In all LWRs, control rods function to control the fission rate, or reactivity, by inserting or withdrawing neutron-adsorbing material from the reactor fuel core. Current large PWRs typically use 17 X 17 fuel assemblies that include guide tubes for 24 control rod fingers which are operated together through a spider assembly. All fuel assemblies are capable of hosting a control rod assembly, but not all fuel assemblies will be fitted with a control rod assembly in any given fuel cycle. Control rod assemblies are generally split into two groups; control groups and shutdown groups. The shutdown control rod groups are completely withdrawn from the core to provide a large source of negative reactivity to shut down the reactor in the event of an accident. The control groups are generally partially inserted into the core and are slowly withdrawn over the fuel cycle to compensate for fuel burn up.

Current large PWRs also use a second method to control reactivity by adding soluble boric acid to the reactor coolant. Boric acid is a strong neutron absorber and is referred to as chemical shim. Current large PWRs heavily borate the RCS at the beginning of a fuel cycle and slowly dilute the boron concentration over the fuel cycle in conjunction with control rod motion to maintain operating temperature.

Many iPWR designs plan to use a half-height version of proven 17 X 17 array fuel assemblies. This will allow similar spider assembly control rods to be used. The iPWR cores contain fewer fuel assemblies than the current large PWRs and will typically have a higher percentage of fuel assemblies fitted with a control rod spider assembly. In addition, some iPWR designs may fit a very high percentage of the fuel assemblies with a control rod spider assembly and opt not to implement a chemical shim for normal reactor operation. However, it is anticipated that most iPWR designs will use concentrated boron as an emergency backup to shut down the reactor in the event not all the control rods are able to be inserted into the core. Therefore, support systems for batching boric acid will be required on most iPWR designs.

5.2.2 Control rod drive mechanisms

In all current large PWR designs, the control rod drive mechanisms (CRDMs) are external to the reactor vessel above the reactor vessel head. Prior to removing the PWR head for refueling, the CRDMs are decoupled from the control rod spider apparatus. The control rods are then left in the respective fuel assemblies while the head is removed.

Some iPWR designs plan to continue the practice of using CRDMs that are external to the reactor vessel. However, because the iPWR reactor pressure vessel is much taller than current PWR pressure vessels and the iPWR pressure vessel flange is not located at the top of the pressure vessel, some design consideration will be necessary regarding how the shaft of the control rod spider is decoupled from the CRDM and protected when the upper reactor pressure vessel is removed from the lower reactor pressure vessel for refueling. Integral PWR designs planning on the use of external control rods include the SMART reactor and the NuScale reactor (Lee, 2010; NuScale, 2012).

Other iPWR designs intend to use CRDMs that are internal to the reactor pressure vessel. This will require some materials and reliability testing given the high temperature, pressure, and radiation environment inside the reactor pressure vessel. In addition, electrical cabling will be necessary that passes through the reactor pressure vessel flange to operate the control rod drive mechanisms and provide control rod position indication. Current designs planning on the use of internal CRDMs include the Generation mPowerTM reactor and the Westinghouse SMR reactor (Kim, 2010; Memmott et al., 2012). In addition, the IRIS design had planned to use internal

CRDMs. (Carelli et al, 2004). The CAREM reactor uses internal hydraulic CRDMs, which will not require electrical cabling for operation; just position indication (Mazzi, 2011).

The US Nuclear Regulatory Commission’s (NRC) review guidance on HSIs, NUREG-0700, defines the HSI as ‘that part of the nuclear power plant through which personnel interact to perform their functions and tasks. Major HSIs include alarms, information displays, controls, and procedures’ (O’Hara et al., 2002). The HSI is used to manipulate a device or system, to request and display stored data, or to actuate a single process or various pre-programmed command routines. HSIs can be organised into workstations consisting of consoles and panels and the arrangement of workstations and supporting equipment could be organised into physical work areas such as a main control room (MCR), remote shutdown station, local control station (LCS), technical support centre (TSC) and emergency operations facility (EOF). The HSI could also be characterised in terms of the environmental conditions in which the HSIs are used, including radiation, temperature, humidity, ventilation, illumination and noise.

The NUREG-0700 definition is generally valid for HSIs currently in use, but it does not take into account the latest advances in HSI hardware and software. (At the time of writing NUREG-0700 was being prepared for a revision.) The following characteristics would generally be associated with ‘advanced HSIs’.

