Generation m-Power design (180 MWe per module) has developed its integrated systems test (IST) facility in Bedford County, Virginia, representing a scaled prototype of the Babcock and Wilcox (B&W) m-Power reactor. All of the technical features of the B&W m-Power LW-SMR are included in the IST with the heat source being electricity versus nuclear. The overall objective of the IST is to conduct a three-year

Cooling

pool Containment volume ^

Reactor Vent pressure lines vessel

testing program to collect data to verify the reactor design and safety performance, supporting B&W’s licensing activities with the NRC. More specifically, the IST will provide data and information on the integrated system performance but also the steam generator and other key components, evaluation model development, control and protection systems development, and operating procedures and training development. Table 14.7 presents some of the key IST testing features [16].

16.2.1 Development of research reactors

In the 1950s Argentina decided to start the design and construction of research reactors. The main purpose was the production of radioisotopes, but this also was very important for the development of engineering and construction capabilities. The RA-1, an Argonaut-type critical facility, was entirely constructed in the country

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following a basic design provided by Argonne National Lab and became critical in January 1958. Later it was redesigned locally as a 40 kW research reactor. RA-1 research reactor fuel was designed and produced in the country.

In the 1960s the Atomic Energy Commission of Argentina (CNEA) designed RA-3, a 5 MW open tank reactor for radioisotope production. The reactor became critical in 1967 and its construction had a very important local participation. RA-3 was also a milestone in achieving expertise in the field of research reactors. In the late 1970s CNEA exported to Peru a 10 MW research and radioisotope production reactor.

INVAP was created in 1976, and in the following years became involved in many technological projects: uranium enrichment, research reactors, etc. A research and training reactor (RA-6) was designed and built by INVAP in Bariloche, Argentina. INVAP has also built the research reactors NUR in Algiers, Algeria, ETRR-2 reactor in Cairo, Egypt, and OPAL in Lucas Heights, Australia. The company has participated in the construction of the RP-10 reactor in Peru.

Increased axial and radial complexity in fuel designs becomes even more relevant in those iPWRs that rely on control rods or heavy BP loadings rather than soluble boron in the coolant to control the excess reactivity. Unlike large PWRs where there may be a control rod located in one out of three or four assemblies, the iPWRs have a much higher density of control rods, particularly if the control rods are the means by which the excess reactivity is controlled. For example, in the mPower and SMR-160 designs, every assembly contains a control rod. In a large PWR where boron is used to control the excess reactivity, the control rods are normally fully withdrawn. But in some of the iPWRs, these control rods, and the sequence in which they are maneuvered, allows for compensation in the reduction in excess reactivity (as the fissile material depletes), for axial power changes, and for xenon variations (radially and axially) — see Table 4.1. The removal of the need for boron simplifies the chemical volume control systems and results in capital and operational cost savings, e. g., waste handling of depleted soluble boron.

However, at the same time, inserting a control rod moves the power away from its location to somewhere else in the core, and excessive movements can thermally ‘cycle’ the fuel, potentially resulting in fuel failures. There is also a need to control the rate of withdrawal due to fuel performance (such as pellet clad mechanical interaction, or thermal cycling) and for reactivity insertion limits (as explained in Section 4.2). Therefore, in terms of nuclear design, this leads to the need not only to optimize the fuel and core design, but also to develop a control rod sequence for the various control rod groupings to ensure appropriate compensation in reactivity, while at the same time, also ensuring other limits, such as changes in reactivity or power ramp rates for the fuel rods, are not violated. This development of the control rod sequence includes:

• determining which control rods get moved together, in so-called ‘control rod banks’;

• when they get moved in or out, how far and at what rate;

• what the overlap is between one bank and the next one starting to move in/out.

Even with the use of soluble boron, similar challenges regarding control of excess reactivity are seen for those iPWRs looking to achieve long cycle lengths between refueling outages, as greater fissile loadings are required to achieve the cycle length of greater than 24 months.

