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Technical Specification for a Small Integral Pressurised Water Reactor Basic Principles Simulator

Nuclear Power Technology Development Section (NPTDS)

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EXECUTIVE SUMMARY

The IAEA maintains a suite of basic principles simulators for nuclear power plants that are used for educational purposes. Upon request, these are freely distributed to its Member States and used on personal computers to aid in the understanding of reactor plant fundamentals, operational characteristics, and various approaches to reactor designs.

There is continuing growth in global interest in the development of small modular reactor (SMR) designs and technologies. One variety of reactors that is being developed in several countries is the small integral pressurised water reactor (iPWR). In this design, primary circuit components are located within the reactor pressure vessel, eliminating the need for primary circuit pipework, with the intention of enhancing safety and reliability.

In order to continue to support the interests of its Member States, the IAEA would like to add a basic principles simulator (hereinafter referred to as “the Simulator”) describing the basic operation of an iPWR to its suite of simulators. This document proposes a plant design as the basis for the Simulator. This design is specified so as to best represent typical designs of iPWR whilst using publically available information. This document then specifies functional and design requirements for the Simulator before detailing other requirements associated with its supply, including documentation, after sales support and warranties.

Several options for the supply of the Simulator are presented in Chapter 6 – a base option and several augmented features. The Contractor is invited to submit tender documentation to the IAEA with a quote for the supply of each option. The Contractor may submit a proposal for an existing product that is similar to what is being specified. In this case the proposal shall identify and highlight the differences between what is being offered and what has been specified.

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CONTENTS

EXECUTIVE SUMMARY ...... 2 CONTENTS ...... 3 1. INTRODUCTION ...... 6 1.1. BACKGROUND ...... 6 1.2. SCOPE ...... 8 1.2.1. Purpose ...... 8 1.2.2. Overview ...... 8 2. DESCRIPTION OF THE iPWR REACTOR ...... 9 2.1. OVERVIEW ...... 9 2.2. REACTOR PRESSURE VESSEL ...... 13 2.3. PRESSURE CONTROL ...... 15 2.4. SECONDARY SYSTEM ...... 15 2.5. CHEMICAL AND VOLUME CONTROL SYSTEM ...... 16 2.6. REACTOR PROTECTION SYSTEM ...... 18 2.7. EMERGENCY CORE AND CONTAINMENT COOLING SYSTEMS ...... 18 3. SIMULATOR FUNCTIONAL REQUIREMENTS ...... 22 3.1. SYSTEM BREAKDOWN OF REQUIREMENTS ...... 22 3.1.1. Overview ...... 22 3.1.2. Reactor core ...... 24 3.1.3. Reactor Coolant ...... 25 3.1.4. Steam & Feedwater ...... 25 3.1.5. Turbine and Generator ...... 26 3.1.6. Condensate cooling ...... 26 3.1.7. Containment ...... 26 3.1.8. Protection Systems ...... 27 3.2. PLANT CONTROL AND PROTECTION SYSTEMS ...... 27 3.3. OPERATING SITUATIONS ...... 30 3.4. MALFUNCTIONS AND ACCIDENTS ...... 30 4. SIMULATOR DESIGN REQUIREMENTS ...... 32 4.1. OVERVIEW ...... 32 4.2. INTERFACE ...... 32 Page 3 of 90

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4.3. SOFTWARE REQUIREMENTS ...... 41 4.4. HARDWARE ...... 42 4.5. SIMULATION CONTROL ...... 42 4.5.1. Initiation ...... 42 4.5.2. Initial Conditions ...... 42 4.5.3. Simulation Time ...... 42 4.5.4. Snapshot ...... 43 4.5.5. Backtrack and Replay ...... 43 4.5.6. Manual Parameter Variation...... 43 4.5.7. Malfunctions ...... 43 4.5.8. Plant Configuration Variations (Optional Requirement) ...... 44 5. OTHER REQUIREMENTS ...... 45 5.1. DOCUMENTATION ...... 45 5.1.1. Minimum Content of User Manual ...... 45 5.2. PROJECT MANAGEMENT ...... 46 5.3. SCOPE OF SUPPLY ...... 46 5.4. QUALITY ASSURANCE ...... 46 5.5. WARRANTY AND SOFTWARE UPGRADES ...... 47 5.6. AFTER SALES SUPPORT ...... 47 6. OPTIONS FOR TENDER ...... 48 7. MATHEMATICAL MODELS ...... 49 7.1. OVERVIEW ...... 49 7.2. REACTOR CORE ...... 49 7.2.1. Reactor spatial kinetic model (IAEA 2003) ...... 49 7.2.2. Approximation method for coupling coefficients ...... 52 7.2.3. Reactor core kinetics and decay heat model summary ...... 54 7.3. FUEL HEAT TRANSFER (IAEA 2009) ...... 60 7.4. NATURAL CIRCULATION FLOW ...... 61 7.4.1. Single-phase natural circulation ...... 62 7.4.2. Two-phase natural circulation ...... 63 7.5. PUMPED FLOW (IAEA 2003) ...... 67 7.6. PRESSURISER ...... 67 7.6.1. Basic pressuriser model (IAEA 2003) ...... 67 7.6.2. Linearised Pressuriser Model for Integral Reactor (Kuridan and Beynon 1998) ...... 71 Page 4 of 90

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7.7. STEAM GENERATOR ...... 79 7.8. FEEDWATER (IAEA 2003) ...... 80 7.9. MAIN STEAM SYSTEM (IAEA 2003) ...... 81 7.10. CONTROL AND PROTECTION SYSTEMS (IAEA 2003) ...... 83 7.10.1. Pressuriser Pressure Control System ...... 83 7.10.2. Pressuriser Level Control System ...... 85 7.10.3. Steam Generator Pressure Control System ...... 86 7.10.4. Steam Dump Control System ...... 86 7.10.5. Rod Control System ...... 86 7.11. BREAK DISCHARGE (Micro-Simulation Technology 2011) ...... 87 7.12. PIPING ...... 87 8. ACRONYMS AND DEFINITIONS ...... 89 9. BIBLIOGRAPHY ...... 90

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1. INTRODUCTION

1.1. BACKGROUND

The International Atomic Energy Agency (IAEA) has established a suite of nuclear reactor simulation computer programs to assist its Member States in education and training. The objective is to provide, for a variety of advanced reactor types, insight and practice in reactor operational characteristics and their response to perturbations and accident situations. To achieve this, the IAEA’s Nuclear Power Technology Development Section (NPTDS) arranges for the supply and development of a suite of basic principles nuclear power plant simulators which are available to Member States upon request and are intended for educational purposes. The IAEA also provides associated training material, sponsors training courses and workshops, and distributes the documentation and computer programs.

The use of basic principle simulators to aid in teaching complex system interactions can considerably improves students’ comprehension and retention of engineering course materials. In addition, the use of simulators on nuclear fundamentals-type training courses can greatly add to trainees’ understanding of reactor operation and the role of various systems, especially safety systems. The learning pyramid model, developed by the National Training Laboratories in the USA, suggests a relationship between knowledge retention rate and teaching methods. According to this model, students retain only 5% of knowledge from lectures and 10% by reading material; this increases to 75% when doing (e.g., accomplishing simulator exercises), and 90% when having to teach others.

Figure 1: Learning pyramid model (National Training Laboratories n.d.)

The IAEA’s existing suite of basic principles simulators are based on a variety of large scale, water- cooled nuclear reactor technologies. The IAEA defines a basic principles simulator as follows: “A basic principle simulator illustrates general concepts, demonstrating and displaying the fundamental physical processes of the plant. This type of simulator also serves training objectives such as providing an overview of plant behaviour or a basic understanding of the main operation modes. Such simulators may consist of complete primary and secondary circuits, sometimes with a reduced number of loops or redundancies. The scope of simulation focusses on the main systems where Page 6 of 90

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auxiliary or supporting systems may be neglected. The control room or panels very often have a fundamentally different design in comparison with conventional control room design. Other types of basic principle simulators may use video displays to illustrate fundamental processes such as neutron flux control or boiler level control.” (IAEA 1993).

