Small Fast Reactor Research Articles

BELLA: a multi-point dynamics code for safety-informed design of fast reactors

In this paper, the multi-point dynamics code BELLA and its benchmarking with respect to SAS4A/SASSYS-1 is described for a small fast reactor cooled with natural convection of lead (ELECTRA). It is shown that BELLA is capable of reproducing the magnitude of mass-flow, reactivity, power and temperature excursions during design extension conditions with an accuracy better than 10%. Hence, the BELLA code can be used for safety-informed design and stability analyses of fast reactor systems, permitting to isolate essential phenomena and trends of significance for their safety assessment.

Modular Lead-Bismuth Fast Reactors in Nuclear Power

On the basis of the unique experience of operating reactors with heavy liquid metal coolant–eutectic lead-bismuth alloy in nuclear submarines, the concept of modular small fast reactors SVBR-100 for civilian nuclear power has been developed and validated. The features of this innovative technology are as follows: a monoblock (integral) design of the reactor with fast neutron spectrum, which can operate using different types of fuel in various fuel cycles including MOX fuel in a self-providing mode. The reactor is distinct in that it has a high level of self-protection and passive safety, it is factory manufactured and the assembled reactor can be transported by railway. Multipurpose application of the reactor is presumed, primarily, it can be used for regional power to produce electricity, heat and for water desalination. The Project is being realized within the framework of state-private partnership with joint venture OJSC “AKME-Engineering” established on a parity basis by the State Atomic Energy Corporation “Rosatom” and the Limited Liability Company “EuroSibEnergo”.

Remote field eddy current testing for steam generator inspection of fast reactor

We confirmed the defect detection performances of the remote field eddy current testing (RFECT) in order to inspect the helical-coil-type double wall tube steam generator (DWTSG) with the wire mesh layer for the new small fast reactor 4S (Super-Safe, Small and Simple). As the high sensitivity techniques, we tried to increase the direct magnetic field intensity in the vicinity of the inner wall of the tube and decrease the direct magnetic field around the central axis of the tube using the exciter coil with the flux guide made of the iron–nickel alloy. We adopted the horizontal type multiple detector coils with the flux guides arrayed circumferentially to enhance the sensitivity of the radial component. According to the experimental results, the output voltage of the detector coil in the region of indirect magnetic field increased more than 100 times by the application of the exciter and detector coils with the flux guides. Finally, we were able to detect the small hole defect of 1 mm in diameter and 20% of the outer tube thickness in depth over the wire mesh layer by the adoption of the exciter coil and horizontal type multiple detector coils with the flux guides. We also confirmed that the RFECT probe is useful for detecting thinning defects. These experimental results indicated that there is the possibility that we can inspect the double wall tube with the wire mesh layer using the RFECT.

ICONE19-43422 EXPERIMENTAL STUDY ON THERMAL INTERACTION OF ETHANOL JETS IN HIGH TEMPERATURE FLUORINERT

As a fundamental study for the direct contact heat exchange which was employed for in-vessel heat exchange in Pb-Bi cooled direct contact boiling water small fast reactor (PBWFR) or Steam Generator Tube Rupture (SGTR) accident in lead alloy-cooled fast reactor (LFR), ethanol jet was injected into high temperature fluorinert (FC-3283) in order to investigate the jet boiling phenomena just after volatile water contacting with the high temperature continuous lead alloy liquid. Two series of tests (no-boiling and boiling) were initiated to evaluate ethanol vapor volume which generated around the ethanol jet. With synchronized temperature measurement around ethanol jet, the overview of the boiling behavior showed that jet boiling occurred at bottom part of jet firstly and developed to the upper part within very narrow area around jet.