As mentioned earlier, selecting and implementing HSIs is only one small part of the overall engineering effort of the NPP. However, the control room and the HSI

are so tightly integrated with the overall architecture of the plant and, together with the operators, play such a vital role in plant efficiency and safety that it should be treated with the same rigorous engineering discipline as all other parts of the plant. Designers therefore need to attend to all areas of analysis, design and implementation that affect human interaction with plant and systems. We know that operator performance in a control room may be equally influenced by the design of the control system software, the system architecture, the physical architecture of the control room or the design of procedures and documentation. To address these issues, it is necessary to integrate human factors considerations into the processes used for the design of the technical system.

NUREG-0711 emphasises the importance of following a formal engineering process also for all human factors elements of the plant and to provide traceable documentation for all design decisions. This requires designers to compare the final HSIs, procedures and training with the detailed description and specifications of the design to verify that they conform to the planned design resulting from the HFE design process and verification and validation activities (O’Hara et al., 2012).

This section describes the key aspects of an integrated approach to HFE that will ensure that the technology and design choices will actually serve the purpose for which they were selected.

The team should include experts in all required and relevant technical areas, as well as expertise in conducting the elicitation in an unbiased manner, with full description of the range of opinions. An example of an expert elicitation process for performing this work is provided in GenlV International Forum (2011b).

9.4.1.1 Step 4: develop a plan describing the approach and desired results (activity M)

Before undertaking this major analysis, the evaluation plan should be thoroughly developed, reviewed, and documented. In addition, the staff resources, costs, schedule, form of the results and documentation must be clearly defined. Milestones should be developed, particularly for regular reporting to sponsors. A detailed plan for the conduct and use of peer reviews is also important to ensure quality. While developing the plan and implementing the information gathering and analysis tasks, coordination with safety evaluation, safeguards and physical security work for the SMR could provide significant benefits.

Current SMR designs are not only safer and simpler, but also more secure. Security and safeguards are built into the design features including primary safety systems located underground, smaller size (target), greater redundancy and more passive features. This gives an opportunity to determine the appropriate design basis threat, develop effective emergency preparedness plans, and integrate protection measures for: (a) physical security for detection, deterrence, and defense; (b) cyber security; and (c) material control and accounting (MC&A). SMR designs are expected to integrate all aspects of security, including security staffing and size of the protected area, into final design for NRC review. Again, the NRC will consider changes to existing regulatory guidance or new guidance concerning safeguards for an SMR to support licensing of SMR designs. SMRs should have an appropriate number of security staff and size of the protected or exclusion area based on design and engineering features, and reduced reliance on human actions.

SMRs are uniquely suited to tightly coupled, integrated energy applications. SMRs are distinguished by their relatively small power production (10s to 100s of MW — electric) and design for inherent, passive safety. Plants could incorporate multiple SMR units at these production levels, such that they can more easily be sized to meet the specific end-user demand for the output streams (e. g. electricity, thermal input to a process application) or to maximize plant thermal efficiency. The smaller per-unit size offers increased flexibility for investors (lower initial capital outlay), reduces costs associated with load-balancing, eases siting and integration challenges, and ensures increased operational flexibility. The inherent, passive safety designed into SMR concepts supports the NHES goals of system safety, resiliency and environmental stewardship by minimizing the potential for negative consequences (e. g. radiological release) of a design basis or beyond design basis event.

Potential hybrid systems could utilize proven light-water reactor (LWR) technology or proposed advanced reactor technologies that would operate at higher temperature and, hence, provide higher temperature heat for non-electrical applications. Most currently operating LWRs produce on the order of gigawatts (GW) of electricity. Retrofit of existing LWRs to incorporate a non-electric output stream is considered among the potential applications of NHES technology. This option could offer an opportunity for extending the life of operating nuclear power plants that are currently experiencing the effects of competition from low-cost natural gas (potentially resulting in plant shutdown before license expiration) and increased grid penetration by subsidized renewable energy production sources. However, retrofitting an existing reactor facility could introduce significant challenges and hurdles in the relicensing process and may not be a worthwhile investment given the limited remaining plant life.

Reactor designs currently considered for NHES primarily fall into the category of SMRs (< 300 MW electric power), as these plant sizes couple well with many of the process applications considered and are an excellent fit for small-scale regional grids or isolated industrial applications that demand both thermal and electrical resources. Most currently operating electrical generation plants are less than 500 MWe in capacity, so an integrated energy system on this scale could be envisioned to replace aging plants, particularly aging coal plants that have significant CO2 emissions. Significantly smaller systems (~50 MWe) could be a good fit for NHES coupled with wind generation, as this is the approximate capacity of most individual wind farms. Because individual wind farms are affected by regional storm systems, the collective change in generation at the grid level can be on the order of GW in several tens of minutes. This magnitude of fluctuation has challenged the ability of natural gas-fired co-generation plants to accommodate [2].