The IAEA’s suite of simulators is used on personal computers to develop understanding of the various reactor designs as well as their operational characteristics. They are intended for a broad audience of both technical and non-technical personnel as introductory educational tools. The preferred audience, however, are faculty members interested in developing nuclear engineering curriculum and in training the next generation of nuclear professionals. The simulators are not expected to produce accurate results but do demonstrate realistic trends and transients in response to changes made by the user.

In recent years the IAEA has seen an increase in the participation of its Member States in its programme for the technology development of small modular reactors (SMRs). Various designs are under development in various Member States. A large number of the designs that are in development are light water cooled and moderated small integral pressurised water reactors (iPWRs). Common features of iPWR designs include modularity, passive safety systems for core and containment cooling, and integrated design – where most or all primary components are located inside the reactor vessel. In order to best support its Member States in their development and understanding of this design variant, the IAEA would like to obtain a basic principles simulator for an iPWR.

A complicating factor in the specification for such a simulator is that, due to the lack of maturity of the various iPWR designs, there are only two iPWRs with final design details (i.e. SMART and CAREM25). However, there are not any operational data. This complicates the design, verification and validation of the simulator.

Following a review of the various small iPWR designs that are under development a design has been specified as the basis for this simulator. This is similar to the Idaho National Engineering and Environmental Laboratory’s Multi-Application Small Light Water Reactor (MASLWR) which is the basis of NuScale, a multi-module iPWR with natural circulation. This particular design was selected since there is a relatively large amount of technical data that is already publically available as opposed to iPWRs that are currently being developed by industry. Its design parameters are similar to several other iPWR designs that are currently being pursued (e.g., Argentina’s natural circulation CAREM-25 prototype) and some test results from a scaled, non-fuelled, version are also available. This specification includes features which will allow the resulting simulator to be of value when considering other designs as well.

This simulator will be based on physical laws, avoiding the use of predefined transients, and shall be adequately dynamic to allow simulation of transient conditions in real time. It shall have sufficient fidelity to give realistic plant responses during both normal operating conditions and accident situations. It is desirable that the simulator is able to accommodate certain specified changes in plant configuration and design parameters to allow it to represent various different designs.

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1.2. SCOPE

1.2.1. Purpose

This document specifies the technical and functional requirements for the supply of a basic principles simulator for a small integral pressurized water reactor (iPWR). The IAEA will add this to its existing suite of basic principles simulators for use for educational purposes within its Member States. 1.2.2. Overview

Chapter 2 presents an overview of the iPWR reactor design that is to be simulated and provides a brief overview of each of its major systems.

Chapters 3 and 4 specify the functional, design and control requirements of the Simulator.

Chapter 5 details the additional requirements associated with the supply of the Simulator, which are not directly related to its design or functionality.

It is requested that the Contractor provide quotes associated with it meeting the requirements for each of six options, as specified in Chapter 6. These include a base option, then five options which provide additional features. Contractors can submit a proposal for an existing product that is similar to what is being specified. In this case the proposal shall identify and highlight the differences between what is being offered and what has been specified.

Some mathematical models are included in Chapter 7. These illustrate the kind of basic principle models which might be used by the Contractor in the development of the Simulator.

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2. DESCRIPTION OF THE iPWR REACTOR

2.1. OVERVIEW

The Simulator shall represent a small iPWR. The design consists of a primary pressure vessel which contains the reactor core, steam generator, and pressuriser; as such, there is no primary circuit pipework as is found in a loop-type conventional pressurised water reactor (PWR). Core cooling in iPWRs may be achieved by forced or natural circulation of light water within this pressure vessel.

As discussed in Chapter 1.1, the design specified in this Simulator is largely based upon the MASLWR design that was developed by Idaho National Engineering and Environmental Laboratory for the U.S. Department of Energy. This is due to the large amount of data relating to this design that is publically available. However, to allow the Simulator to provide value to a broader range of users, this specification is for a Simulator that can accommodate certain changes in design configuration. Certain parameters have been modified to make the design more representative of several design variants being pursued around the world.

Based on NPTDS’ experience with its existing suite of simulators it is considered to be feasible to develop a basic principles simulator which allows the user to modify plant parameters to adjust various aspects of the plant configuration to accommodate differences between specific designs. An example of one such variation could be the inclusion of horizontally mounted coolant pumps onto the pressure vessel, or the method of cooling the condenser cooling water. This feature is highly desirable so its feasibility shall be clearly stated by the Contractor within the tender documentation. This feature is discussed further in Chapter 4.5.8. Figure 2 presents an overview of the key systems that comprise the iPWR plant.

Table 1 presents some of the key plant design parameters. For the purposes of this Simulator, a single iPWR unit shall be considered along with its associated secondary systems. It is noted that in commercial applications several units may be installed in modular fashion.

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Figure 2: Overview of iPWR heat cycle (Modro, et al. 2003)

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Table 1: Key design parameters for iPWR General Data Reactor Type Integral Pressurised Water Reactor

Thermal Power 150 MWth

Electrical Power 45 MWe Plant efficiency 30% Primary coolant Water Secondary coolant Water Moderator Water Thermodynamic Cycle Indirect Rankine Reactor Core and Coolant Core height 1.35 m Core diameter 1.2 m Core coolant outlet temperature 308oC Primary coolant flow rate 424 kg/s System pressure 9.6 MPa

Fuel Type UO2 pellets Fuel Enrichment 4.95% Fuel assembly type Standard 17x17 PWR assembly, shortened active axial height Clad Material Zircalloy-4 Clad outer diameter 9.5mm Clad wall thickness 0.6mm Fuel pellet diameter 8.2mm Steam Supply System Steam generator type Vertical, helical tube Steam generator number 1 Tube material Thermally treated Inconel 690 Tube outer diameter 16 mm Tube thickness 0.9 mm Number of tubes 1012 Tube length 22.3 m Tube transverse pitch ratio 1.8 Tube vertical pitch ratio 1.5 Steam flow rate 67kg/s Feedwater temperature ~ 195 oC Turbine throttle conditions ~1.52 MPa / 199oC Reactor Pressure Vessel (RPV) RPV height 13.7 m RPV diameter 2.74 m Page 11 of 90

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Design pressure 11.52 MPa Containment Type Steel vessel Volume Since the Simulator represents a new and emerging technology that has not yet been built and fully demonstrated, the containment volume will need to be negotiated between the Contractor and the buyer’s technical representative during the development of the software. Turbo-generator Steam turbine type 3600 rpm, single pressure, two-flow Generator type Self-exciting three phase AC 60Hz Rated power 45MWe Condenser Type Shell and tube heat exchanger Condenser pressure 0.05 bar Feedwater pumps Type AC driven Number 2

Figure 3: Layout for a single unit plant (Modro, et al. 2003)

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2.2. REACTOR PRESSURE VESSEL The stainless steel reactor pressure vessel (RPV) contains the reactor core and steam generator. The pressuriser is formed by the inclusion of a baffle plate divider in the top of the vessel which separates the coolant liquid from the saturated water/steam above it. The baffle plate has orifices which allow flow of water in/out of the pressurizer. The control rod drive mechanisms are mounted on the upper head of the RPV. Nozzles on the upper head provide connections for the reactor safety valves, the reactor vent valves, and the primary system piping. The design of the RPV, along with some illustrative dimensions is presented in Figure 4.

The core design is based upon a scaled version of a typical PWR core, using 24 standard 17x17 fuel assemblies with a heated length of 1.35m and equivalent diameter of 1.2m. The control rods are organized into two groups: a control group, and a shutdown group. The control group is used during normal plant operation to control reactivity. The shutdown group is used during shutdown and scram events. The control rods are gravity actuated so that they will automatically drop into the core upon loss of power.