Using control rod insertion throughout a cycle results in strongly varying axial power shapes; inserting control rods in an iPWR will skew the power to the bottom of the core and result in higher peaking factors. This effect is mitigated by using higher burnable poison loadings in the lower parts of the fuel, or varying the fissile enrichment axially. In addition to the axial power effects, those assemblies that have the control rods inserted during the cycle tend to have their peak pin powers pushed radially to the edge of the assembly, i. e., away from the inserted control rod locations. Therefore, additional radial BP loadings tend to be needed to reduce the radial peaking factors. Overall, this tends to lead to a heterogeneous nuclear design in terms of fuel assemblies, and areas of notable flux suppression in the fuel pins around the BPs and control rods, and results in power peaking around the inserted control rods and BP locations. The disadvantage of a more heterogeneous core is a non-uniform radial and axial depletion throughout the core, which results in either inefficient use of the fuel (as burnup limits are met in some, but not all fuel batches), or difficulty in achieving power peaking limits. Furthermore, for the iPWRs that use natural circulation (e. g., NuScale and SMR-160), the non-uniformity in the power and resulting moderator density will require further thermal hydraulic analysis to ensure sufficient cooling and an accurate prediction and coupling of the neutronic and thermal hydraulic feedback. This issue, which needs to be addressed for all natural circulation designs, is exacerbated by the fuel non-uniformity.

In the case of high gadolinium loadings being used, the iPWR’s nuclear designers will have to consider very carefully the fissile enrichment of the fuel pins carrying the gadolinium. As explained above, those pins will have a lower fissile enrichment compared to the non-BP fuel pins to assist in the licensing of the gadolinium pins, i. e., lower thermal conductivity, and they should not be limiting, considering the lack of irradiation data and fuel performance validation. This will mean that to offset the lower enrichments in these BP fuel pins, the other pins will need to have their enrichment increased correspondingly. The larger the number of gadolinium rods in the core, the greater the level of additional enrichment required in the other fuel pins; thus achieving as long an operating cycle as possible, which requires average fuel enrichments as high as possible, could become in conflict with the current licensing limit of 5 wt% U-235.

In large PWRs operating today, there is extensive experience with different BP types and loadings, as well as control rod sequences and calculation of control rod worths. However, the combination and extent of some of these does result in challenges for code predictions, and the associated validation of nuclear design tools in extremely heterogeneous cores, and regions of notable flux variation, in particular ensuring that BP and control rod worths are accurately predicted. Furthermore, with numerous axial zones in the fuel designs, and with significant control rod movements, advances in nodal methods may be required, including the need for variable size meshes to accommodate some of the nuclear designs.

BOP instrumentation provides additional measurements that are beyond the scope of NSSS instrumentation. BOP instrumentation deals with systems associated with the turbine generator and the electric power making part of the plant. The instrumentation parameters are the same (pressure, temperature, level, and flow), but the quality pedigree is not as rigid. In many cases the secondary side vendors supply the instrumentation and control required for their supplied equipment. For example, most instrumentation required for turbine generator control is provided by the supplier of the turbine generator system. In many traditional PWRs, the vendors supply the equipment and all related instrumentation and controls.

Condenser-related measurements are another BOP instrumentation category. The condenser itself requires temperature, pressure, and level measurements. Additionally, condenser cooling systems require temperature and flow measurements. The pumps and piping that recycle the condenser water into the feedwater system need flow measurements to control flow through flow control valves.

Some heating, ventilation, and air conditioning (HVAC) system measurements may also be grouped into the BOP instrumentation category. To the extent that cooling water is recycled to provide cooling for the HVAC in certain plant areas, the HVAC-related cooling systems will need to have control systems for cooling water control valves. The HVAC systems may also need to control or throttle air flow, and may require air flow or temperature sensors to facilitate this function.

It is expected that some of the same instrumentation used in traditional PWRs will be used for iPWR BOP instrumentation. In some cases, size constraints will demand a newer device or a redesigned device, but since the temperature, pressure, and radiation environments for BOP instrumentation are in general not as extreme as those for RPS and NSSS environments, the flexibility to use newer technology will allow for more instrumentation options.

