Abstract
A comparison of the Japan sodium-cooled fast reactor (JSFR) design with the future French sodium-cooled fast reactor (SFR) concept has been done based on the requirements of Electricité de France (EDF), the investor-operator of the future French SFR, and the French safety baseline, under the framework of an EDF and Japan Atomic Energy Agency (JAEA) bilateral agreement of research and development cooperation in future SFRs..
Keywords:
1. Introduction
In 2008, Electricité de France (EDF) and Japan Atomic Energy Agency (JAEA) signed a bilateral agreement for research and development (R&D) cooperation and information exchange on future sodium-cooled fast reactors (SFRs). Within the bilateral framework, a comparison of the Japan SFR (JSFR) design with the future French SFR concept has been done. First, the comparison has been conducted based on the requirements of the investor-operator of future French SFRs. EDF's primary requirements as an electric utility for future operation of an industrial generation four (Gen IV) SFR are briefly presented in reference [Citation1]. Second, the comparison has been done based on the French safety baseline that could be applicable to the future French SFRs, which is currently under preparation. This study shows that the French baseline, EDF's requirements for commercial SFRs, safety approach, design practices and codes, differ from the baseline which has been taken into account for JSFRs. Therefore, particular items in the JSFR design do not exactly meet EDF's requirements or other baselines. Throughout the comparison by EDF, some options of safety measures of JSFR such as self-actuated shutdown system (SASS) and fuel assembly with inner duct structure (FAIDUS) are considered interesting or positive. Some options for economic optimization such as reduction of the number of components or the number of loops could bring significant cost reduction. This paper describes the results of the comparison work of JSFR and EDF requirements for future SFRs where the specific designs of JSFR were evaluated as interesting from the EDF point of view. The comparison work pointed out the differences in safety baselines between two countries as well.
2. JSFR design outline
A conceptual design study and related R&D for the demonstration reactor and the commercial reactor has been carried out in the framework of the Japanese fast reactor cycle technology development (FaCT) project. JSFR is an advanced loop-type SFR selected in view of the potential advantage of a loop-type SFR for operation and maintenance capability as well as the perspective for the economical plant configuration. As a next generation plant, JSFR has adopted a number of innovative technologies in order to achieve economic competitiveness and enhancement of its reliability and safety.
Main innovative concept studies in JSFR are as follows: A compact reactor vessel (RV) without a vessel wall cooling system is pursued in consideration of the wall thickness enough to resist the severest seismic condition. A two-loop cooling system with shortened high-chromium steel piping is a crucial feature, and studies on the hydraulics in the L-shaped pipe elbow and the fabrication capability of the pipes are being carried out. A double-walled, straight tube steam generator (SG) is investigated to enhance the reliability against sodium/water reaction, and developmental works are progressing, including the thermal–hydraulic design and trial manufacturing for components. SASS is being developed with safety analysis of the applicability for JSFR. As an in-vessel retention concept, JSFR adopted FAIDUS to avoid re-criticality in the case of core disruptive accidents (CDA). An advanced fuel handling system (FHS) is pursued to enhance economic performance.
2.1. JSFR overview and innovative technologies
JSFR major parameters and the innovative technologies adopted are shown in and . The innovative technologies are categorized into 10 categories in terms of the plant design system, as shown in . The adoption judgments of the innovative technologies have been made with the view of design feasibility such as structural integrity, manufacturability, operation and maintenance, and economy. R&D and technical investigations are underway for some options such as application of the oxide dispersion strengthened steel (ODS) for cladding, and final judgments will be made according to its results.
Table 1. Major parameter of JSFR.
Table 2. Innovative JSFR technologies.
2.2. Core and fuel
One of specific aims of the core and fuel design is to achieve high burn-up, where the design target is 150 GWd/MT for the core region. ODS is used as fuel cladding material, as it has potential resistance to high neutron dose and high temperature.
Enhancement of safety, based on defense-in-depth principles, is also a key factor to be considered in the core and fuel design. The reactor shutdown system consists of two independent subsystems to prevent fuel failure for design basis accidents (DBA). As a passive safety feature, SASS is provided for reactor shutdown to enhance the prevention capability against CDA from anticipated transient without scram (ATWS).
Furthermore, FAIDUS is applied to the fuel subassembly to enhance molten fuel discharge in the case of CDA. This design allows molten fuel to be rapidly expelled from the core, so that the severe energetics due to nuclear excursion could be avoided coupled with restricting the core performance such as sodium void reactivity worth.
As CDA has been historically regarded as beyond design basis events (BDBE), JSFR accepted the traditional conservative hypothesis. However, JSFR is designed to mitigate CDA consequences to eliminate severe energetics.
2.3. Reactor system
A compact reactor system is indispensable for the purpose of achieving an economic competitiveness for JSFR, while keeping the accessibility to as many in-vessel components as possible in view of the capability of in-service inspection and repair (ISI&R) and maintaining the robustness even against the severest earthquake condition. A compact RV is a notable feature for this purpose, and an upper internal structure (UIS) with slit (slit UIS) design is applied to realize the compact RV. By applying the slit UIS design, the JSFR in-vessel fuel handling approach is compacted, which consists of a single rotating plug, a slit UIS and a fuel handling machine (FHM) with a pantograph. An FHS for JSFR has adopted some advanced concepts, consisting of an in-vessel FHM, in-vessel fuel transfer equipment (transfer pot), an external-vessel fuel transfer machine (EVTM), an external-vessel fuel storage tank (EVST) and a spent fuel storage water pool, spent fuel cleaning facilities and new fuel handling facilities. This FHS enables significant reduction of the RV diameter.