The method of core coolant circulation is one of the design parameters which it is desirable to be able to change in the Simulator. As standard, coolant flow is driven round the RPV due to buoyancy induced natural circulation resulting from temperature differentials in the water. If a forced circulation configuration is selected, the coolant flow will be driven by four horizontal canned pumps installed within the RPV.

Relatively cold water enters the core at its base. This water is then heated and flows up through the plenum. It then flows down over the steam generator (SG) secondary side tubes, in which water boils to form steam. Each steam generator is a helical-tube, once-through heat exchanger, located within the reactor pressure vessel at a suitable height above the core. Its tubes are made of thermally treated Inconel 690. Cold feedwater enters the secondary side tubes at the base of the SG and slightly superheated steam is collected from its top. The tubes are 16mm outer diameter with 0.9mm thick walls.

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Figure 4: Reactor Pressure Vessel (Modro, et al. 2003)

Figure 5: Some Reactor Pressure Vessel Piping Connections

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If the option for forced circulation within the RPV is included in this Simulator, this shall be achieved by including four canned reactor coolant pumps, mounted horizontally onto the RPV. An illustrative coast down characteristic for such a larger centrifugal pump is presented in Figure 6. If no more detailed coast down characteristics are available, transient behavior for the canned pumps may be modelled based on this, following agreement by the buyer’s technical representative.

Figure 6: Coast-down curve (Micro-Simulation Technology 2011)

2.3. PRESSURE CONTROL

The pressuriser is integral to the RPV, separated from the main RPV volume by a baffle plate. The pressuriser serves as the primary means of RPV pressure control and is designed to maintain constant reactor coolant pressure during operation. A bank of heaters is installed above the pressuriser baffle plate. Coolant pressure is increased by applying electrical power to these heaters. Coolant pressure is reduced by spraying water into the pressuriser from spray nozzles which are fed by the Chemical and Volume Control System (see Chapter 2.5).

2.4. SECONDARY SYSTEM

Figure 7 presents an overview of the main components that comprise the secondary system. Some valves and features described here are not illustrated in this diagram.

Slightly superheated steam leaves the SG(s) through an isolation valve on each SG steam line. Each line is also connected to two pressure relief valves. The steam lines join in a manifold and the steam then passes through a main steam control valve and isolation valve into a one stage low pressure steam turbine. The expansion of the steam drives the turbine which is connected to a generator. The generator produces electricity that is transmitted to the grid via a step up transformer. Steam is extracted from the turbine stages at three positions to preheat the feedwater in order to increase the efficiency of the plant.

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After leaving the turbine, the steam enters a condenser where it is cooled and enters the liquid phase. The condenser is cooled with cooling water in either a closed loop (with heat being removed via an evaporative, mechanical-draft cooling tower) or open loop (with heat being removed by water from a natural body of water). This condenser cooling loop configuration is another feature which could be variable in the Simulator. Condenser coolant pumps circulate this cooling water. After leaving the condenser, the condensate is pumped through a condensate polishing system by condensate pumps. Here, a portion (e.g. 5%) of the condensate is ‘polished’ using filters and ion exchange beds. The condensate is then pumped back to the SG by variable speed feedwater pumps – two of these are installed in parallel to provide redundancy. A feedwater control valve and an adjacent isolation valve control the feed into each SG.

A turbine bypass line is fitted to allow steam to pass straight from the SG to the condenser via a control valve and isolation valve. This is used to allow reactor cooling to be maintained in the event of a turbine trip. This means that a turbine trip does not necessitate a reactor trip. There is also a steam discharge line which allows steam to be dumped to atmosphere via steam release valves.

Figure 7: Illustration of equivalent secondary systems that are used on the NuScale design (NuScale Power 2014)

2.5. CHEMICAL AND VOLUME CONTROL SYSTEM

The Chemical and Volume Control System (CVCS) is used to treat a portion of the reactor’s primary coolant to maintain coolant cleanliness and water chemistry (for reactivity and corrosion control reasons). It does this by using a system of chemical dosing tanks, ion exchange beds and filters. The CVCS also provides the supply for the pressuriser spray that is used to control reactor pressure. Note that simulation of this system is not required. Details of this system are presented here for information only. Inclusions of a model for the CVCS system could be a feature of a future upgrade to the Simulator. Page 16 of 90

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Reactor coolant inventory is controlled by injection of additional water when the reactor coolant level is low or letdown of reactor coolant to the liquid radioactive waste system when coolant level is high.

Boron concentration in the reactor coolant system is controlled by a feed-and-bleed process. The boron concentration is varied for long term control of core reactivity. Injection pumps provide borated water or clean demineralized water that is delivered into the RPV while excess reactor coolant being is letdown to the radioactive waste system.

A detailed description of a similar CVCS is presented by Westinghouse in its Technology Systems Manual (Westinghouse n.d.). Some illustrations of a typical CVCS are presented in Figure 8 and Figure 9.

Figure 8: Overview of a typical PWR chemical and volume control system (Grove and Travis 1995)

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Figure 9: A typical chemical and volume controls system configuration (Grove and Travis 1995)

2.6. REACTOR PROTECTION SYSTEM

When the plant’s operating parameters exceed certain defined safety limits, the plants’ protection systems will trip to put the plant into a safe state. A list of these trip parameters is provided in Chapter 0.

Where control rod insertion fails, a gravity-driven liquid boron injection system provides negative reactivity to put the reactor into a safe state.

2.7. EMERGENCY CORE AND CONTAINMENT COOLING SYSTEMS

The various small reactor designs have common features relating to emergency core and containment cooling:  Automatic depressurisation system (ADS)  Passive emergency coolant injection  Passive core decay heat removal  Containment cooling system  Large heat sink

The iPWR to be modelled in this Simulator uses a design for containment and cooling systems which is similar to several designs being developed internationally. The RPV is located within a steel lined concrete containment building. This containment design has similarities to both the CAREM and SMART reactor designs, as illustrated in Figure 10 and Figure 11.

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Two independent passive decay heat removal trains remove heat from the core, through the establishment of natural circulation loops, in the case of loss of heat sink. The SG is connected to decay heat removal condensers within dedicated pools inside the containment. The inlet valve to this line is always open but the outlet valve is normally closed; therefore the condenser tubes fill with condensate. Upon activation of the decay heat removal system, the outlet valve opens. Water drains from the tubes into the SG and this draws steam into the condensers. Here it transfers its heat to the water pool. This establishes a natural circulation loop. This loop is illustrated by item 3 in Figure 10. Each train of this decay heat removal system has sufficient capacity to remove the decay heat after a shutdown from full power operation.

In the case of a loss of coolant accident (LOCA), the primary system is depressurised by the ADS. This involves valves automatically opening to vent steam from the RPV to the containment, via a suppression pool. Three pressure relief valves provide redundancy, with each being able to provide 100% of the require pressure relief capacity. As the primary system pressure drops below a critical level, a low pressure water injection automatically starts to inject borated water into the RPV. The storage tanks (two independent trains) for this water are pressurised so only flood the RPV when system pressure drops below a specified level (15 bar). This water injection maintains the water level within the RPV above the top of the core. If the water level continues to fall in the RPV, a gravity driven water injection system provides another supply of water from a tank located above the level of the core. Any water leaking from the RPV itself will be gathered within a cavity in the containment that surrounds the lower section of the RPV. This will flood over time, providing cooling for the RPV. Any water that boils to steam within the containment will condense on the steel liner of the containment. A core spray system is used to reduce containment pressure and temperature. This is an active system, with water being pumped from the containment’s suppression pool.

The containment building also contains a large water storage tank which acts as a heat sink for cooling systems. Passive decay heat removal is achieved by routing the steam from the SG to a condenser within the water storage tank. In a LOCA accident, the ADS releases steam into the containment where it condenses on the steel liner and falls into the water storage tank. A low pressure gravity driven coolant injection system feeds coolant into the RPV once core pressure drops.