Many of the SMR plant concepts would ultimately incorporate multiple units. For such an implementation, additional capacity can be added incrementally, with individual units built in phases as necessary to meet growth in market demand. These units could be operated independently or in concert as a group depending on the overarching control strategy. Modular build-out improves the financial investment profile of the overall project, where the plant owner could choose to first construct the basic plant units (e. g. nuclear generation, energy conversion system, and electrical power generation) to establish a revenue stream while the remainder of the plant is completed, building in the necessary interconnection points and control system structure to allow later addition of additional generation sources (e. g. renewable energy system or additional nuclear units) and thermal energy applications.

Small-scale, modular reactors incorporate significantly smaller components than large-scale plants, such that they can be factory-built. Large system components for traditional, large-scale baseload nuclear plants are often built on-site and are reliant on foreign suppliers. SMR component factories could utilize a domestic supply chain and could be sited very near to the intended plant site, or components could be easily transported to the intended plant location. One might even envision a future hybrid system implementation powering a domestic SMR component factory.

Modular construction also allows alternate operational scenarios and integrated system control strategies than would be possible for a hybrid system that incorporates a single large-scale nuclear plant. In a multi-unit plant in which each of those units provides a modest amount of thermal energy input, some of the input units could be dedicated to a particular output application. Other units could then be designated as ‘swing plants’ that switch output between applications as necessary based on customer demand, economic factors, required maintenance or refueling activities, etc.

The siting of an SMR plant having one or more nuclear units is significantly more flexible than traditional large-scale plants. The possibility to locate SMRs in densely populated regions (due to reduced exclusion zone) introduces the opportunity to site the plant closer to the final customer. The estimated US land availability (suitability) for small-scale versus traditional large-scale nuclear plants is discussed in Section 13.4.3. In a hybrid implementation, siting flexibility translates to siting of the industrial heat-use application near those population centers as well. By producing non-electricity products (heat, chemicals, etc.) near the point of use, the economical attractiveness of the planned facility is increased and the market size is enlarged (particularly as aging coal plants require replacement).

Smart grids could enable the implementation of smaller input sources, such as SMRs, by balancing the load dynamics at a local scale rather than at the large scale required by traditional, large-scale nuclear plants. In this case, SMR plants could be located based on other (non-grid) subsystem requirements. Siting could be in the vicinity of the process feedstock resource (e. g. coal, natural gas, biomass), near the end-user (e. g. local community or commercial industry), or near the coupled renewable input source. Such siting would reduce transport distances for both electricity and thermal energy, thereby minimizing transmission losses. Hence, SMRs offer operational flexibility by introducing a broad range of production opportunities and simplified coupling to renewable sources and to more process applications than large-scale nuclear implementations.

Multiple deployment opportunities can be envisioned for nuclear hybrid energy systems, particularly those utilizing small modular reactors. Early implementations might provide electricity and thermal energy for independent industrial implementations without an intention to connect to the main electrical grid, allowing the system to be optimized based only on internal energy demands that are likely more predictable than external demand from the grid. Alternately, an early hybrid energy park could provide electricity and heat to small, remote communities that currently rely on diesel power that must be trucked in to the region. Later implementations might integrate the hybrid system directly to the large-scale grid, while internally managing the thermal and electrical energy resources to meet the grid demand and maximize economic return. Considerations for potential hybrid system architectures are discussed throughout this chapter. The specific needs (desired commodities) of a potential customer and resources located at the intended site will aid in the design of an optimal energy system.

To this point, none of the several LW-SMR concepts under development in the US has been licensed or constructed. ‘The mission of the SMR Licensing Technical Support program is to promote the accelerated deployment of SMRs by supporting certification and licensing requirements for US-based SMR projects through cooperative agreements with industry partners, and by supporting the resolution of generic SMR issues,’ [1]. The cooperative agreement is a cost-shared arrangement between DOE and the industry partner(s). The LTS is currently planned as a six — year program, providing a total of $452 M of government funding to be matched by the industry partners.