The “hot vessel” concept is pursued, where an RV wall cooling system is not installed, to realize a compact RV. This is adaptable for a potential benefit of SFR as a low-pressure system. A thinner RV wall is feasible to restrict the impact of thermal transient on the structural integrity of RV. On the other hand, specific care is given to the RV wall thickness so as to be sufficiently tough against severe earthquakes. In Japan, the hot vessel without cooling system design has successfully accumulated operating experiences in Joyo (Japanese experimental SFR) [Citation2] and Monju (Japanese prototype SFR) [Citation3]. This hot vessel concept was also employed in Rapsodie and SNR-300 RV designs. The JSFR RV protection is further simplified than Joyo and Monju without an external-vessel overflow system, while Joyo and Monju have external-vessel overflow systems to maintain a steady sodium level during start-up operation to reduce transient thermal stress.
A high-performance radial shielding with zirconium hydride (Zr-H) has been adopted, taking advantage of its high capability of neutron moderation, to reduce the peripheral diameter of the radial shielding. Meanwhile, the concept of radial shielding with Zr-H assemblies was re-examined after the accident at Fukushima Daiichi Nuclear Power station (NPS) where severe hydrogen explosions have seriously damaged the reactor buildings. Because the hydride existing in the core has the potential to cause hydrogen production, Zr-H shielding concept was excluded from the options of compacting RV.
2.4. Cooling system
In the JSFR design, the cooling system is simplified by a two-loop configuration that can satisfy the high-flow-velocity conditions in large-diameter pipes even with a large electricity generation of 1500 MWe, as well as by shortened piping with less elbows than that of a conventional fast reactor. The shortened primary and secondary pipes are designed with high chromium steel that has a high strength in the elevated temperature and a low thermal expansion coefficient. The primary hot-leg circuit has simple L-shaped piping. The curvature radius of the L-shaped short elbows for the hot-leg piping system is equivalent to the piping diameter, in order to compact the reactor building volume. As the thermal expansion of the L-shaped piping is absorbed by only one elbow, the modified 9 chromium–1 molybdenum (Mod. 9Cr-1Mo) steel is applied as piping material, as it has high strength and a low thermal expansion coefficient. The design of the JSFR hot-leg pipe was evaluated using the allowable stress, and the results demonstrated that the compact pipe design is assured [Citation4].
The JSFR design adopts fully natural convection to achieve reliable decay heat removal. The decay heat removal system (DHRS) of JSFR consists of a combination of one loop of direct reactor auxiliary cooling system (DRACS) and two loops of primary reactor auxiliary cooling system (PRACS) adopting full natural convection system. The heat exchanger of DRACS is dipped in the upper plenum within the RV. The heat exchanger of each PRACS is located in the primary-side upper plenum of an intermediate heat exchanger (IHX). DHRS can be operated by a fully passive feature with natural convection, which requires no active components such as pumps [Citation5]. The JSFR DHRS performance was confirmed by a 1/10 scale water test and evaluation tools were verified and validated by the experimental data [Citation6,7]. From the view point of probabilistic safety assessment (PSA), thanks to the passive features, frequency of protected loss of heat sink (PLOHS) was evaluated to be 2×10−8 per reactor year [Citation8] and less than 10−8 per reactor year taking into account accident management [Citation9].
2.5. Reactor building
The layout plan of reactor building has been investigated with an attention to maintaining an adequate seismic reliability and a space enough to keep operability of the plant and maintainability of the components and equipment arranged in the reactor building. A steel-plate-reinforced concrete containment vessel (SCCV) is a notable feature of the reactor building design. This innovative containment vessel (CV) design is a part of the reactor building, and has a potential to shorten the construction period by a unit construction method aimed at achieving a high economic performance.
While the double-walled structure of coolant boundaries for JSFR can reduce the possibility of sodium leak in the containment, further enhancement of safety is desirable to assure that the containment boundary function and structural strength of the CV can be maintained as the final barrier for the release of radioactive material even if sodium is assumed to leak in the CV.
For seismic design, JSFR adopts an advanced seismic isolation system for SFR which mitigates the horizontal seismic force by thicker laminated rubber bearings with a longer period and improvement of damping performance by adopting oil dampers. To confirm feasibility of the thicker laminated rubber bearings, a basic characteristic test with a 1/8 reduced scale model was carried out. As a result, the possibility of application of the thicker laminated rubber bearings was confirmed [Citation10].
3. Remarks on JSFR design from electricity utility's point of view
JSFR is pursuing a compact RV design by adopting innovative technologies, such as the UIS with a radial slit, the FHM with a pantograph arm, hot vessel concept, shortening of piping and reduction of loop number.
3.1. Hot vessel
The hot vessel concept is adopted to realize a compact RV in JSFR design, as described in Section 2.3.
The thermal ratcheting at the sodium level combined with creep on the JSFR hot vessel has been evaluated by French side with the European fast reactor (EFR) [Citation11] loadings and condition of 60-year lifetime. The result showed that the hot vessel option is not compatible with these conditions. There still exist difficulties with the thermo-mechanical justification of a hot vessel. And as it is probable that the design margin for seismic reliability becomes stricter in the future, the alternative designs are considered for JSFR – for instance, a sodium dam and a cold vessel. In France, it is implicitly assumed for the design of a nuclear reactor primary system that components that participate in the confinement safety function (a fortiori if they also participate in the core support function, which is the case of the RV) must remain within the negligible creep range (for example, the primary vessel and the core support). This recognition is shared with the French operator and the French nuclear safety authority (ASN); the recognition is based on several considerations: creep is a damage that cannot be or can only slightly be checked by inspection; it is important to maintain the “creep capacity” for accident cases (loss of heat removal function, etc.). It is true and well-known that creep damage is not detectable with nondestructive examination means. When it is detectable, it has reached an advanced stage (tertiary creep) so that is too late to ensure the integrity of the vessel. If serious and unknown creep damage affects the vessel, and then a long accident involving high temperatures occurs (e.g. loss of decay heat removal), the vessel could reach the limit of creep capacity.