Validation of the sodium void reactivity effect prediction using JEFF-3.1 nuclear data

Near 60 Na void experiments performed in the zero power reactors MASURCA (CEA-Cadarache) and ZPPR (Argonne West – Idaho) have been analyzed using JEFF-3.1 nuclear data and the ERANOS-2.1 (deterministic) and TRIPOLI-4 (Monte-Carlo) codes. Some comparative calculations have been performed also using either JEFF-3.1, ENDF/B-VII.0 or JENDL-3.3 nuclear data for 23Na, as these three 23Na evaluations show marked differences. The Na void experiments have been selected to cover spectral conditions ranging from the relatively hard flux in the outer zone of a small fast reactor to the relatively soft flux in the inner zone of a large fast reactor. For in-fuel Na void patterns, there is a good agreement between ERANOS and TRIPOLI computations, while the deterministic calculations significantly underestimate the leakage component for Na void patterns in fertile regions. The agreement between ERANOS-2.1 + JEFF-3.1 predictions and experimental values is excellent for in-fuel Na void patterns in MASURCA experiments, but a significant underestimation of the leakage component occurs for in-fuel Na void patterns in ZPPR. For fertile Na void patterns, there is a clear underestimation of the leakage component, quantitatively different for MASURCA and ZPPR experiments. Variations in 23Na cross-section data also result in significant differences: ENDF/B-VII.0 and JENDL-3.3 nuclear data for 23Na increase noticeably the predicted Na void worth values with respect to JEFF-3.1 data. The three 23Na evaluations differ at high energy (>500 keV, and especially >2 MeV), and this stresses the need for accurate additional measurements in this energy range.

小型高速炉4Sの大口径主循環電磁ポンプの開発研究(<特集>第14回動力・エネルギー技術シンポジウム)

小型高速炉4Sの大口径主循環電磁ポンプの開発研究(<特集>第14回動力・エネルギー技術シンポジウム)

E218 プラント動特性解析コードCERESの検証 : 小型高速炉を対象としたSAS4A/SASSYS-1コードとの比較(原子力技術開発の概観と小型炉,OS8 軽水炉・新型炉・核燃料サイクル)

E218 プラント動特性解析コードCERESの検証 : 小型高速炉を対象としたSAS4A/SASSYS-1コードとの比較(原子力技術開発の概観と小型炉,OS8 軽水炉・新型炉・核燃料サイクル)

E217 小型高速炉4Sの開発 : 大口径高温電磁ポンプの開発研究(原子力技術開発の概観と小型炉,OS8 軽水炉・新型炉・核燃料サイクル)

E217 小型高速炉4Sの開発 : 大口径高温電磁ポンプの開発研究(原子力技術開発の概観と小型炉,OS8 軽水炉・新型炉・核燃料サイクル)

Design study on reactor structure of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR) can produce steam by direct contact of feedwater with primary Pb–Bi coolant above the core, and circulate Pb–Bi coolant by means of buoyancy of steam bubbles. The PBWFR is capable of eliminating components of the cooling system such as primary pumps and steam generators, and thereby making the reactor system simple and compact. The specifications of the PBWFR are as follows: the fuel is Pu–U nitride; the core height is 75 cm; the core diameter is 278 cm; the average burnup is 80 GWd/t; the refueling interval is 10 years; the rated electric power is 150 MWe; the rated thermal power is 450 MWt; the core outlet/inlet temperatures are 460 °C/310 °C, respectively; and the operating steam pressure is 7 MPa. The reactor structure design has been formulated, where reactor vessel sizes are 4200 mm (ID) × 19,750 mm (H), the guard vessel is a closed type, the upper structure is made of chimneys, and the core support structure is hung up. An ultrasonic flow meter is installed inside the vessel. The seismic evaluation, design of refueling procedure and cost evaluation have been performed.

Safety design of Pb–Bi-cooled direct contact boiling water fast reactor (PBWFR)

In Pb–Bi-cooled direct contact boiling water small fast reactor (PBWFR), steam is generated by direct contact of feedwater with primary Pb–Bi coolant above the core, and Pb–Bi coolant is circulated by steam lift pump in chimneys. Safety design has been developed to show safety features of PBWFR. Negative void reactivity is inserted even if whole of the core and upper plenum are voided hypothetically by steam intrusion from above. The control rod ejection due to coolant pressure is prevented using in-vessel type control rod driving mechanism. At coolant leak from reactor vessel and feedwater pipes, Pb–Bi coolant level in the reactor vessel required for decay heat removal is kept using closed guard vessel. Dual pipes for feedwater are employed to avoid leak of water. Although there is no concern of loss of flow accident due to primary pump trip, feedwater pump trip initiates loss of coolant flow (LOF). Injection of high pressure water slows down the flow coast down of feedwater at the LOF event. The unprotected loss of flow and heat sink (ATWS) has been evaluated, which shows that the fuel temperatures are kept lower than the safety limits.

