Nuclear Pressure Equipment Research Articles

Molten salt breeding blanket: Investigations and proposals of pre-conceptual design options for testing in DEMO

In Europe, the DEMO test facility for Breeding Blankets (BB) will be the next step for fusion energy after ITER. The BB is a key component facing the plasma and ensuring tritium self-sufficiency, shielding against neutrons and heat extraction for electricity production. Within the EU research program, the two concepts candidates as driver BB in DEMO consider high-pressure coolants (water or helium) and Lithium-Lead or Beryllium pebble beds as breeder. In this article, an advanced BB concept is propounded as a candidate to be tested in DEMO taking into account Molten Salt (MS) as breeder and coolant and, consequently, named Breeder and Coolant Molten Salt (BCMS) BB. BCMS BB have several advantages, which lead to their investigations. As the MS is not pressurized during operation, high requirements associated to Nuclear Pressure Equipment (ESPN / PED) could be relaxed. Furthermore, the MS magnetohydrodynamic (MHD) flow is low due to its weak electrical conductivity. Besides, MS BB have identified drawbacks such as the structural material compatibility and the MS chemistry control needed during irradiation. A bibliographic review of MS and compatible materials is conducted. Promising BCMS BB are analysed regarding the Tritium Breeding Ratio (TBR) to test the capability of such MS breeders to reach the targeted TBR. The main nuclear quantities are evaluated with various MS on a dedicated model via the TRIPOLI-4® Monte-Carlo code: TBR, nuclear heating, neutron flux, displacement damage and helium production are reported and discussed. Moreover, the thermal and thermo-hydraulic responses of the MS BB are evaluated regarding the pressure drops (ΔP), the Pumping Power (PP) and the maximum temperatures in the structures. Nonetheless, preliminary design options are investigated. Finally, the major purpose is to evaluate the feasibility of a BCMS BB and identify their advantages and drawbacks.

Advanced design of the ITER core CXRS shutter and integration into the diagnostic shield module of the Upper Port Plug No. 3

For the ITER core CXRS diagnostic, a shutter is mandatory to protect the first mirror when no measurement takes place. With the continuing design development of the diagnostic and the knowledge obtained from previous shutter prototypes, the frictionless fast shutter did undergo some design changes and manufacturing improvements without degrading the performance (opening/closing <1 s). In this paper, the major modifications of the shutter design are presented. An adjustment system was developed to allow compensation of manufacturing deviations and precise positioning of the protective blade according to the final aperture position, which is not known before the final assembly with the as-built dimensions of the port plug and diagnostic shield module (DSM). The shutter components and their adjustment systems are implemented into the shutter‘s actively cooled base structure, which in turn simplifies the remote handling procedure. For the helium circuit an activity study is presented, which allows to access the nuclear pressure equipment classification and to determine the number of required purges during the ITER lifetime. With the heat loads obtained from the known heat flow on the blade and from the neutronics analysis, a thermal FE-analysis was performed on the full shutter model by which the expected heat flow at the mechanical interfaces can be given.

Manufacturing design of typical attachments in the vacuum vessel of ITER

• This paper proposes design rules for permanent attachment to ITER’s vacuum vessel by considering regulatory requirements and code and standard. • Four design factors and related design rules are summarized. Allowable criteria for P-type damage and S-type damage are listed by applicable NDE. • As one example for application, manufacturing design of the platform is presented with its structural calculation result. The vacuum vessel is a nuclear fusion engineering equipment in ITER. One of its functions is to support various components under different loading conditions during normal and off-normal operations. In order to provide the support to these components, many attachments shall be welded on the wall of the vacuum vessel. Therefore, the manufacturing design of the vacuum vessel should take into account the requirements for permanent welded attachment on nuclear pressure equipment. However, the application of these requirements into the design of attachment is not straightforward, due to conflicted rules, complex configuration and complicated operating loads of the vacuum vessel. Manufacturing design, regulatory requirements, applicable codes and standards, key design factors and types of typical attachment in sector 6 are summarized in this paper. The latest practice of Sector 6 is considered as it is the first of a kind among sectors. By rule, design verification on this platform is introduced as one example of manufacturing design of typical attachment. The result of this paper could be applied to manufacturing design of any other attachments permanently welded to the vacuum vessel wall.