JAEA specifies that for the JSFR RV, thermal ratcheting and creep fatigue damage on RV structures as sodium level, thermal stratification boundary and core support have been evaluated and confirmed to meet design limits. The most severe case is the thermal ratcheting on the RV wall at the sodium level. Evaluation of thermal ratcheting at the sodium level based on the Japanese design guide, the “elevated temperature structural design guide for demonstration fast breeder reactor (DDS)” [Citation12], was conducted. The evaluation shows that there is no ratcheting strain with start-up operation longer than 3.5 days [Citation9]. To reduce damage on the RV wall, a sodium dam concept has been proposed, the concept is shown in . The sodium dam can keep the sodium level constant during the start-up. This simple structure can reduce thermal stress on the RV damage dramatically even though the RV is still exposed to hot sodium in the normal operation. The evaluation shows that the RV with the sodium dam can achieve 1.5-day start-up even without ring forged material, where the JSFR RV adopts ring forged low-carbon and medium-nitrogen type 316 stainless steel (316FR) instead of type 304 stainless steel (304SS) to improve high temperature strength.
Figure 1. Sodium dam concept.
From the EDF point of view, the hot vessel needs further R&D, but this option would result in a saving of construction cost. Therefore, it would be desirable to continue studies on the hot vessel option. There is a great deal of interest if different countries could harmonize their rules internationally, as French and Japanese positions for this particular aspect do not rely on the same requirement level.
3.2. Two-loop primary cooling system
JSFR adopts a two-loop primary cooling system and achieves a compact component arrangement adopting the L-shape pipe for the primary hot-leg piping, as described in Section 2.4. The flow rate of sodium in a primary cooling system increases because the number of loops is minimized, thus a large-diameter piping system is required.
Major issues: performance of DHRS, loss-of-flow type events and hydraulics property in two-loop system were clarified and evaluated in the previous study [Citation5], and the basic feasibility of the two-loop cooling system has already been confirmed. As for the design basis events (DBE), the primary pump (PP) seizure in one loop accident (loss-of-flow type event) has appeared to be the severest event and the transient analysis taking into account the latest design has shown that the two-loop cooling system meets safety criteria [Citation13]. For DHRS, as mentioned in Section 2.4, the evaluation shows that the frequency of PLOHS was 2×10−8 per reactor year [Citation8].
From the hydraulic point of view, the L-shaped pipes have better characteristics of flow-induced vibrations resistance than U-shaped pipes, but their thermo-mechanical behavior is more difficult to justify than U-shaped ones. The risk of cavitation in elbows requires either high pressurization of the reactor argon blanket or an increase in the loop diameter, but the thermo-mechanical behavior deteriorates in the latter case because the pipes become more rigid. The resistance to flow-induced vibrations has been estimated on the French side with a simplified way of which the criterion is being much simplified (natural frequency ≥2, frequency corresponding to the Stroudhal number). Since this criterion is simplified, the extrapolation is uncertain. The resistance to cavitation was also estimated by French side, using a simplified criterion [Citation14], and it is concluded that the resistance of L-shaped pipes is not guaranteed, even with a high reactor cover gas pressure.
Nevertheless the criterion of the reference [Citation14] is considered preliminary to make a final decision because it is worth only for diameters less than 304 mm and Reynolds number less than 8×106, so it is not representative of JSFR reactor features. Then the extrapolation to JSFR is uncertain.
EDF also pointed out a possible problem resulting from high cover gas pressure. In France, there is a tendency to emphasize the negative effects of a high reactor argon blanket pressure because of the following two issues:
Submission to ESPN (French statutory rule for pressurized nuclear equipment) is required when the cover gas pressure exceeds 1.5 bars absolute (0.5 bars over atmospheric pressure), and pressure resistance tests, generally by the method of hydrostatic pressure test, will be required.
The geyser effect will cause sodium leak due to high cover gas pressure. This only occurs when there are simultaneous leaks on the hot leg, the pipe, and the double envelope.
The ESPN rule has been written for existing plants, and at this time only pressurized water reactors (PWR) were operated, no SFR was planned, so its application to sodium fast reactors raises difficulties. It might be hoped that an adaptation of ESPN (keeping the main basis of the rule but taking into account the specificities of sodium technology) will be examined for SFRs. Regarding the geyser effect, as JSFR adopts a double boundary for primary cooling system, a large sodium leak could be prevented by sodium leak detection and following reactor trip. The cover pressure should be decreased by the normal reactor shutdown procedure after the reactor trip and following pump trip, but the double boundary will strongly mitigate the geyser effect if there is no failure in a series of trips process and reactor shutdown procedure.
JAEA points out that for JSFR, the resistance to vibrations has been studied by using a 1/3 scale hot-leg pipe water experiments with an acryl pipe for visualization and with a stainless pipe for vibration data accumulation [Citation5,Citation15], and the results led to low predicted mechanical stresses. Detail vibration data was also accumulated from the water experiment with the stainless steel pipe. With the accumulated data, conservative design power spectrum density (PSD) for stress analysis on random vibration has been defined as shown in . And from the results of the tests carried out, it is estimated that the risk of vibration is controlled. The demonstration for the cold leg remains to be made, but it should not cause any problem due to lower velocities in the cold-leg piping.
Figure 2. PSD for hot-leg piping design.
Concerning the cavitation in JSFR design, the cavitation factor k was used to estimate the onset condition of the vortex cavitation in the reactor, where the factor k is given as k = (P − Ps)/0.5ρv2.
The 1/10 scaled model water experiment for the upper plenum of RV was conducted by JAEA to confirm the probability of cavitation occurrence, where the cavitation factor of JSFR design is around 8. Criteria for onset of cavitation for the experiment were selected based on JSFR cavitation factor taking into account the effects of differences between experiments and JSFR (sodium/water, pressure, scale size etc.). As a result of the experiments, cavitation inception factor was found to satisfy the criteria of the experiment, and thus cavitation can be prevented by design for JSFR [Citation16,17].