ICONE15-10817 VISUALIZATION OF LEAD-BISMUTH-WATER/STEAM DIRECT CONTACT BOILING TWO-PHASE FLOW USING GAMMA-RAY RADIOGRAPHY

A concept of lead-bismuth (Pb-Bi) cooled direct contact boiling water small fast reactor (PBWFR) is proposed as safety fast breeding reactors. Gas lift pump, injecting and boiling water in a molten lead-bismuth pool, can simplify the reactor constitution and make the volume smaller. The visualization of boiling bubbles by gamma-ray radiography is useful for investigation of thermal-hydraulics in the gas lift pump. A conceptual design of a test apparatus for the visualization test is provided.

06/01194 Sodium cooled small fast long-life reactor ‘4S’: Ueda, N. et al. Progress in Nuclear Energy, 2005, 47, (1–4), 222–230

06/01194 Sodium cooled small fast long-life reactor ‘4S’: Ueda, N. et al. Progress in Nuclear Energy, 2005, 47, (1–4), 222–230

A Conceptual Design Study of a Small Natural Convection Lead-Bismuth–Cooled Reactor Without Refueling for 30 Years

A small fast reactor has the potential to be utilized as a power source applicable to diversified social needs and to reduce capital risks. At remote sites where the population is small and plants cannot be economically connected to a power grid, power sources without refueling whose capacities are <50 MW(electric) are required because the fuel transfer cost is expensive at such sites. In the present study, a small lead-bismuth–cooled core with a 30-yr lifetime has been developed, and a simple plant system without refueling has been sketched. The dimensions of the major components are determined to evaluate its economic potential. Transient analyses of anticipated-transient-without-scram events show that the design has passive safety features suitable for a remote power source.

2639 ナトリウム冷却小型高速炉4Sの炉心冷却評価 : スタック機能喪失時のRVACS性能の検討(S57-3 原子炉システムおよびその要素技術(3),S57 原子炉システムおよびその要素技術)

2639 ナトリウム冷却小型高速炉4Sの炉心冷却評価 : スタック機能喪失時のRVACS性能の検討(S57-3 原子炉システムおよびその要素技術(3),S57 原子炉システムおよびその要素技術)

1万kWeナトリウム冷却小型高速炉（４S）の除熱特性

1万kWeナトリウム冷却小型高速炉（４S）の除熱特性

The prospect of mox fuel based Pb [sbnd]Bi cooled small nuclear power reactors

Safety performance of MOX fuel based Pb Bi cooled small fast power reactors has been analyzed and discussed. Though the thermal conductivity of MOX fuel is not large relative to that of nitride or metal fuel, but by proper combination of relatively small power density and relatively large natural circulation it can compensate fuel temperature decrease with coolant temperature increase smartly during unprotected loss of flow accident. Under such condition, accident analysis discussed in this paper show that under unprotected total loss of flow (ULOF) accident the reactor can survive inherently using combination of reactivity feedback. For unprotected rod run out transient over power (UTOP) accident the MOX reactor can overcome external reactivity by smaller power increase compared to that of nitride fueled reactors case. In this case doppler feedback plays much more important role compared to radial expansion component. So the MOX fueled small power reactors discussed here can survive both UTOP and ULOF accident with still enough temperature margin.

Sodium cooled small fast long-life reactor “4S”

CRIEPI and Toshiba Corp. have been exploring to realize a small-sized nuclear reactor for the needs of dispersed energy source and multi-purpose reactor. A conceptual design of 4S ( Super- Safe, Small and Simple) reactor is proposed to meet the following design requirements: (1) All temperature feedback reactivity coefficients including whole core sodium void reactivity are negative; (2) The core integrity is secured against all anticipated transient without reactor scram; (3) No emergency power nor active mitigating system is required; (4) The reactivity core lifetime is more than 10 years. The 4S reactor is a metallic fueled sodium cooled fast reactor. A target of an electrical output is 10–50 MW. A remarkable feature of 4S is that its reactivity is not controlled by neutron absorber rods but by neutron reflectors to cope with a long core lifetime and a negative coolant void reactivity. This study includes a design consideration of 4S. Design discussions are mainly focused on various core designs to meet above requirements. A tall core active height is adopted to gain long core lifetime. An averaged fuel burn-up is tried to be increased up to 100 GWd/ton from a point of economic view. The reference 4S designs are 10 MWe (30 years core lifetime) and 50 MWe (10 years core lifetime).