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Figure 10: Illustrative containment structure, as used in CNEA’s CAREM design, based on (Ruben 2005)

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Figure 11: Illustrative containment structure, as used in KAERI’s SMART design (IAEA 2014)

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3. SIMULATOR FUNCTIONAL REQUIREMENTS

3.1. SYSTEM BREAKDOWN OF REQUIREMENTS

3.1.1. Overview

The Simulator shall represent the iPWR. An example of such a reactor is described in Chapter 2. Table 2 presents a summary of features which the Simulator shall include. Deviation from the requirements of this table shall be clearly highlighted within the tender documentation for consideration by the Buyer.

Table 2: Summary of Simulator Functional Requirements System Simulation Scope Operator Controls Malfunctions (defined further in Chapter 3.4) Reactor core  neutron flux levels over  reactor power and  reactor setback and a range of 0.001 to rate of change (input stepback fail 110% full power, 6 to control computer)  one bank of shutdown delayed neutron groups  manual control of control rods drop into  decay heat (3 groups) reactivity devices - the reactor core  Two groups of control control rods and rods will be modelled - boron a control group and a addition/removal shutdown group  reactor trip  soluble boron reactivity  reactor setback control.  reactor stepback  Xenon/Iodine changes  reactor power control system  reactor shutdown system  Boron effects  Note: Simulation of reactor startup is not required. Reactor  main circuit coolant  Pressurizer pressure  Pressurizer pressure coolant loop riser, steam control: heaters; relief valve fails open generator, minimum of spray; pressure relief  charging (feed) valve six equivalent “lumped” valve fails open reactor  pressurizer heaters coolant channels turned "ON" by  pressure control and malfunction pressure relief  Failure of a single  operating range is from reactor coolant pump zero power hot to full (if applicable) power  Failure of all reactor coolant pumps (if applicable) Steam and  boiler dynamics  feed pump on/off  all feed pumps trip

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feedwater  steam supply to turbine operation  all steam safety valves  turbine by-pass to open condenser  steam header break  extraction steam to feed heating  boiler feed system Turbine  simple turbine model  turbine trip  turbine spurious trip generator  mechanical power and  turbine run-back  turbine spurious run- generator output are  turbine run-up and back proportional to steam synchronization  turbine trip with flow bypass valves failed  speeder gear and shut governor valve allow synchronized and non- synchronized operation Condensate  Model of forced flow of  Manual control of  Condenser cooling cooling condenser cooling loop heat sink pumps trip in either an open or temperature;  Loss of condenser closed loop.  Manual control of vacuum  Simple thermodynamic condenser cooling model to allow pumps investigation of effects of changing temperature of heat sink  Simple vacuum model to show condensate depression Overall unit  fully dynamic interaction between all simulated systems  overall unit power control with reactor leading mode; and turbine leading mode  Integration of the feedwater valve control and power control.  unit annunciation & time trends  computer control of all major system functions Safety  Decay heat removal  Inadvertent initiation system system of ADS  Emergency core cooling  Inadvertent initiation system with automatic of decay heat removal depressurisation system system (ADS), gravity driven water injection, low pressure water injection and containment cooling sprays. Page 23 of 90

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Control  Pressuriser pressure  Malfunction of Systems control system pressuriser pressure  Turbine and steam control system dump control system  Reactor power and rods control system  Reactor protection system

The following sections present details of the parameters that, as a minimum, shall be available as transient outputs for each of the various systems within the iPWR plant. 3.1.2. Reactor core

Table 3 presents the parameters relating to the reactor core that shall be available as transient outputs from the Simulator. Table 3 Parameters name Display Units Permit user to change the parameter, Yes or No Reactor thermal power MW No Reactor thermal power % No Neutron power % No Reactivity Pcm No Control rod length in core % total length Yes Control rod insertion/withdrawal rate %/min Yes Control rod worth @ current position Pcm No Clad surface temperature (peak) oC No Clad surface temperature (average) oC No Average fuel temperature oC No Peak fuel (centre line) temperature oC No Reactivity (total) %dk/k No Reactivity fuel (Doppler) %dk/k No Reactivity moderator temperature %dk/k No Reactivity rod %dk/k No Reactivity soluble boron %dk/k No

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3.1.3. Reactor Coolant

Table 4 presents the parameters relating to the reactor coolant system that shall be available as transient outputs from the Simulator. Table 4 Parameters name Display Units Permit user to change the parameter, Yes or No Coolant flow rate kg/s No Coolant average temperature oC No Coolant temperature at core inlet oC No Coolant temperature at core exit oC No RPV water level % No Reactor coolant pump speed (if RPM Yes applicable) Reactor coolant pump power (if kW No applicable) Pressure MPa No Pressure setpoint MPa Yes Level setpoint % Yes Pressuriser heater power kW No Pressuriser water level % No

3.1.4. Steam & Feedwater

Table 5 presents the parameters relating to the steam and feedwater system that shall be available as transient outputs from the Simulator. Parameters marked by an † shall also be available for the steam and feedwater headers. Table 5 Parameters name Display Units Permit user to change the parameter, Yes or No Steam flow rate† kg/s No Feedwater flow rate† kg/s No Steam pressure† MPa No Steam temperature† oC No Feedwater temperature oC No Main steam control valve opening % Yes Feedwater control valve opening % Yes Steam bypass valve opening % Yes

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3.1.5. Turbine and Generator

Table 6 presents the parameters relating to the turbogenerator system that shall be available as transient outputs from the Simulator. Table 6 Parameters name Display Units Permit user to change the parameter, Yes or No Turbine power demand setpoint % Yes Turbine demand ramp rate %/min Yes Turbine speed RPM No Generator synchronisation Degrees Yes Generator load MW No Generator breaker status Open/closed Yes

3.1.6. Condensate cooling

Table 7 presents the parameters relating to the condensate cooling system that shall be available as transient outputs from the Simulator. Table 7 Parameters name Display Units Permit user to change the parameter, Yes or No Condenser coolant flow kg/s No Coolant pump power kW No Coolant temperature at condenser inlet oC No Coolant temperature at condenser outlet oC No Temperature of heat sink water (if open oC Yes loop from river, lake or ocean) Wet-bulb temperature of cooling air (if oC Yes closed loop with cooling tower) Condenser vacuum bar Yes

3.1.7. Containment

Parameters name Display Units Permit user to change the parameter, Yes or No Containment pressure MPa No Containment temperature oC No Pool level (for each pool) % of full level No Pool temperature (for each pool) oC No

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3.1.8. Protection Systems

Parameters name Display Units Permit user to change the parameter, Yes or No ADS valve flow kg/s No Boron injection system flow rate kg/s No Boron concentration in coolant ppm No Decay heat removal system flow rate kg/s No Water injection flow from low pressure kg/s No injection system Water injection flow rate from gravity kg/s No driven injection system Containment coolant spray flow rate kg/s No

3.2. PLANT CONTROL AND PROTECTION SYSTEMS

Various plant control models will need to be developed and linked in order for the Simulator to respond realistically to transients. Some illustrative plant control system mathematical models are presented in Chapter 7.10. Figure 12 and Figure 13 illustrate the various control models required in the reactor control loop and some details of the reactor power control logic, respectively.

Figure 12: Control Loops Overview (Cassiopeia Technologies Inc. 2007)

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Figure 13: Reactor Power Control Logic (Cassiopeia Technologies Inc. 2007)

The simulator models shall include a reactor protection system which causes the reactor or turbine to trip (as appropriate) when plant parameters exceed trip setpoints or if AC power is lost. Trip setpoints are specified in Table 8. Users shall be able to vary these setpoints.