Large, high-density, high-resolution, high-definition displays are already common in consumer and commercial markets and many process and manufacturing industries are also using a variety of these displays. Typical applications include multi-monitor configurations, tiled flat panels and also projection-based systems that can display images several metres wide (Ni et al., 2006). In conventional NPP control rooms the implementation of these large displays present numerous technical difficulties, mainly because of the lack of space in an area that was originally designed for large, hardwired consoles without any digital displays. This is where designers of new plants have a distinct advantage: they can design state-of-the-art control rooms without trying to retrofit advanced HSIs into cramped spaces.

Although large displays may seem an attractive option to overcome the distributed nature of information typical of older control rooms, designers need to consider that this will not necessarily address the fundamental question of ensuring that operators benefit from increased size and resolution. Vendor hype often leads designers to assume too easily that large displays will automatically outperform small ones (Ni et al., 2006). Before equipping a control room with a multitude of large displays, human factors engineers should understand under what conditions increased size and resolution may be advantageous and how they may contribute to situation awareness. In many cases a number of standard-size displays on the operator’s workstation may be more effective than a large overview display.

To evaluate the response of an SMR to proliferation, theft, and sabotage threats, analysts need to consider both technical and institutional characteristics of the SMR. The system response is evaluated using a pathway analysis method. Pathways are defined as potential sequences of events followed by actors to achieve their objectives of proliferation, theft, or sabotage.

Before analyzing pathways, it is important to define the system under consideration and identify its main elements. After identification of the system elements, it is possible to identify and categorize potential targets for each of the threats and identify pathways for those targets. The steps used to evaluate the system response are illustrated in Figure 9.3 and discussed below.

As seen, multiple SMRs on the same site may be considered as an investment option alternative to a power station based on LRs with the same overall power output. The SMR investment case bears a loss of economy of scale which may be mitigated by some specific cost benefits. These economic benefits, presented in the previous sections, are enhanced by deploying multiple units on a same site. On the construction side, learning accumulation, modularization and co-siting economies are expected to be fostered by the multiple units ‘philosophy’ and the ‘mini-serial production’ of a number of smaller and simpler plant units. In addition, design simplification is expected to further contribute to cost-reduction of SMRs, but its evaluation is strictly plant-specific and deserves further analysis and approaches.

On the investment side, the fractioning of total investment into multiple smaller batches may represent a risk mitigation factor against possible cost/time overruns and an opportunity to adapt the investment plan and the power installed rate to the market conditions. All these economic features may be summarized into an ‘Economy of Multiple’ concept that may counterbalance the ‘Economy of Scale’ philosophy, especially when uncertainty is introduced in the analysis, affecting market conditions or construction process time schedule.

Some concepts apply to the operating and decommissioning cost as well, with a loss of economy of scale to be partially recovered by the simplification of operating or dismantling procedures. SMR design simplification has a relevant impact both on construction and decommissioning costs and any economic assessment that does not take fully into account such issues tends to be very conservative against SMR.

Hot isostatic pressing (Hipping) is the consolidation and densification under high temperature and pressure of metal powder, housed within a canister that represents the final geometry desired. HIP components offer the potential for nett-shape (NS) or

near-nett-shape (NNS) component fabrication with high densification of the volume, resulting in a small metallurgical grain size and thus providing superior mechanical properties to cast or even forged components.

HIPped components have been, and still are used in the automotive, aerospace and medical industries, and some experience of HIPped components has been gained in nuclear applications

The HIP process is well-established and consists of five key stages:

1. Procurement of metal powder to desired specification. This is a fundamental quality phase in the overall fabrication route. The powder must be of the right quality, with metal particle size distribution and morphology well defined. Failure to define and attain the correct powder will result in a poor-quality HIPped product.

2. Canister modelling and fabrication, including application of deformation modelling to optimise NNS potential. This is the most labour-intensive phase of the fabrication route. Canisters are typically low-alloy steel in sheet-metal form, that are formed and welded manually. Robustness of the canister and its welds is critical — failure of the canister during the HIP cycle will result in a defective HIP cycle — so each canister undergoes extensive inspection. The adoption of the HIP process over the unit volumes presented by small reactors depends on the effective automation of canister manufacturing.

3. Loading of canister in to the HIP furnace. The canister is loaded with powder, vibrated to maximise powder-fill, evacuated then sealed prior to loading to the HIP furnace. A typical loaded HIP canister is shown in Figure 12.9.