For the thermo-mechanical behavior of hot primary loops, a simplified calculation based on RCC-MR (French design and construction rules for mechanical components of fast breeder reactor nuclear islands) for similar diameter as JSFR has been done by French side, where the pipe thickness was 20 mm instead of 15.9 mm, and the pipe material was austenitic steel instead of Mod. 9Cr steel. Therefore, the L-shape pipe resistance was not justified. From material point of view, according to the French experts, the main problems of Mod. 9Cr piping are: cyclic softening, little knowledge of long-term thermal aging, need for post-welding heat treatment which is difficult to implement.
Concerning the design criteria, JAEA has developed them from the viewpoint of thermal stress and material strength [Citation4]. Creep strength of welded joints of Mod.9Cr-1Mo steel decreases in comparison with base metal under the conditions of high temperature of over 600 °C and long-time period; that is known as Type-IV damage. The creep strength reduction by Type-IV damage is taken into account in the hot-leg piping design of primary cooling system, although definite reduction has not been observed at the JSFR temperature of 550 °C. Since experimental results are still limited for the JSFR condition of 550 °C, a conservative stress limit has been proposed based on the available experimental data of 600 °C and 650 °C conditions. The limit of the initial thermal stress has been evaluated taking into account stress relaxation. The elastic follow-up parameter for the major part of the hot-leg piping has been evaluated based on the finite element analysis with shell elements. A conservative elastic follow-up parameter of 2.0 has been decided from the result to evaluate stress relaxation in the piping design. With this conservative elastic follow-up parameter, the limit of the initial stress for welded parts is evaluated to be 150 MPa. The initial thermal stress has been evaluated as shown in . It is confirmed that the evaluated initial stresses at all welded parts are lower the proposed design limit of 150 MPa.
Figure 3. Stress analysis results on the hot-leg piping.
Mod. 9Cr-1Mo steel is used as a high-temperature structural material; its low thermal expansion coefficient coupled with better monotonic tensile and creep strengths at elevated temperature favored the application of this material to JSFR primary system. It is well known that the Mod. 9Cr material generally shows a cyclic softening behavior. The cyclic softening effect is one of factors to determine allowable stress, while the effect alleviate secondary stress which is the dominant stress in SFR. And this effect is taken into account in JSFR design.
In France, a concept with two PPs is not recommended by safety experts (higher severity of PP failure accidents, including break of pump-diagrid pipe). But this concept is justified by safety studies in other countries (Japan, India). That means that the safety approaches of designers are not identical (probably not the same criteria on maximum temperatures of fuel cladding).
An option, linked to the loop-type reactor, is the presence of guard pipes for all primary pipes. From EDF point of view, ISI&R on the double piping seems to be a problem.
JSFR employs continuous leak monitoring on the primary pipes and periodical leak rate tests on guard pipes as regulatory inspections. Volumetric inspection on high-stress part is planned as voluntary inspection [Citation18]. For the inspection, the accesses to the area between primary and guard pipes is taken into account in the JSFR design as shown in , and inspection devices are under investigation. For the repair, repair method of hot and cold leg piping has been conceptually studied taken into account disassemble process and workspace.
Figure 4. Inspection and maintenance guide tubes on guard pipe.
From EDF viewpoint, further R&D would be necessary for the two-loop cooling system. Nevertheless, this option offers a good potential for savings in the investment cost. Therefore, it is worth continuing studies on this option.
3.3. Slit UIS
As described in Section 2.3, in order to realize a compact RV, JSFR adopted an in-vessel FHS that consists of a single rotating plug, UIS with a vertically penetrating slit, and FHM with a pantograph arm. With this UIS design, FHM with a pantograph arm can move in the slit area. And also, this type of UIS does not need to be transferred from the region above the core at refueling and contributes to reducing the RV diameter.
From EDF point of view, the assemblies under the slit that is fortunately narrow then cannot have two instrumentations – the cladding failure localizing system that does not trigger safeguard actions and the double or triple thermal instrumentation that is essential to satisfy French requirements – at the sodium outlet from each fuel assembly. The latter instrumentation is called the TRTC (rapid processing of core temperatures system) in France, and the emergency shutdown stations are controlled by TRTC. It is particularly needed for detection of the early blocked assembly detection station. Therefore, it is essential that the TRTC is provided and operative at the assemblies’ outlet. In France, the lack of blockage monitoring for some assemblies would not be accepted even if it is not imposed by a written rule.
Japanese strategy is different; although all the Monju core assemblies are provided with thermal instrumentation, the thermal instrumentation for outlet temperature detection of all assemblies is not essential and, therefore, was not adopted for JSFR. The outlet temperature instrumentations in JSFR are provided only for monitoring of core performance and of core coolant flow where the flow rates are controlled by flow rate regulating orifices structure at entrance nozzle of assemblies and by the core support structure.
It seems necessary to understand the reason for the very different positions taken in France and in Japan. It is essential to share internationally the recognitions on safety instrumentation requirement of the designers and the operators, since “thermal instrumentation” is imposed in certain countries while it is not in the other countries. In France, TRTC is essential not only for the first-of-a-kind, but also for series of SFR reactors according to the current consideration. Thus, it would be useful to list the French incidents or accidents, including accidents involving the fourth defense-in-depth level, where the first and second safety stations activation will be done by thermal instrumentation in France. Then JAEA should analyze those incidents or accidents with the JSFR safety stations, to demonstrate that the mitigation robustness is equivalent, with same hypothesis and criteria.
In EDF opinion, the slit UIS modify core exit flows and therefore make it necessary to modify the core exit thermo-hydraulics file (it is very different from the “flat jet” at the Superphénix core exit), particularly considering risks of stratification and gas entrainment.