Hydrostatic pressure test of the ITER lower port stub extension for factory acceptance

To check the structural integrity of the ITER Lower Port Stub Extension (LPSE), the hydrostatic pressure test has been conducted as part of the Factory Acceptance Test (FAT). The test is conducted as complying with the RCC-MR 2007 code and the French regulations of nuclear pressure equipment (ESPN) because the LPSE is classified as a nuclear pressure equipment. Because the LPSE is double wall structure with inner shell, ribs and outer shell, the inside coolant passage is filled with water for test. Pressurization is kept until 37.8 bar using the certified deionized water. During the test, before-after pressurization, and relief the pressure, the permanent deformation is checked. To verify permanent deformation, the actual displacement is measured as applying 3D metrology method using a laser tracker on the 18 locations of the LPSE, which are specified by the structural analysis results. As a result, there is no leakage on the welded and any interface area. Especially, after pressurization, the measured displacement is appeared similar to the numerically estimated value, the maximum is 0.39 mm. Finally, after relief the pressure, no permanent deformation is occurred, so that the structural integrity of the LPSE is confirmed through the selected hydrostatic pressure test.

ITER Tokamak Cooling Water System Design Status

The tokamak cooling water system (TCWS) is the primary cooling system of the ITER tokamak machine, providing cooling water to the vacuum vessel and in-vacuum vessel components. In addition, it provides water and gas baking to its clients as well as drying them prior to maintenance activities. It is a Safety Important System that is subject to the French Order on Nuclear Pressure Equipment. The TCWS design has been modified significantly since the preliminary design. Such changes required the approval of the French Nuclear Safety Authority (ASN). The first main modification was to relocate the main equipment of the cooling loop for in-vessel components from level L4 to level L3 of the Tokamak Building in order to improve the overall building shielding capacities, and the second dealt with the ability of the TCWS to significantly reduce the mass and energy released in case of a loss-of-coolant accident (LOCA). This second improvement was required to allow a strong improvement of the vacuum vessel pressure suppression system (VVPSS), i.e., the system that prevents overpressure in the vacuum vessel in case of a LOCA. The green light for the two aforementioned modifications of the TCWS (together with the new VVPSS configuration) has been recently obtained from the ASN. Most recently, the three major subsystems of the TCWS needed for first plasma operations (to start the end of 2025) have passed the final design review process. The final design has been approved and procurement/manufacturing has started. In addition, the final design of all piping not functionally required for first plasma operation but nevertheless needed due to installation constraints has been reviewed and approved as well. A few kilometers of nuclear-grade piping have been delivered to the ITER Organization (IO) together with the main drain tanks of the system. All the other components are expected to be received on site starting May 2021. This paper gives a detailed description and status of the main TCWS subsystems needed for first plasma either because they are functionally required or because of installation constraints. The major modifications highlighted above as well as all the improvements accomplished in the final design are also detailed, together with a status on procurement and construction activities.

Convective baking test of the ITER lower port for factory acceptance

The Factory Acceptance Test (FAT) for ITER components should be complied with the RCC-MR 2007 code and French regulations of nuclear pressure equipment (ESPN) to assure quality of nuclear pressure equipment. The procedure of FAT for lower port has been set to that visual test, pressure test, baking, vacuum leak test, and final dimensional inspection. Especially, the baking is critical cleaning method to satisfy required vacuum condition. The baking condition should be satisfied with both the given ramp-up/down condition, which is 5 °C/hr, and the temperature difference of the object within 40 °C, simultaneously. The air heating and circulating furnace has been specially designed to apply convective heating and cooling method. In this study, using the mock-up of the lower port stub extension, convective baking test is performed with the maximum holding temperature of 220 °C during 24 h. As a result, the maximum temperature difference of the mock-up is appeared 15.4 °C at the ramp-down condition. Using the measured result, transient thermo-structural analysis is performed. The maximum stress of 85.8 MPa is occurred, and this is reasonable stress intensity comparing to the allowable stress 195 MPa. It is proven that applying convective baking is feasible for the baking of lower port FAT.

