analysis presented hereafter have been performed following a best estimate approach which requires, among the other things, a high-fidelity Simulation

<sup>1</sup> Within the NEMM (NINE Evaluation Model Methodology), NINE adopts the conventional and internationally acknowledged process to achieve the validation of computation tools and define the inherent uncertainties. It includes three steps: the analytical compliance test - the verification -, the qualification through code-to-code comparisons and benchmarks, the actual validation – supported by scaling analysis – on experimental data originating from mock-up experiments and the outcomes of the

operating experience, including downgraded operation and emergency.


**244**

Model (SM), i.e., a SM that represents with a high level of detail the hardware subject of the analysis to avoid the introduction of inaccuracies due to rough approximations and assumptions.