For JSFR design, in the JAEA laboratories, the flow pattern of slit UIS, the gas entrainment at the sodium level, and the gas entrainment conditions have been revealed by the 1/10-scaled and 1/1.8-scaled water experiments [Citation19,20], and a three-dimensional calculation has shown good performance of the single dipped plate design [Citation17]. Prevention of those vortex with an optimized upper plenum design with flow control devices was basically confirmed by the 1/10 upper plenum water experiment and numerical analyses [Citation21,22].
Above all, from EDF point of view, JSFR solution is very innovative because of the presence of the slit and of the fact that the UIS is transparent, in other words it does not have the outer shell usually used on UISs. EDF is wondering whether or not crippling problems could be encountered in the design of this slit UIS.
In EDF's opinion, the lifetime of the UIS is a real problem (from experience of Phénix, Superphénix, and EFR project), with the target of 60 years.
The lifetime of the JSFR UIS has been evaluated to be 60 years with 316FR steel without thermal resistance material [Citation9]. For thermal striping at the bottom of UIS, an optimized UIS design reducing thermal striping has been proposed and experimentally demonstrated [Citation17,Citation23,Citation24].
3.4. Fuel handling
In JSFR design, FHS is simplified with advanced technologies as mentioned in Section 2.3. A spent fuel subassembly (S/A) is taken in a sodium pot and transported from RV to EVST by EVTM same as Monju. The sodium-filled fuel transfer pot in JSFR is designed to be able to contain and transfer two S/A, it is designed to reduce refueling time and thereby increase plant availability [Citation25]. Maximum decay heat rate of one discharged spent fuel subassembly after 17 days following the reactor shutdown is estimated to be 22.5 kW, then the total heat capacity of the sodium pot is 45 kW and an active cooling is not necessary during transportation from RV to EVST. The EVST has enough capacity for full core evacuation to enhance the plant's ISI&R capability.
From EDF point of view, spent S/A withdrawn two at a time is favorable for handling rate. On the other hand, for EVST, as JSFR allows for the possibility of whole core defueling (known as whole core discharge (WCD), described in the reference [Citation26]) followed by sodium draining for exceptional ISI&R purposes, the requirement becomes more severe than ever before because the unloaded assemblies will possibly be reloaded again. Therefore, the EVST must have a high capacity (a complete core).
EDF points out that, high uplift forces must be possible due to the risk of assemblies bowing, so EDF prefers not to pull assemblies with a large cantilever distance, which is difficult with a pantograph arm. From EDF point of view, different risks of jamming of this arm must be taken into account.
From JAEA point of view, JSFR pantograph FHM can handle discharge/charge load up to 24.5 kN taking into account subassembly bowing under irradiation in a restricted core (a core with a core former and a barrel). A full-scale FHM mock-up has shown that the pantograph FHM could meet design requirements including the discharge/charge load [Citation27]. Prevention and recovery from the FHM jamming has also been conceptually studied showing that the severest condition is a gripper up/down jamming. In case of gripper up/down jamming, a special device has to access the failed gripper from the RV plug [Citation28].
3.5. FAIDUS
As mentioned in Section 2.2, JSFR adopted FAIDUS; FAIDUS removal channels produce a controlled materials relocation (CMR).
In JSFR, fuel discharge of FAIDUS was analyzed by the fast reactor safety analysis code, Sn, implicit, multifield, multicomponent, Eulerian, recriticality code (SIMMER-III), and discharge capability was evaluated to be 19% providing subcritical condition [Citation29]. And key early discharge mechanisms of FAIDUS have been confirmed experimentally in the experimental acquisition of generalized logic to eliminate re-criticalities (EAGLE) tests in the impulse graphite reactor (IGR) in Kazakhstan [Citation30]. During relocation, remaining fuel in the core region starts to melt by decay heat and gradually moves downward. Assuming the gradual movement of core material, static neutronic calculations have shown that a significant reactivity insertion would be avoided [Citation13]. For long-term cooling, JSFR has a multi-layer core catcher at the bottom of the RV providing 100% fuel inventory capacity. The DEBNET code (code which can evaluate the temperature distribution of debris beds and the reactor cooling system simultaneously) analysis has shown that the JSFR core catcher can provide long-term cooling capacity of 100% relocated fuel by natural circulation [Citation31].
From EDF point of view, FAIDUS-removal channels produce a CMR effect, and this solution deserves to be evaluated in France in order to determine if it could be adopted with a very robust safety demonstration. Having said this, another solution that appears to be preferred for SFRs in France is to aim at prevention rather than reduction of the consequences, by designing a core with negative or low sodium void coefficient (CFV) [Citation32].
3.6. SASS
It is mentioned in Section 2 that JSFR adopted SASS, where SASS systems provide an advantage for accidents in which safety systems fail.
In EDF opinion, there is a risk of loss of the supporting force due to particle bonding on supporting faces, based on operating experience from the backup control rod installed on Phénix (SAC in French term), and it should be tested further with representative conditions and long durations. There was a decrease of lifting-force of Phénix SAC rod, and its cause is not completely confirmed. For Superphénix, the risk of a spurious drop on SAC rods has made ASN hostile to transients requiring inhibition of the emergency shutdown station by negative reactivity, and not causing the reactor shutdown. The large number of safety rods (17 backup control rods with SASS in JSFR) increases the probability of a spurious drop.
For JSFR, the demonstration test using the reduced-scale experimental equipment of SASS was conducted in Joyo. And the control-rod-holding stability under the actual reactor operational environment was successfully confirmed [Citation33]. The results also indicate there is no essential problem for the practical use of SASS about its operational trouble involving the unexpected drop during reactor operation. The results also demonstrate that the neutron irradiation has little influence on the SASS function. The test repeatedly conducted for the recovering functions of the driving system to reconnect and pull out the control rod showed excellent results without any single failure and the propriety of SASS design. The effectiveness of SASS for the reference core design of JSFR has been evaluated through all ATWS types and as a result, it is ensured that JSFR will have a reliable passive shutdown system [Citation34].