Detail procedure of pressure and baking test for ITER vacuum vessel lower port stub extension

The procedure of factory acceptance test for ITER vacuum vessel lower port stub extension has been set to that visual test, pressure test, baking, vacuum leak test, and final dimensional inspection as listed order. Among them, the pressure test and the baking are challengeable to meet the required conditions because of the nuclear pressure equipment. The full pressure test condition has been set to 3.78 MPa, and the baking condition has been set to 200 °C during 24 h as well. The structural and thermal analysis have been conducted to verify the structural safety and predict thermal effect under the actual conditions, respectively. As a result, the optimal test configurations have been established. During the pressure test, dimensional verifying points are set based on the structural analysis results. Moreover, the convective baking method has been regarded as the optimal way because of complex double wall structure and difficulty of heating rate control. In this paper, whole procedure of the factory acceptance test for the lower port stub extension is introduced, and detail method of pressure test and baking have been reported.

ITER vacuum vessel structural analysis completion during manufacturing phase

Some years ago, analyses were performed by ITER Organization Central Team (IO-CT) to verify the structural integrity of the ITER vacuum vessel baseline design fixed in 2010 and classified as a Protection Important Component (PIC). The manufacturing phase leads the ITER Organization domestic agencies (IO-DA) and their contracted manufacturers to propose detailed design improvements to optimize the manufacturing or inspection process. These design and quality inspection changes can affect the structural margins with regards to the Codes&Standards and thus oblige to evaluate one more time the modified areas. This paper proposes an overview of the additional analyses already performed to guarantee the structural integrity of the manufacturing designs. In this way, CT and DAs have been strongly involved to keep the considerable margins obtained previously which were used to fix reasonable compensatory measures for the lack of In Service Inspections of a Nuclear Pressure Equipment (NPE).

Qualification of phased array ultrasonic examination on T-joint weld of austenitic stainless steel for ITER vacuum vessel

Full penetration welding and 100% volumetric examination are required for all welds of pressure retaining parts of the ITER Vacuum Vessel (VV) according to RCC-MR Code and French Order of Nuclear Pressure Equipment (ESPN). The NDE requirement is one of important technical issues because radiographic examination (RT) is not applicable to many welding joints. Therefore the ultrasonic examination (UT) has been selected as an alternative method. Generally the UT on the austenitic welds is regarded as a great challenge due to the high attenuation and dispersion of the ultrasonic signal.In this paper, Phased array ultrasonic examination (PAUT) has been introduced on double sided T-shape austenitic welds of the ITER VV as a major NDE method as well as RT. Several dozens of qualification blocks with artificial defects, which are parallel side drilled hole, embedded lack of fusion, embedded repair weld notch, embedded parallel vertical notch, and so on, have been designed and fabricated to simulate all potential defects during welding process. PAUT techniques on the thick austenitic welds have been developed taking into account the acceptance criteria. Test procedure including calibration of equipment is derived and qualified through official demonstrations using qualification blocks under the supervision of ITER organization and an agreed notified body (ANB). The developed PAUT techniques show excellent detectability and applicability for ITER VV.

Starting manufacturing phase of ITER upper ports

The ITER Vacuum Vessel (VV) features upper, equatorial and lower ports. The upper and regular equatorial ports are occupied by the port plugs. Although the port design has been overall completed in the past, the design of some remaining interfaces was still in progress: in particular, the Sealing Flange package, which includes the high-vacuum seals and the plug fasteners. As the ITER construction phase has started, the procurement of the VV ports has been launched. The VV upper ports will be procured by the Russian Federation Domestic Agency. The main suppliers were selected and the manufacturing design of the first parts is in full progress now. Since the VV is classified at nuclear level N2, the design and manufacture of its components are to be compliant with the French RCC-MR code and regulations for nuclear pressure equipment in France. These regulations make a strong impact to the port design and manufacturing process.

Progress of ITER vacuum vessel

Design modifications were implemented in the vacuum vessel (VV) baseline design in 2011–2012 for finalization. The modifications are mostly due to interface components, such as support rails and feedthroughs for the in-vessel coils (IVC). Manufacturing designs are being developed at the domestic agencies (DAs) based on the baseline design. The VV support design was also finalized and tests on scale mock-ups are under preparation. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. Further modifications are required to be consistent with the DAs’ manufacturing designs. Dynamic tests on the inter-modular and stub keys to support the blanket modules are being performed to measure the dynamic amplification factor (DAF). An in-service inspection (ISI) plan has been developed and R&D was launched for ISI. Conceptual design of the VV instrumentation has been developed. The VV baseline design was approved by the agreed notified body (ANB) in accordance with the French Nuclear Pressure Equipment Order procedure.