The DOE has made two different solicitations via funding opportunity announcements for SMR concepts. The m-Power America Partnership was selected for the first award with a goal to support commercial operation of an SMR by 2022. The emphasis of this first solicitation was to complete design certification, site characterization, licensing, first-of-a-kind engineering activities, and the associated NRC safety review processes to support the 2022 goal. The NuScale Power Partnership is the second award recipient. The objective of the second solicitation was to seek out innovative and effective solutions for enhanced safety, operations, and performance beyond designs currently certified by the NRC that should be expected to realize commercial operation in the 2025 time period.

While DOE-NE is not conducting specific R&D in support of these LW-SMR designs per se, it is recognized that there is a need and basis for addressing a number of generic issues for all potential SMR licensees, including for LW — SMRs and future A-SMRs. Some of these generic issues include evaluating the following:

• SMR source terms;

• staffing requirements for operations, maintenance, and security; and

• economic potential of SMRs via improved modeling.

The DOE-NE is also engaged with the nuclear industry’s Electric Power Institute (EPRI) in preparing a utility requirements document (URD) for SMRs aimed at developing a generic set of design requirements. The URD would assist the SMR industry in developing more focused designs for commercial deployment as well as an enhanced understanding of the licensing requirements early in the design process.

Since the SMART fuel is almost identical to the 17 X 17 standard PWR fuel except for its height, the fuel-handling equipment of SMART is similar to that of the commercial PWRs, which are currently operating in the Republic of Korea. Fuel­handling equipment provides for the safe handling of fuel assemblies and control rod assemblies (CRAs) under all specified conditions and for the required assembly, disassembly and storage of reactor vessel head and internals during refueling.

The principal design criteria specify the following:

• fuel is inserted, removed and transported in a safe manner;

• sub-criticality is maintained in all operations.

Near-term management of used fuel would be improved through the large spent fuel storage pool located in the secured fuel building for the reactors lifetime.

As outlined above there are a number of design criteria that have to be taken into account when completing the nuclear design of an iPWR. This section highlights

the major safety design criteria for PWRs (including iPWRs); the other criteria, including economics, are discussed in later sections.

As with traditional light-water PWRs, the safety analysis and regulatory requirements drive the safety instrumentation design. The regulatory requirements stem largely from the national regulations as well as several other regulation guides and referenced documents. The safety analysis is the process of evaluating the plant response to anticipated occurrences, accidents, and abnormal events to ensure the safe response of the reactor. The safety analysis defines and models the expected events, steady state outcomes, anticipated occurrences, transients and accident conditions to ensure fuel design limits are not exceeded. It defines and models the reactor vessel pressure boundary limits and the containment pressure boundary limits to ensure they are not exceeded. It also analyzes radiation conditions in order to minimize the radiation dose. The design response to the safety analysis and regulatory requirements is the called the reactor protection system (RPS) and is usually defined as the set of all reactor trips (RTs), engineered safety features (ESFs), and monitoring, required to meet all safety system analyses and regulations in order to provide safe monitoring and safe shutdown of the reactor.

For example, the need for a reactor trip when reactor pressure drops to a certain level is derived from the safety analysis requirement to protect the core from a core damage condition known as departure from nucleate boiling (DNB). This requirement drives the need to measure reactor pressure and temperature in order to know the conditions associated with DNB, and so protect the reactor from this potential accident condition.

The safety system instrumentation must be designed to protect the core from damage and to operate the reactor within safe limits as defined by the safety analysis and applicable regulations. For these reasons the following instrumentation requirements are generally required. Unique iPWR designs may preclude some of the instrumentation measurements itemized below, or may add additional instrumentation, but for the most part the following instrumentation will be required:

• pressurizer pressure and reactor pressure;

• pressurizer level;

• core temperature;

• reactor coolant temperature (wide range, and narrow range, hot leg and cold leg);

• reactor vessel level;

• steam generator level;

• reactor coolant flow;

• reactor water storage tank level;

• feedwater flow;

• main steam flow;

• main steam pressure and temperature;

• reactor power;

• core power flux (power range, intermediate range and source range);

• reactor coolant pump voltage and frequency;

• containment pressure temperature and level.

The measurement requirements listed above are used in the RPS, where RTs and ESF actuations are automatic based on sensed parameter values. Some of the sensed RPS signals are provided downstream, through isolators, to the non-safety NSSS control system, where control actions are taken in automatic or manual to keep the reactor operating within established limits. Additional safety related parameters that may be needed based on some iPWR designs are:

• safety accumulator tank pressure;

• safety accumulator tank level;

• safety valve positions;

• feedwater flow;

• boron concentration, temperature, tank levels and mixing amounts.

The following sections describe the traditional and new technological devices that are used and will be used to measure the parameters described above.