As well as reactor trip (scram), the plant shall be able to curtail reactor power in response to setpoints being exceeded by two other means, under certain conditions where a full scram is not necessary.  Reactor stepback – reduction of reactor power in a large step.  Reactor setback – ramping of reactor power at a fixed rate to a setback target.

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Table 8: Trip parameters adapted from (James E. Fisher n.d.) and (Cassiopeia Technologies Inc. 2011)

Signal Trip Set Point Resulting Trip Low upper plenum pressure 8.5 MPa Reactor scram Low upper plenum level 0.5 m Reactor scram & ADS actuation Low steam generator level Reactor scram Manual scram Operator Reactor scram ADS Actuated Reactor scram Low downcomer flow 350 kg/s Reactor scram High core outlet temperature Reactor scram High reactor neutron flux 120% at full power Reactor scram High log rate Reactor scram High reactor coolant pressure Reactor stepback Loss of one coolant pump (if Loss of pump Reactor stepback applicable) Loss of two coolant pumps (if Loss of pumps Reactor stepback applicable) High zone flux If zone flux >115% of nominal Reactor stepback zone flux at full power High steam header pressure Reactor setback High pressuriser level Reactor setback Manual trip Operator Turbine trip Reactor scram Scram Turbine trip Turbine trip Tripped Feedwater pumps trip Low steam header pressure 1.2 MPa Feedwater pumps trip Manual trip Initiated by operator Feedwater pumps trip High containment pressure 0.5 MPa ADS actuation High upper plenum pressure 12 MPa ADS actuation Low upper plenum pressure 8.5 MPa ADS actuation Governor valve and bypass valve Valve indication Decay heat removal system closed actuation

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3.3. OPERATING SITUATIONS

The Simulator shall be capable of simulating the following standard operational events:

 Reactor start-up from hot-standby zero power

 Reactor power increase and decrease

 Reactor scram and restart

 Variation in speed of MCPs (if applicable)

 Operation in either turbine leading or reactor leading mode

For simplicity, simulation of reactor start up from cold is not required. Additionally, natural circulation may be assumed to already be established in the hot standby condition. This will simplify the modelling requirements.

3.4. MALFUNCTIONS AND ACCIDENTS

The Simulator shall be capable of simulating malfunctions and accidents to allow the user to understand the transient plant response for each event. The Simulator shall enable the user to be able to trigger a range of predetermined malfunctions by selecting them from a list. Table 9 presents a list of malfunctions that shall be available as standard within the Simulator. When selecting a malfunction the Simulator shall enable the user to be able to specify whether it occurs immediately, after a specified time interval, or if it is triggered by a specific event (such as a specific plant parameter reaching a specified value). It shall also be possible for the user to select to insert a ‘random’ malfunction from this list. In doing so, the user will not know which malfunction has been initiated.

Additionally, the Simulator shall enable the user to be able to cause individual components (e.g. valves or pumps) to fail completely or reduce their performance at any time during a simulation.

Table 9: Malfunctions to be included in the Simulator Number Description

Anticipated Operational Occurences

1 Reduction in feedwater temperature (loss of feedwater heating)

2 Abnormal increase in feedwater flow

3 Inadvertant opening of a steam generator relief valve

4 Inadvertant actuation of decay heat removal system

5 Loss of containment vacuum

6 Turbine trip

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7 Turbine trip with bypass valve failed closed

8 Loss of condenser vacuum

9 Closure of mainsteam isolation valve

10 Uncontrolled control rod assembly withdrawal at power

11 Loss of normal feedwater flow (feedwater pumps trip)

12 Failure of main coolant pumps (if applicable)

13 Inadvertant opening of a pressure relief valve

14 Steam generator tube failure

15 Inadvertant ECCS valve opening

16 Inadvertent operation of pressuriser heaters

17 Spurious turbine run-back

18 Condenser coolant pumps trip

19 Inadvertant reactor isolation – closure of all main steam isolation valves

20 Earthquake

Postulated Accidents

21 Major steam system piping failure within containment

22 Major steam system piping failure outside containment

23 Station blackout – loss of AC power

24 Feedwater system pipe break

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4. SIMULATOR DESIGN REQUIREMENTS

4.1. OVERVIEW

The Simulator shall be designed to run on standard Windows 7 or Windows 8 desktop computers with user interaction via a single display, keyboard and mouse. It shall feature an intuitive, user-friendly Graphical User Interface (GUI) to allow the user to focus on understanding the principles represented by the Simulator, not on its operation. Details of the proposed GUI and programming language shall be specified by the Contractor in the tender documentation.

4.2. INTERFACE

The Simulator GUI shall be as intuitive and user-friendly as is reasonably achievable. It shall be based around a series of display pages showing different plant views; an illustrative list of pages that could be included is presented in Table 10. This table was populated with features from various products within the IAEA’s suite of basic principles simulators. The use of several pages will prevent a single page from becoming cluttered. The GUI of the Simulator shall be able to resize to fit monitors of various resolutions.

An additional feature that could be included as an option is the ability to spread the various simulator mimics across several monitors and select a different display page to be shown on each monitor (optional requirement).

Table 10: Illustrative List of Display Pages Display Name Details Page Number

1 Home page List of the Simulator’s display pages with hyperlinks to them.

Schematic of the whole iPWR plant. This could be in either two or three dimensions.

2 Plant overview Simplified schematic diagram of the whole iPWR plant showing main components and flow paths.

Key plant parameters shown including:  Neutron power (% full power)  Neutron power rate (%/s)  Reactor thermal power (% full power)  Core pressure and flow  Core water level  Average coolant temperature

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 Average fuel temperature  Pressuriser pressure and level  Flow to/from pressuriser  SG steam flows, pressures and temperatures  Total steam flow through main steam isolation valve  Main steam isolation valve status  Main steam governor valve status  Condenser steam bypass valve status  Atmospheric steam discharge valve status  Generator output  Feedwater flow and temperature  Boiler feed pump status  Pressure relief valves status 3 iPWR control overview / control A schematic representation of the relationships loops between the various control models used within the Simulator.

4 Control rods and reactivity Schematic of control rod positions and locations.

Status of the control rods.

Core flow vs power map.

Reactivity contributions from the different types of rods, from poisons, from boron additions, and from variation in moderator and fuel temperature.

Schematic showing a ‘map’ of core flux intensity in its different sections.

Schematic of route that coolant takes through core.

6 Trip parameters Displays a list of parameters which could cause the reactor or turbine to trip, step back or set back. Each parameter shall be automatically highlighted as it occurs so that the reasons for a trip can be understood.

7 Reactor coolant system Schematic of the relevant areas of iPWR plant.

At the relevant locations on the diagram the following parameters shall be shown:

 Average fuel temperature  Average coolant temperature  Coolant temperature at base of core  Coolant temperature at top of core Page 33 of 90

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 ΔT (coolant temperature at top of core – coolant temperature at base of core)  Coolant flow  Coolant pressure  Feedwater flow in steam generator  Coolant flow to/from pressuriser  Steam generator pressure  Steam flow rate  Pressuriser heater status  Pressuriser vapour pressure  Pressuriser liquid level  Spray flow into pressuriser  Pressure relief flow  Coolant makeup tank level  Coolant feed/bleed flows  Valve statuses and % open values  Pressuriser level setpoint  Reactor pressure setpoint  Condenser vacuum  Status of condenser cooling pumps

Relevant trend plots shown, for example:

 Core coolant temperature  Pressuriser pressure; reactor core inlet pressure; reactor core outlet pressure  Reactor power  Coolant inventory 8 Turbine generator and condenser Schematic for relevant areas of iPWR plant.

At the relevant locations on the diagram the following parameters shall be shown:

 Steam pressure  Steam flow  Valve statuses and % open values  Steam flow to turbine  Steam bypass flow  Generator electrical output  Generator rotation speed  Generator breaker trip status  Generator synchronisation angle  Power demand (%)  Power demand rate (%/min)  Turbine trip status  Manual turbine runback and trip options.