ITER Vacuum Vessel design and construction

After implementing a few design modifications (referred to as the “Modified Reference Design”) in 2009, the Vacuum Vessel (VV) design had been stabilized. The VV design is being finalized, including interface components such as support rails and feedthroughs for the in-vessel coils. It is necessary to make adjustments to the locations of the blanket supports and manifolds to accommodate design modifications to the in-vessel coils. The VV support design is also being finalized considering a structural simplification. Design of the in-wall shielding (IWS) has progressed, considering the assembly methods and the required tolerances. The detailed layout of ferritic steel plates and borated steel plates was optimized based on the toroidal field ripple analysis. A dynamic test on the inter-modular key to support the blanket modules was performed to measure the dynamic amplification factor (DAF). An R&D program has started to select and qualify the welding and cutting processes for the port flange lip seal. The ITER VV material 316 L(N) IG was already qualified and the Modified Reference Design was approved by the Agreed Notified Body (ANB) in accordance with the Nuclear Pressure Equipment Order procedure.

Classification of ITER Tokamak Cooling Water System in Accordance with French Regulations Concerning Pressure and Nuclear Pressure Equipment

ITER is a joint international fusion facility to demonstrate the scientific and technological feasibility of fusion power for future commercial electric power facilities. ITER is being designed and constructed in France with support from seven domestic agencies. In accordance with the Article 14 of the ITER Agreement, ITER shall observe French Regulations. Among various existing regulatory documents the French Decree 99-1046 concerning pressure equipment and the French Order dated 12th December 2005 concerning nuclear pressure equipment formulate the requirements for design, manufacture and operation of the pressure and nuclear pressure equipment.The ITER Tokamak Cooling Water System (TCWS) is comprised of 4 primary heat transfer systems and their supporting systems. TCWS provides the cooling water to client systems for heat removal during plasma operations and provides the primary confinement for the radioactive substances entrained in the cooling water. The main sources of radioactive substances include Tritium, Activated Corrosion Products (ACP), 14C isotope, 16N and 17N isotope. The concentration of these radioactive substances is a key parameter for the classification of TCWS equipment in accordance with French regulations.The paper will describe the process for classifying TCWS pressure equipment in accordance with French Regulations.

ITER’s Tokamak Cooling Water System and the Use of ASME Codes to Comply with French Regulations for Nuclear Pressure Equipment

During inductive plasma operation of ITER, fusion power will reach 500 MW with an energy multiplication factor of 10. The heat will be transferred by the Tokamak Cooling Water System (TCWS) to the environment using the secondary cooling system. Plasma operations are inherently safe even under the most severe postulated accident condition - a large, in-vessel break that results in a loss-of-coolant accident. A functioning cooling water system is not required to ensure safe shutdown.Even though ITER is inherently safe, TCWS equipment (e.g., heat exchangers, piping, pressurizers) are classified as safety important components. This is because the water is predicted to contain low-levels of radionuclides (e.g., activated corrosion products, tritium) with activity levels high enough to require the design of components to be in accordance with French regulations for nuclear pressure equipment, i.e., the French Order dated 12 December 2005 (ESPN). ESPN has extended the practical application of the methodology established by the Pressure Equipment Directive (97/23/EC) to nuclear pressure equipment, under French Decree 99-1046 dated 13 December 1999, and Order dated 21 December 1999 (ESP). ASME codes and supplementary analyses (e.g., Failure Modes and Effects Analysis) will be used to demonstrate that the TCWS equipment meets these essential safety requirements.TCWS is being designed to provide not only cooling, with a capacity of approximately 1 GW energy removal, but also elevated temperature baking of first-wall/blanket, vacuum vessel, and divertor. Additional TCWS functions include chemical control of water, draining and drying for maintenance, and facilitation of leak detection/localization. The TCWS interfaces with the majority of ITER systems, including the secondary cooling system.U.S. ITER is responsible for design, engineering, and procurement of the TCWS with industry support from an Engineering Services Organization (ESO) (AREVA Federal Services, with support from Northrop Grumman, and OneCIS). ITER–International Organization (ITER-IO) is responsible for design oversight and equipment installation in Cadarache, France. TCWS equipment will be fabricated using ASME design codes with quality assurance and oversight by an Agreed Notified Body (approved by the French regulator) that will ensure regulatory compliance.This paper describes the TCWS design and how U.S. ITER and fabricators will use ASME codes to comply with EU Directives and French Orders and Decrees.