4. HIP cycle stage — typically in excess of 96 MN/m2 (14 000 psi) and 1200 °C Cycle time, pressure and temperature are dependent upon powder material and canister geometry, but typically this phase in the process is a small number of hours.

5. Removal of canister post-HIP cycle, via acid pickle or machining. A typical finished HIP component is shown in Figure 12.10.

HIP furnaces are available worldwide, each with their own ‘working envelope’ limiting the physical size of component which can be HIPped. This has an implication for the nuclear sector where some candidate components such as large valve bodies or pressure vessels exceed even the largest HIP vessel commercially available. The largest HIP vessel in existence today is in Japan and has a working envelope of some 2 m in diameter and 4.2 m in length. Nevertheless design schemes do exist for HIP furnace working diameters in excess of 3.5 m which would suit a number of large nuclear-grade components.

The potential for HIPping of components to save on unit cost and lead-time is great, especially when compared with forged components. The most time-consuming phase of the HIPping cycle is the canister development, but a regular drum-beat of components offers the opportunity to drive-down cost through automation of the canister fabrication process. A further distinct advantage of the HIP process is the inherent repeatability and robustness in the material properties: large forged components are notoriously difficult to produce and will always contain defects and variations in grain structure from component to component. HIPped components will always have the same fine grain structure throughout the bulk of the component and possess isotropic mechanical properties.

HIPped components are yet to be realised in a civil nuclear arena, but experience of HIPped product in an industrial environment is being gained through the oil and gas sector where HIPped large-bore pipework has been introduced. In the marine sector there are further opportunities where HIPping has yet to be realised, and the UK’s naval programme is already using some HIPped components and has gained good credibility in this part of the nuclear sector.

The small reactor programme offers a huge opportunity to introduce new HIPped components to a new reactor design, which could align well with a regulatory code case approval scheme. Retrospective introduction of HIPped components is a further opportunity but this would involve the like-for-like replacement of components and the design would not change. HIPping for small reactors offers the chance to influence the design of the components to suit the HIPping process.

Conventional water desalination uses low-temperature heat to purify seawater. Three types of desalination are considered: multi-stage flash distillation, multiple effect distillation, and reverse osmosis. The multi-stage flash distillation is a process where incoming seawater is pumped to a higher pressure and heated to near boiling. Through a series of stages, the seawater pressure is decreased to generate vapor that is condensed by the incoming seawater. The multiple effect distillation process uses a steam heat source and a series of evaporators at successively lower pressures to create water. The reverse osmosis process uses membranes to separate the pure water from the brine. The conventional case would use steam from the bottoming cycle of a natural gas combined power cycle as the steam source. For the nuclear-integrated cases, the steam would come from the low-pressure turbine. As stated previously, the temperatures needed for water desalination are higher than the conventional heat rejection temperature of the power cycle. Therefore, power generation would need to be decreased to compensate, but the steam extracted from this point in the cycle would be used for the desalination process. These combined trends would result in an overall increase in the efficiency of the combined system [4].

The development of the SFR technology in Korea entered a new phase from 2007 with Korea’s participation in the Generation IV SFR collaboration project. The advanced SFR design concept that can better meet the Generation IV (Gen IV) technology goals was developed in 2010. R&D efforts were made to develop the conceptual design of the advanced SFR, focusing on the core and reactor systems, and a development of the advanced SFR technologies necessary for its commercialization and basic key technologies. To develop these advanced technologies, R&D was performed to improve the economics, safety assurance, and metal fuel performance of an SFR in the areas of safety, fuels and materials, reactor systems and the balance-of-plant (BOP).

Korea is seeking a role in the international fast reactor technology community, not only to benefit from the advanced countries experiences, but also to contribute to the international efforts to advance the fast reactor technology. To this end, KAERI, the main body responsible for the fast reactor development in Korea, is actively pursuing cooperation with foreign countries’ organizations: CIAE (China), CEA (France),

JAEA (Japan), IPPE (Russia) and ANL (USA); international organizations: IAEA and OECD/NEA; and international collaboration projects: IAEA INPRO, GIF and ISTC.