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Relevant trend plots shown, for example:

 Reactor neutron and thermal power (%)  Generator output  Turbine steam flow; steam bypass flow  Turbine speed 9 Feedwater and extraction steam Schematic for relevant areas of iPWR plant.

At the relevant locations on the diagram the following parameters shall be shown:

 Main steam header pressure  Steam flow through governor valve  Steam flow through bypass valve  Feedwater pump statuses  Valve statuses and % open values

Relevant trend plots shown, for example:

 Reactor neutron and thermal power (%)  Main steam header pressure 10 Passive core cooling and Schematic for relevant areas of iPWR plant containment including view of containment structure.

Key parameters and valves statuses shown.

Water levels illustrated graphically in the vessels.

Temperatures illustrated using a colour scale.

A link to an explanation of the passive core cooling systems would be useful.

11 Trends An additional page which contains several plot windows that the user can use to plot trends as desired.

There are certain features that shall be displayed on each of the pages within the GUI in a consistent location:

 Simulation status – running, frozen, not initiated, replay.

 Simulation time.

 Simulation run speed.

 Simulator control buttons which are common to all pages, for example – run, freeze, stop, save, manual trip for reactor, manual trip for turbine, help, plot.

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 Key plant parameters – Reactor neutron and thermal power, generator output, main steam header pressure, pressure vessel water level.

 Malfunction selection controls – this will allow a malfunction to be initiated from any display page.

 Indicators for if the turbine or reactor has tripped and what has caused it – this could be achieved by the inclusion of alarm indicators for key alarms (e.g. reactor scram, turbine trip, malfunction active).

 Indicators for if passive safety systems have been actuated (e.g. yellow flashing light) – for decay heat removal system, automatic depressurisation system, water injection system, containment cooling sprays.

Other common features of the GUI shall be:

 Components shall have name labels next to them to identify them to the user.

 The status of all components (on/off, open/closed, malfunction) shall be clearly represented, for example using colour coding.

 The display of each component shall clearly indicate whether it is under automatic or manual control.

 Key parameters shall be shown, such as those listed in Table 10. Units shall be given for all parameters.

 The display shall clearly indicate whether a parameter or component is able to be controlled manually.

 Certain plant features or malfunctions will not always be present, such as sprays or leaks. When these occur, these shall be shown on the schematic diagrams in the relevant locations.

 Arrows shall be used to indicate flow directions on schematic diagrams.

 Parameters shall be identified by name where possible, not by code/acronym, to simplify interpretation by new users.

Where transient plots are shown within a display page, it shall be possible for the user to change their scales and ranges to allow manual inspection of trends. In any display page, it shall be possible to create a plot in a new window displaying any single, or combination of, user defined parameter(s) against time.

To make the Simulator more useful as an educational aid, it is desirable for it to provide basic information about each plant system and each component’s purpose and operation, along with the basic equations and theory for the physics model of that system or component. This will assist in the understanding of the basic principles associated with the reactor plant. It is suggested that this information might be displayed by clicking on an ‘information’ button next to each component (optional requirement).

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It is desirable for the Simulator to include a ‘Help’ feature that is easily located within the tool bar. This shall provide a basic overview of the controls of the Simulator and how to operate it. It shall include a search function so that the user can look for help by inserting key words. It could include an electronic version of the user manual.

On schematic diagrams, the containment position shall be clearly identified. For example, on a diagram which does not show the structure of the containment but does show flow paths in/out of the containment, a line shall be included to represent the containment boundary.

The units associated with each parameter shall be included in every occurrence of that parameter. Any acronyms or codes that are used shall be clearly explained in a glossary within the user manual. Some examples of the GUI from the IAEA’s existing suite of simulators are presented for illustration purposes in Error! Reference source not found. to Error! Reference source not found.19.

Figure 144 (Cassiopeia Technologies Inc. 2007)

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Figure 155 (Cassiopeia Technologies Inc. 2007)

Figure 166 (Cassiopeia Technologies Inc. 2007)

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Figure 176 (Cassiopeia Technologies Inc. 2007)

Figure 187 (Cassiopeia Technologies Inc. 2007) Page 39 of 90

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Figure 18 (Cassiopeia Technologies Inc. 2007)

Figure 19 (Cassiopeia Technologies Inc. 2009)

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4.3. SOFTWARE REQUIREMENTS

The Simulator will likely be used by individuals in many different countries, using computers of various specifications. As a minimum, the Simulator shall be capable of running in the Windows XP, Windows 7 and Windows 8 operating systems. The Simulator shall be flexible to accommodate future changes in operating systems. The Simulator shall have a GUI as discussed in Chapter 4.2.

To ensure that the Simulator is of value to as broad a range of users as possible, it would be highly desirable for the Simulator to be programmed to be platform independent and able to run on a wide range of operating systems. This could be achieved through a web browser interface that is common to various operating systems, for example (optional requirement).

The Simulator shall include facilities to allow the user to record the results and transient data from any simulation for reporting or reference purposes. The Simulator shall be capable of saving transient simulation data to standard file formats (e.g. comma separated variable, Microsoft Office Excel) so that it may be analysed by users at a later date.

The Simulator shall be capable of plotting several transients in real time as it operates. It shall be possible to plot more than one user defined simulation parameter on a single chart. Plots shall be able to be exported into Microsoft Office programs or exported as image (.png, .jpg) files. The user shall be able to manually change the scales and ranges on plots in addition to there being an automatic scaling feature (to give optimum display). The Simulator shall be capable of generating a full transient report along with any associated plots that the user may require. These reports shall show the details and time of occurrence of any user or Simulator initiated plant status changes. The Simulator shall be capable of using the computer’s printer drivers to print charts or reports directly from the simulator software. It shall also be capable of saving these to .pdf format.

Where possible, standard programming languages, communications protocols, subroutines, macros, software parts and software development tools shall be used. Exceptions shall be clearly stated in the tender submission. Any parts of the code that are hardware or operating system dependent shall also be identified explicitly. It is thought to be likely that the Simulator will employ a modular modelling system with each plant component and process being represented in a separate software module.

All software written by the Contractor shall use standard coding practices. Comments shall be used extensively to explain the operation of the program. The following items shall be considered when developing the code:

 Variables and constants must follow the naming convention as defined in the user manual’s glossary.

 Variables and constants within the code shall be appropriately defined with correct units.

 Derived or empirical values generated during development must be identified as such. The Contractor shall provide to the IAEA the underlying simulator code and data, for maintenance, development and use in other simulators.

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4.4. HARDWARE

The Simulator shall be capable of running comfortably on general purpose, desktop personal computers. Ideally at least 40% free memory and 25% spare CPU time shall be left available during operation.

4.5. SIMULATION CONTROL

4.5.1. Initiation

The Simulator shall be dynamic and model the reactor operation in real time. The Simulator shall enable the user to be able to be able to initiate a simulation by selecting a set of initial conditions (ICs) and pressing a ‘run’ button. The Simulator shall include a ‘freeze’ button that shall allow the user to pause the simulation at any point which shall also pause associated data recorders and plots. The Simulator shall include a ‘run’ button that shall restart the simulation from the point at which it was paused and a ‘stop’ button shall invite the user to save the Simulator’s data records and plots before exiting the Simulation.

4.5.2. Initial Conditions

The Simulator shall enable the user to be able to start a simulation from various plant conditions by choosing from a list of predefined ICs. An illustrative list is presented in Table 11. Additionally, the Simulator shall enable the user to be able to specify a custom set of initial conditions.