Design finalization and start of construction of ITER vacuum vessel

The vacuum vessel (VV) design is being finalized including interface components, such as the support rails and feedthroughs of coils for mitigation of edge localized modes (ELM) and vertical stabilization (VS) of the plasma (ELM/VS coils). It was necessary to make adjustments in the locations of the blanket supports and manifolds to accommodate the design modifications in the ELM/VS coils. The lower port gussets were reinforced to keep a sufficient margin under the increased VV load conditions. The VV support design is being finalized as well, with an emphasis on structure simplification. The design of the in-wall shielding (IWS) has progressed, considering assembly and required tolerances. The layout of ferritic steel plates and borated steel plates will be optimized based on on-going toroidal field ripple analysis. The VV instrumentation was defined in detail. Strain gauges, thermocouples, displacement meters and accelerometers shall be installed to monitor the status of the VV in normal and off-normal conditions to confirm all safety functions are performed correctly. The ITER VV design was preliminarily approved, and the VV materials including 316L(N) IG were already qualified by the Agreed Notified Body (ANB) according to the procedure of Nuclear Pressure Equipment Order.

Design finalization and material qualification towards procurement of the ITER vacuum vessel

Procurement arrangements for ITER key components including the vacuum vessel (VV) have been signed and the ITER activities are now fully devoted towards construction. Final design reviews have been carried out for the main vessel and ports. One of the design review topics is the selection of materials, material procurement, and assessment of material performance during operation. The width of the inner shell splice plates was increased from 120mm to 160mm to minimize risk during the assembly of the Thermal shields and the VV. Instead of facet shaping, 3D shaping was introduced for the outboard inner shell. The material qualification procedures have been started for VV structural materials such as 316L(N) IG for licensing as a nuclear pressure equipment component. In accordance with the regulatory requirements and quality requirements for operation, common material specifications have been prepared in collaboration with the domestic agencies.

ITER vacuum vessel design and construction

According to recent design review results, the original reference vacuum vessel (VV) design was selected with a number of modifications including 3D shaping of the outboard inner shell. The VV load conditions were updated based on reviews of the plasma disruption and vertical displacement event (VDE) database. The lower port gussets have been reinforced based on structural analysis results, including non-linear buckling. Design of in-vessel coils for the mitigation of edge localized modes (ELM) and plasma vertical stabilization (VS) has progressed. Design of the in-wall-shielding (IWS) has progressed in details. The detailed layout of ferritic steel plates and borated steel plates is optimized based on the toroidal field ripple analysis. The procurement arrangements (PAs) for the VV including ports and IWS have been prepared or signed. Final design reviews were carried out to check readiness for the PA signature. The procedure for licensing the ITER VV according to the French Order on Nuclear Pressure Equipment (ESPN) has started and conformity assessment is being performed by an Agreed Notified Body (ANB). A VV design description document, VV load specification document, hazard and stress analysis reports and particular material appraisal were submitted according to the guideline and RCC-MR requirements.

Codes and standards and regulation issues for design and construction of the ITER mechanical components

A coherent set of Codes and Standards (C&S) for the ITER structures, system and components is needed to fulfil the safety and licensing requirements, and to provide the specified quality of the mechanical components manufactured by the different ITER Parties. In general the basic regulatory documents do not specify particular design and manufacturing codes, but for pressure equipment general Essential Safety Requirements are formulated and the means to fulfil them from a technical and legal point of view are given. This paper presents requirements for the ITER pressure and nuclear pressure equipment. It describes the selection of Codes and Standards for the mechanical components. Available industrial codes and standards (e.g. ASME, RCC-MR, and European standards) are selected for ITER components. For components not covered by conventional codes, specific structural design criteria and technical specifications have been developed. The remaining issues discussed are related to the completion of the specific C&S and the assessment of conformity with regulatory requirements before and during construction of the components.