Study on Pb-Bi natural circulation phenomena

For the development of 45w%Pb-55w%Bi cooled direct contact boiling water small fast reactor (PBWFR), experimental study on Pb-Bi-water direct contact boiling two-phase flow has been performed using Pb-Bi-water direct contact boiling two-phase flow loop. For stable start-up of the boiling flow operation, Pb-Bi single-phase natural circulation must be realized in a Pb-Bi flow system of the loop before water injection into Pb-Bi. The Pb-Bi flow system consists of a four-heater-pin bundle, a chimney, an upper plenum, a level meter tank, a cooler, and an electromagnetic flow meter. A stable Pb-Bi single-phase natural circulation was realized in the range of flow rate from 1.5 l/min to 4.8 l/min by heating Pb-Bi in the heater-pin bundle with a power up to 7.7 kW. The inlet and outlet temperatures of the heater bundle were in the ranges from 243°C to 278°C, and from 251°C to 278°C, respectively. The natural circulation flow was simulated analytically using one-dimensional flow model including frictional, form and drag forces. Total hydraulic head through the loop were calculated from Pb-Bi densities at measured Pb-Bi temperatures in the loop. It was found that the calculated flow rate agreed well with the measured ones, which indicated the validity of the analytical models.

Pb [sbnd]Bi-cooled direct contact boiling water small reactor

A design concept of Pb Bi cooled direct contact boiling water small fast reactor (PBWFR) has been formulated with some design parameters identified. Water is injected into hot Pb Bi above the core, and direct contact boiling takes place in chimneys. Boiling bubbles rise due to buoyancy effects, which works as a lift pump for Pb Bi circulation. The generated steam passes through separators and dryers for the removal of Pb Bi droplets, and then flows into turbines for the generation of electricity. The system pressure of 7 MPa is as the same as that of the conventional boiling water reactors (BWRs). The outlet steam is superheated by 10°C to avoid the accumulation of condensate on a Pb Bi free surface in the reactor vessel. The control rods are inserted from above, which is different from the original concept. This insertion was chosen since the seal of steam at the top of the reactor vessel is technically much easier than the seal of Pb Bi at the bottom of the reactor vessel. The electric power of 150 MWe may be the maximum which is practically possible as a small reactor with economic competitiveness to conventional LWRs. A two-region core is designed. A decrease in reactivity was estimated to be 1.5% dk/kk′ for 15 years. A fuel assembly has 271 fuel rods with 12.0 mm in diameter and 15.9 mm in pitch in a hexagonal wrapper tube. The design limit of cladding temperature is specified to be 650°C for compatibility of cladding material with Pb Bi. As a result, the Pb Bi core outlet temperature becomes 460°C. The Pb Bi temperature rise in the core is 150°C. The conditions of the secondary coolant steam are as the same as those of conventional BWRs with thermal efficiency of 33%. The core is designed to have the breeding ratio of 1.1 and the refueling interval of 15 years as a reactor with a long-life core. Direct heat exchangers (DHX), reactor vessel air cooling systems (RVACS) and guard vessel are designed.

3534 超臨界CO_2サイクルの発電プラントへの応用と現状 : 先進エネルギーシステム(S49-5 超臨界CO_2サイクル(1),S49 原子炉システムおよびその要素技術)

The Advanced Energy System Project (AESP) aims to develop commercially viable systems and technologies that provide electricity and heat such as hydrogen, methane, methanol, fuel cells using heat from small nuclear reactors, gas-turbine power plants and the sun. AESP has three objectives: (1) Save energy resources and reduce global warming gas emissions, attaining total energy utilization efficiency higher than 85% through waste heat recovery and utilization. (2) Foster a recycling society that consumes that garbage, waste wood and used paper that are produced in urban centers, along with excreta produced through farming to generate hydrogen, methane and methanol for fuel cells using waste heat and solar energy. (3) Consume trans-uranium elements produced in light water reactors as fuel for the small fast reactors, and reduce long-lived radioactive wastes or environmental loads of long term geological disposal.