Table 11: Initial Conditions Initial Condition Number Plant State 1 100% power, end of core life. 2 100% power, middle of core life. 3 100% power beginning of core life. 4 50% power, end of core life. 5 50% power, middle of core life. 6 50% power beginning of core life. 7 Hot zero power at the point of adding heat 8 Hot reactor critical condition

4.5.3. Simulation Time

As standard, the Simulation shall run in real time. It shall be possible for the user to choose to slow down or speed up simulation time. It is suggested that time may be slowed by a factor of ten or accelerated by a factor of 2, 4, 8, 16 or 32. This will allow demonstration of rapid or long term transients or for fast entering of desired plant conditions from an initial condition.

The Simulation shall be able to be frozen and resumed at any time by the user. It shall also be possible for the user to freeze the simulation and then manually advance it one second at a time. At the end of each time step (simulation second) the simulation shall freeze until the user initiates the next time step manually. This shall be useful for tutorial purposes or for debugging behaviour during transients. Page 42 of 90

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4.5.4. Snapshot

At any point during simulation, the Simulator shall enable the user to be able to copy the plant conditions to a new custom set of initial conditions in a ‘snapshot’. This will allow new simulations to be started from that state in future.

4.5.5. Backtrack and Replay

During the simulation, the Simulator shall automatically record plant conditions at a standard (or user defined) interval (for example – every 30 seconds). The user will then be able to ‘backtrack’ the simulation to one of these stored sets of plant conditions shall they desire. This will remove the need to restart a simulation completely if the user makes a mistake or has forgotten to ‘snapshot’ a condition. Storage space requirements for a high number of backtrack initial conditions shall be specified.

It is also desirable for the user to be able to ‘replay’ the stored plant conditions from a specified point so that they may look again at the plant characteristics, perhaps from a different display page. This ‘replay’ mode shall be able to be exited at any point so that the normal simulation mode may be continued (optional requirement).

4.5.6. Manual Parameter Variation

For plant systems, particularly where complicated automatic operation logic is involved, the operation shall be defaulted to automatic mode. At any time, the Simulator shall enable the user to be able to choose to assume manual control and override plant condition parameters (for example – pressure values, valve states, pump % performance, plant power demand setpoint, trip setpoints) if desired, to investigate the effects that doing so has on plant operation. The Simulator shall also enable the user to be able to manually adjust plant trip points and set points. Upon reselecting automatic operation for a parameter, the simulator shall resume automatic control.

The Simulator shall enable the user to be able to edit the basic plant data (before or during simulator operation) to change the specification of the plant being simulated. Examples of parameters which could be changed include rated thermal power, valve setpoints, ultimate heat sink temperature etc.

4.5.7. Malfunctions

The Simulator shall enable the user to be able to initiate one or more malfunctions either from a predefined list of generic malfunctions or by manually overriding the performance of one or more specific components. This will allow the user to see how the plant responds and to learn how best to operate the plant in fault conditions. A list of suggested generic malfunctions is presented in Chapter 3.4.

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4.5.8. Plant Configuration Variations (Optional Requirement)

As discussed in Chapter 2.1, it is desirable for the simulator to be able to include plant design configuration modifications. The desired configurations are presented in Table 12. If this feature is included, the GUI shall make it simple for the user to change the configuration before starting a simulation. Plant configuration shall be locked during simulation runs.

Table 12: Plant Configuration Options Plant Design Feature Possible Configurations

Circulation Natural / Forced with four horizontally mounted coolant pumps

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5. OTHER REQUIREMENTS

5.1. DOCUMENTATION

As a minimum, the Contractor shall provide the IAEA with the following documentation:

 A design specification – A detailed design specification shall be provided to the IAEA for review prior to the start of the construction of the simulator models.

 A complete user manual for the Simulator (see Chapter 5.1.1);

 Project progress documentation to be supplied during the project (see Chapter 5.2).

5.1.1. Minimum Content of User Manual

Prior to customer acceptance of the Simulator, a user manual shall be supplied to the IAEA in printed copy and one set of electronic media (Microsoft Word or .pdf format). The manual shall include the following content, as a minimum. Units shall be provided for all parameters when specified within the manual.

 Introduction

 Purpose

 Historical background

 Simulation principles

 Description of the iPWR design and its various systems. The manual shall clearly state any parts of the plant which have not been simulated. Inclusion of diagrams in this section is desirable.

 Simulator installation, start-up and initialisation instructions.

 Discussion of the main features of the Simulator and its operation

 List of Simulator display screens and description of the main features of each.

 Simulator Exercises for standard operations – steps to be taken and expected results.

o Introductory exercises

o Reactor Start-up and Heat up

o Power manoeuvre: 10% power reduction and return to full power (for both turbine lead and reactor lead)

o Reduction to 0% full power and back to 100% full power

o Turbine trip and recovery

o Reactor trip and recovery

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 Simulator exercises for malfunction transient events – steps to be taken and expected results.

o A selection of events from those listed in Table 9 in Chapter 3.4.  Detailed description of mathematical models used. This shall be broken down by model, in a similar manner to that presented in Chapter 7. This would ideally include discussion of the modelling assumptions, solution technique, convergence criteria and associated numerical errors.

 Description of control logic used in the Simulator.

 References

 Glossary explaining any acronyms used.

 Appendix – software description including program description, data fields, arrays etc.

 Appendix – hardware requirements for Simulator.

5.2. PROJECT MANAGEMENT

As part of the tender documentation, the Contractor shall provide a proposed project schedule including progress review dates. Also provided shall be the project cost break down and details of the quality assurance arrangements. The project schedule shall be updated in agreement with the IAEA following IAEA acceptance of the tender.

During the project, the Contractor shall provide the IAEA with monthly project management updates in the form of progress reports. These shall include as a minimum:

 Report on the status of modelling for each system.  Report on the status of simulator software programming.  Detailed project schedule updates.  Example display pages from the Simulator GUI as they become available.  Discrepancy reports (if applicable).

This reporting shall allow the IAEA to identify any problems at an early stage.

5.3. SCOPE OF SUPPLY

The simulation software shall be supplied to the IAEA with complete ownership rights. The IAEA shall be able to use and distribute the software freely as it wishes.

5.4. QUALITY ASSURANCE

The Contractor shall provide the IAEA with details of its quality assurance plan in the tender documentation. As a minimum, it is expected that this shall include details of basic document control, configuration management, software coding practices, and verification. Contractors with proven experience of developing nuclear power plant simulators are likely to be preferred.

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The following process is suggested for initial verification and validation of the Simulator:

1. Verification conducted by Contractor to confirm that the program code is a correct implementation of the agreed physics models and control logic. 2. Acceptance testing by IAEA staff and subject matter experts through an IAEA consultation meeting. This will involve extensive trials of the Simulator for both normal and abnormal reactor operations. The focus for this testing will be ease of use of the Simulator (e.g. GUI) and realism of the plant transients. Following this, changes will be requested that the Contractor shall implement prior to acceptance of the Simulator by the IAEA. 3. Validation against data. This step will be completed under a separate contract in the future if operational plant or licensing data becomes available.

During verification acceptance testing and validation it is important that all discrepancies are recorded and tracked by both the Contractor and the IAEA.

5.5. WARRANTY AND SOFTWARE UPGRADES

The developer shall provide minor software upgrades free of charge to the IAEA for a one-year period to ensure that any computer bugs are resolved.

5.6. AFTER SALES SUPPORT

The tender documentation shall include details of what options the Contractor can offer the IAEA in after sales support. It is desirable that the IAEA shall be able to receive assistance relating to modelling, simulator operations and troubleshooting. Options for instructor training in use of the Simulator would also be desirable.

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6. OPTIONS FOR TENDER

Various simulator features have been discussed in this document. Some of these are necessary, others are desirable but not essential. To aid the procurement process, IAEA has specified three options for the supply of the basic principles simulator, as presented in Table 13. It is requested that the Contractor provide a quote for the price and lead time associated with each of these three options in its tender submission.

Table 13: Simulator Supply Options Option Details A (Basic) Natural circulation iPWR. Simulator runs in Windows XP, 7 and 8. Simulator shall include all features that are specified within this document with the exception of those explicitly stated as part of Options B, C, D, E or F.