In fact, from EDF point of view for SASS, irradiation does not damage the system, and thermal aging of components (not only the sensing alloy, but also all parts, including the insulation material of the electromagnet, etc.) do not appear to create any problems, since the different mechanical and electrical characteristics remain good. Nevertheless, the particle bonding problem (Phénix SAC experience) would need further investigation for all systems relying on in-sodium electro-magnet. The tests undertaken in the laboratory and in the Joyo reactor give a relatively good confidence level, particularly regarding the spurious drop problem which more specifically concerns availability rather than safety. On the whole, SASS performs the same role as a sentinelle passive d'insertion d'anti-reactivite (SEPIA) (a passive shutdown system developed by the commissariat à l'énergie atomique et aux énergies alternatives (CEA) [Citation35]) which is an alternative and/or a complement to SASS systems.
4. Differences in design requirements and safety aspects
4.1. Design requirements
JSFR project was compared with a future French SFR concept by EDF in relation to, first, EDF's functional requirements for the next commercial SFR, second, safety practices in France (Superphénix) and EFR and also safety practices envisaged for future SFRs. The reference data used for the comparison work are the commercial JSFR design requirements as on February 2009 and the conceptual draft of the EDF specification for French industrial SFR as on July 2007.
It is evident that primary difference is a selected reactor type; JSFR is a loop-type and previous French SFRs were pool type. It is identified that there are some outstanding differences in design requirements. JSFR specification is focused on safety and operational performance where around 120 detailed requirements of five categories (general design, safety, plant performance, design process, and economics) are given. EDF draft specification for the industrial SFR was principally distributed on the four axes of the generation IV international forum (GIF) roadmap (sustainability, economics, safety and reliability, and proliferation resistance and physical protection (PRPP)) and 15 high-level design requirements are given. This EDF specification will be reviewed and is scheduled to be updated. In fact, the conceptual draft of 2007 has been enriched (especially by taking into account experience of previous SFRs [Citation36]) and EDF iterates another version thanks to discussions between French partners.
4.2. Safety aspects
The results of the preliminary comparison on differences in safety aspects in JSFR and French SFR are shown in . The differences in safety baseline are recognized during the comparative study and are categorized into the following eight items.
Table 3. Preliminary comparison on safety aspects.
Item 1: In JSFR design, SCCV is applied. Japanese approach considers that an earthquake can be concomitant with or can follow thermal loads due to a sodium fire (for Superphénix sodium spray fires analysis, it was assumed that the initiating earthquake is not concomitant with thermal loads due to the fire, it was concomitant only with the overpressure occurring during the first seconds). In the JSFR design, double-walled pipes and a guard vessel are adopted to significantly reduce the occurrence probability of sodium leak in the containment, so that the sodium leak accident in the containment can be excluded from DBA. The guard pipes are designed to withstand loads against double-ended break of both the primary and secondary pipes inside the containment. Therefore, sodium combustion inside the containment can be prevented in any events. In spite of the above design features, taking the piping configuration into account, hypothetical sodium leak events inside the containment are assumed and evaluated to give the load condition of beyond design basis accidents (BDBA). From the evaluation, mainly two types of sodium combustion, sodium spray fire with small leak rate (less than 10 kg/s) and sodium pool fire, were assumed. In such cases, pressure rise is not so significant and the thermal load due to temperature rise of the inner steel plate is major concern [Citation37].
In France, on the reactor plug (above reactor slab area) higher sodium leak rates than 10 kg/s are considered (the main initiator can be the fly wheel disintegration of a PP, other initiators are studied, for example drops of loads if not totally prevented). For example, for the EFR project, a large secondary sodium leak (25 kg/s) above the reactor slab had been considered as a limiting event; design extension conditions (DEC). The main initiator was the fly wheel disintegration of a PP. Since a close confinement vessel (like Superphénix dome) is a difficulty for some maintenance operations (e.g. handling large components), EDF would prefer a confinement of “remote confinement vessel” type (namely this vessel would be the reactor building, like in PWR plants), and a polar table (like in the EFR project) would be desirable to alleviate loadings on this remote confinement vessel (and it is also necessary to get very resistant secondary galleries structures to ensure low loadings on the remote confinement).
In the wake of the accident at Fukushima Daiichi NPS, JAEA is currently reconsidering the safety criteria. Concerning the safety criteria of SCCV, more severe conditions are under consideration, for instance a large sodium leakage. EDF and JAEA will continue discussion on sodium leak assumption in CV.
Item 2: JSFR cover gas pressure is selected to be 0.15 MPa (gage) to maintain the net positive suction head (NPSH) of the PP. In France, the cover gas pressure is limited based on the French statutory rule ESPN. The submission of the hydraulic tests result to ESPN is required when the pressure exceeds 0.05 MPa (relative pressure). As mentioned in Section 3.2, the French statutory rule for pressurized nuclear equipment, ESPN, was written when no SFR was planned; therefore its application to SFRs raises difficulties. It might be hoped that an adaptation of ESPN (keeping the main basis of the rule but taking into account the specificities of sodium technology) will be examined.
Item 3: As mentioned in Section 3.3, outlet temperature monitoring by thermal instrumentation for each assembly is essential to satisfy French requirements. Contrarily, the outlet temperature instrumentations in JSFR will be installed at least one at each flow region and are provided only for monitoring of core performance and of core coolant flow. EDF and JAEA will continue the discussion for mutual understanding on safety requirements of subassembly outlet temperature monitoring.
Item 4: As the double-wall tube SGs are adopted in JSFR, sodium–water reactions resulting from a break of several SG tubes are treated in JSFR. The JSFR double-wall tube SG can eliminate tube failure propagation as DBE taking into account inspection capabilities; periodical inspections with an inspection capability target on both inner and outer tubes are taken into account to keep reliable sodium–water boundary. Evaluations on sodium–water reaction showed that the maximum tube failure propagation is within the range of a single tube plus its four neighbors, based on sodium–water reaction mechanism obtained from experimental results including wastage and overheating [Citation38]. Yet PFR (UK's prototype fast reactor) experienced the sodium–water reaction in early 1987 in which there were 39 broken tubes. In EDF's opinion, considering that limitation of propagation from tube to tube by wastage or thermal overheating (or other common modes) cannot be robustly demonstrated, the French approach [Citation39] consists of studying the break of all tubes in a 150 MW SG module (and even in a 375 MW SG for a plant with a single SG per secondary loop), which seems more stringent than JSFR approach. For EFR, a sodium–air–water reaction in a SG building is considered in DEC. EDF and JAEA will continue the discussion on hypothetical assumption of SG tube break for mutual understanding and harmonization of safety standard.
Item 5: For the decay heat removal function, as mentioned in Section 2.2, there is one DRACS and two PRACS installed in JSFR, and the Japanese PSA evaluation shows that the frequency of PLOHS is lower than 10−8 per reactor year against internal events. French safety experts are questioning whether or not the redundancy of these means is sufficient.
In France, studies are carried out for additional decay heat removal function with reactor-pit circuits (avoidance of some common modes with DRACS or PRACS). For reliability of the decay heat removal function in JSFR, dampers failure common mode was studied [Citation40]. On the other hand, common failure mode at the above roof area (simultaneous occurrences of loop leakage and DRACS failure) is considered in France (situation which has to be prevented by segregation between sodium circuits on the above roof area). EDF and JAEA will continue the discussion on other common modes for decay heat removal function: sodium frost, loss of heat sink (for example, stacks of sodium/air exchangers are sensitive to diverse external hazards), human factors (errors), and so on. As the practical elimination of a loss of the decay heat removal function is a shared requirement between two countries, the consideration of common failure mode is one of a future discussion items.
Item 6: Occurrences of reactor main vessel leak followed by reactor safety vessel leak are considered in France but not in JSFR. These leaks were already taken into account for Superphénix (after 1987), but they were not considered to be simultaneous. It was considered that the safety vessel leak would appear at least a few months after the main vessel leak. EDF believes that this position could become stricter for Gen IV SFRs. From EDF's point of view, in case of a loop-type reactor, leak in the primary loop piping along with a leak in its guard pipe should be treated in DEC, even without common mode. This event is not taken into account in JSFR because all of the primary sodium piping boundaries is doubled with the leak tight guard pipe and the guard pipe even resists a hypothetical double-ended guillotine break of the primary pipe. However, a very limited leak case in the guard pipe following a leak in the pipework is under investigation in JSFR's approach taking into account leak before break (LBB) phenomenon. In the same spirit, a leak in the component tank, the outer envelope of IHX and PP component, and its double envelope should be examined in DEC with the French doctrine. In fact, in JSFR it is considered that these double leak cases are practically eliminated using arguments based on construction quality, limited defect propagation, margins, and inspections.
As mentioned previously, after the accident at Fukushima Daiichi NPS, JAEA started to reconsider the safety criteria. The double coolant boundary failure is under consideration as DEC event. Assumptions and considerations of the double coolant boundary failure could be one of future cooperative items between EDF and JAEA in order to pursue mutual understanding and safety standard harmonization.
Item 7: The mitigation approach of JSFR is focused on preventing re-criticality and assuring in-vessel retention of core materials if a CDA does occur. JSFR design adopted SASS, as a passive safety feature for enhancement of prevention capability against CDA from ATWS, and FAIDUS, which allows the molten fuel to escape quickly from the core, for elimination of severe re-criticality on conformity with in-vessel retention. On the other hand safety approach of French SFR is based on strong prevention (risk minimization) with innovative cores, for instance the CFV core [Citation32], the target being no sodium boiling (so no large energy release) in cases of unprotected loss of flow (ULOF) or unprotected loss of station supply power (ULOSSP), and a favorable natural behavior in case of control rod withdrawal. Nevertheless, for defense in depth, French approach also consists in mitigation since a large core catcher is intended for French future SFR, advanced sodium technological reactor for industrial demonstration (ASTRID). The project of ASTRID is presented in the reference [Citation41]. In JAEA opinion, the low-void core concept was studied in Japan in the past but the concept has not been adopted as a result of the safety analysis [Citation42].
In the deterministic approach, SFR-initiating events were found very different in the two countries. For example, as for CDA initiators, total instantaneous blockage (TIB) is considered in France. This accident is not considered in JSFR, where the envelope scenario is a 2/3 planar blockage of a fuel assembly in DEC. In fact, in Japan, TIB is practically eliminated by adopting proper entrance subassembly nozzle (this provision has also been adopted on previous French SFRs and will be continued on future SFRs). JSFR approach considers that the TIB is enveloped by the ULOF and the ULOF accident studies are underway. For EDF, these two last accidents have a different phenomenology and both are therefore to be studied. In fact, the consideration of CDA initiator is different between EDF and JAEA, French safety experts consider that all initiating events which could lead to a core meltdown would have to be studied a similar fashion. For local accidents, in France, it seems necessary to examine the possibility that local melting may propagate through several assemblies and then become widespread. Therefore, future discussion on the consideration should be made between two bodies. Since harmonization of CDA approach could lead broad collaboration in safety R&D, this item could be also one of future discussion for mutual understanding.
Item 8: The accident at Fukushima Daiichi NPS has showed that a correct sizing for external hazards is strongly necessary. The consideration of external hazards and a safety design approach for external events in JSFR are under review in JAEA and discussions between JAEA and EDF to compare the respective approaches will be launched.
The preliminary comparison on differences in safety aspects cleared up the very different approaches between the two countries. Furthermore, after the accident at Fukushima Daiichi NPS, consideration on some of safety criteria for JSFR are under review.
After all, as pointed out above for each items, future cooperative study on safety criteria, failure mode (common failure mode at the above roof area, double boundary failure, etc.), screening method, and design approach for external hazard and safety standard harmonization are meaningful to pursue safety standard comparison between Japanese and French sides. And the above-mentioned items could be a future cooperative study in this EDF–JAEA cooperation framework.
5. Mutual interests
One of an important concept toward operational costs reduction in JSFR is to apply rigorous ISI&R strategy. Sodium-cooled reactors have the well-known maintenance challenges; on the other hand low-pressure operation provides the capability of continuous monitoring for sodium leakage that will satisfy LBB strategy. From EDF point of view, in-service inspection (ISI), especially periodic inspections, is an important issue for it provides an alert before leak, which is an early warning before the LBB line of defense. Nevertheless, in France too, on previous SFRs, the LBB approach has always been implemented and this practice will be continued on the next SFRs. A comprehensive ISI program for JSFR has been developed. Inspection methods and frequency have been established based on the ISI program of Monju, the Japanese ISI code for the light water reactor (JSME S NA1) and ISI code for the liquid metal cooled reactor (ASME section XI division 3). In JSFR design, even the reactor system is designed compact, ISI&R capability is not affected. JSFR accommodates in-service inspection programs on the RV and in-vessel structures by allowing accesses for inspection devices. Accesses and maintenance spaces are carefully allocated in the component structural design. For some small components, a whole component removal and overhaul in the reactor building is being planned.
EDF attaches great importance to getting good knowledge of reactor state to anticipate potentially present faults and guarantee reactor safety for the following 10 years by periodic inspection. Taking into account opacity of sodium, this is an important challenge for future SFRs [Citation43].
An advanced FHS is pursued in the balance of plant (BOP) design for JSFR aimed at enhancing the economic competitiveness. The features of this FHS are composed of an in-vessel FHM to be applicable to the compact reactor system, an in-vessel fuel transfer equipment that can transport two assemblies simultaneously for a shorter period of refueling time. Various R&D works have been performed, including a full-scale FHM test in air to acquire data to confirm the basic characteristics required for the FHM design for JSFR. ISI&R approach and FHS design are also one of mutual interests then could be a future collaborative study between EDF and JAEA.
6. Conclusion
Comparative study of JSFR and requirements for the future French SFRs has been conducted by EDF, the investor–operator of future French SFR. From EDF point of view, some of the innovative concepts applied to JSFR might need further R&D, in same time the specific options, hot vessel, two-loop primary cooling system, though they need further R&D, are recognized attractive in terms of saving in construction cost, and therefore it would be desirable to continue studies on these options. For safety features, FAIDUS removal channels produce a CMR effect, and this solution deserves to be evaluated in France as one of the core safety options by determining on French-side the robustness level of this mitigation device. The reliability of SASS function was confirmed with high confidence level by the results of experiments conducted in Japan. SASS could be a prominent solution although it is still necessary to pursue further experimental investigation and demonstration.
Taken as a whole, EDF points out that the economic aspect is placed in high priority in designing JSFR, and it influences the entire system.
It was also recognized during the comparative study that there are differences in safety baseline. EDF and JAEA will continue discussion for mutual understanding on safety aspects; sodium leak assumption in CV (Section 4.2 item 1), nuclear pressure equipment rules for fast reactor to be settled (item 2), safety requirements of subassembly outlet temperature monitoring (item 3), hypothetical assumption of SG tube break (item 4), consideration of common failure mode at RV plug (item 5). And as JAEA started to reconsider the safety criteria after the accident at Fukushima Daiichi NPS, assumptions and considerations of the double coolant boundary failure (item 6), comparison of CDA approach which could lead broad collaboration in safety R&D (item 7) and screening method and design approach for external hazard (item 8) could be challenging future cooperative items between EDF and JAEA. After all, future cooperative study on above mentioned items will be meaningful to pursue safety standard comparison and harmonization between Japanese and French SFR sides. The harmonization in safety aspects works will be even more important cooperative study, as the consideration on safety criteria would be reviewed for JSFR and standard SFRs in the wake of the experiences of the accident at Fukushima Daiichi NPS.
As for mutual interests, from EDF point of view, simplified FHS for JSFR is very innovative, that consists of a single rotating plug, slit UIS, and FHM with a pantograph arm. An advanced FHS, which is pursued in the BOP design for JSFR could enhance economic competitiveness, even there seemed to exist different risks of jamming of the arm. ISI&R is one of an important concept toward operational costs reduction in JSFR. Therefore, collaborative study in an advanced FHS along with FHM system, which is pursued in the BOP design for JSFR aimed at enhancing the economic competitiveness, and ISI&R could be a future study item between EDF and JAEA.
After all, this study highlights that some of the innovative concepts in JSFR deserve to be considered by EDF's point of view. The mutual interests, such as FHS and ISI&R concepts, are pointed out. And the differences of safety aspects between the French and the Japanese clarified the necessity of mutual understanding and harmonization of safety baselines for equal and high requirement level. The collaboration between JAEA and EDF will be continued in view of mutual understanding and comparison of the particular subjects and safety standard, when this is possible.
Acknowledgements
This content includes the outcome of collaborative study between JAEA and Japan Atomic Power Company (JAPC) (as the representative of nine electric utilities, Electric Power Development Co., Ltd., and JAPC) in accordance with “The agreement about the development of a commercialized Fast Breeder Reactor Cycle System”, and the results of “Technical development program on a commercialized fast breeder reactor (FBR) plant” entrusted to JAEA by the Ministry of Economy, Trade and Industry of Japan (METI). And this content includes the outcome of collaborative study between EDF – SEPTEN and AREVA, CEA and EDF – research and development.
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