B Variable plant configuration as specified in Chapter 4.5.8.

C Ability to split the display across several monitors as specified in Chapter 4.2.

D Inclusion of system/component descriptions as specified in Chapter 4.2.

E Platform independent operation as specified in Chapter 4.3.

F Inclusion of replay feature as specified in Chapter 4.5.5.

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7. MATHEMATICAL MODELS

7.1. OVERVIEW

This Chapter presents a set of mathematical models as an illustration of how the various physical systems within the iPWR plant can be modelled. This is a compilation of models from various sources, some of which are existing simulator manuals. As the presentation of the models in these sources is typically very good, some of these models have been directly copied from the sources. The models are included here for information only. The Contractor may use these models, if appropriate, or may use its own if it has superior models available. In the tender submission, the Contractor shall clearly specify what models the simulator will be based upon. A detailed system model was developed by Galvin for an iPWR and the associated thesis may also be a useful reference (Galvin 2009). It is thought likely that the models presented by Galvin are of too high a fidelity for use in a basic principles simulator, however.

7.2. REACTOR CORE

7.2.1. Reactor spatial kinetic model (IAEA 2003)

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7.2.2. Approximation method for coupling coefficients

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7.2.3. Reactor core kinetics and decay heat model summary

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7.3. FUEL HEAT TRANSFER (IAEA 2009)

The lumped parameter technique can be used to calculate the heat transfer from fuel rods:

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Figure 20: Cross section of a fuel pellet enclosed by metal fuel clad. The reactor coolant takes heat from the clad.

For fuel elements in a reactor zone, the transient fuel mean temperature and fuel clad temperature are given by:

Where:

Q = nuclear heating of fuel rod C= thermal capacity for fuel pellet = πr C ρ

C= thermal capacity for fuel clad = 2πr ΔrC ρ R= resistance of fuel and clad gap = 4πk1 2πr1h

k = fuel thermal conductivity; hg = fuel and clad gap conductance T = average fuel pellet temperature in the zone T = average fuel clad temperature in the zone T = average coolant temperature in the zone channel R2 = outside radius of the fuel pellet including the clad.

7.4. NATURAL CIRCULATION FLOW

Natural circulation will occur in the iPWR pressure vessel when buoyant forces caused by differences in fluid densities (due to differences in temperature) are sufficient to overcome the flow resistance of loop components (e.g. steam generators). Accurate modelling of mass, momentum and energy transfer in a nuclear reactor’s coolant is a formidable task and requires numerical solution of complex equations. This is likely to be impractical for the purposes of this basic principles simulator. If possible, simplified models shall be used. A fundamental principles approach to modelling single- phase and two-phase natural circulation within an iPWR is presented in (IAEA 2005) and (IAEA 2012), as summarised below. Note that if the option to include main coolant pumps in the plant Page 61 of 90

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configuration is to be made available, a thermal hydraulic model will also be required for forced circulation flow.

7.4.1. Single-phase natural circulation

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7.4.2. Two-phase natural circulation

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7.5. PUMPED FLOW (IAEA 2003)

The torque balance (angular momentum) equation for the shaft and rotating assembly is

π Ω Where: I= pump moment of inertia Ω = pump Speed (RPM)

T= motor torque T= hydraulic torque T= friction torque

The head and torque characteristics of a pump as a function of flow rate and rotor speed, are determined using the homologous theory. In this theory, the pump parameters are represented by their normalized values. The shapes of the homologous curves depend only on the rated speed of the pump. The homologous modeling relates normalized head h, normalized hydraulic torque β, to normalized flow v, and speed α, by tabulating:

, < 1

, < 1

These curves are fitted with a high order polynomial function of ( ), and ( ) respectively, and are used by the model to compute pump head and torque. The pump head so determined can be used as an input to the primary hydraulic model. The pump torque can also be used, as input to the torque balance equation.

7.6. PRESSURISER

7.6.1. Basic pressuriser model (IAEA 2003)

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7.6.2. Linearised Pressuriser Model for Integral Reactor (Kuridan and Beynon 1998)

A linearised model for the pressuriser in an iPWR was developed by (Kuridan and Beynon 1998). The key features of this model are replicated here. The ‘surge line diode’ represents the element that limits the flow through the baffle plate.

Figure 21: Key features of integral pressurizer (Kuridan and Beynon 1998)

The following assumptions are made:  During insurge (flow from main vessel into pressuriser) the lower control volume assumed to be subcooled and the upper to be superheated.  During outsurge a two phase mixture is assumed in each region.  Spray plus condensate mixture will enter the water phase as saturated liquid.  The process of steam condensation on vessel wall is neglected compared to other mass transfer terms.  Insurge water mixes completely with water already present in the pressuriser.  Pressure inside the pressuriser is uniform.  Heat transfer between the pressuriser contents and the pressuriser wall is negligible.

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Figure 22: Schematic diagram of the pressurizer during an insurge (Kuridan and Beynon 1998)

For insurge flow:

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System of equations for insurge flow:

Spray condensation:

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System of equations for insurge flow can be linearised using:

This gives the linearised insurge equations:

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For outsurge flow:

Figure 23: Schematic diagram of the pressurizer during an insurge (Kuridan and Beynon 1998)

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System of equations for outsurge flow:

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Condensate fall rate:

Bubble rise rate:

The system of equations for outsurge flow can be linearised using:

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Using this, the linearised equations for outsurge flow are:

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7.7. STEAM GENERATOR

The steam generator in this iPWR design is of the helical-tube, once through type. Other simulators have typically used models for U-tube SGs.

A simple SG model is presented in (Micro-Simulation Technology 2011).

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7.8. FEEDWATER (IAEA 2003)

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7.9. MAIN STEAM SYSTEM (IAEA 2003)

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7.10. CONTROL AND PROTECTION SYSTEMS (IAEA 2003)

7.10.1. Pressuriser Pressure Control System

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7.10.2. Pressuriser Level Control System

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7.10.3. Steam Generator Pressure Control System

The steam generator pressure is automatically controlled to be constant.

7.10.4. Steam Dump Control System

7.10.5. Rod Control System

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7.11. BREAK DISCHARGE (Micro-Simulation Technology 2011)

If a break is within the subcooled region and the water level is above the break, the choking orifice flow will be:

If the break is located in the saturated region’s liquid phase, Moody’s liquid critical flow will be used. If it is located in the vapour phase, Moody’s vapour critical flow will be used.

7.12. PIPING

For an idealized adiabatic flow in a pipe system, the transient temperature at the outlet of a pipe section is equivalent to the transient temperature at its inlet delayed by the transport time. This delay will vary continuously in the case of pump trip or coast down. This is known as time delay model. This is not realistic as it does not account for turbulent mixing in pipe, as well as heat storage in the pipe walls. Both of these factors can have a major effect on the transient temperatures and shall be considered to obtain a realistic solution.

The governing equations for the model are given below:

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Heat transfer between light water coolant and metal wall

Heat transfer between metal wall and insulation

Heat transfer between insulation and air

where M is the mass per unit length, C the specific heat, the conductance between coolant and wall, the conductance between wall and insulation, the conductance between insulation to atmosphere, subscripts C, W, in, a referring to coolant, wall, insulation and atmosphere respectively.

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8. ACRONYMS AND DEFINITIONS

Term Definition ADS Automatic Depressurisation System GUI Graphical User Interface CVCS Chemical Volume Control System IAEA International Atomic Energy Agency IC Initial Condition iPWR Integral Pressurised Water Reactor LOCA Loss of Coolant Accident MASLWR Multi-Application Small Light Water Reactor MCP Main Coolant Pump MW Megawatts NPTDS Nuclear Power Technology Development Section PWR Pressurised water reactor RPV Reactor Pressure Vessel SG Steam Generator SMR Small Modular Reactor User Refers to the user of the simulator software

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9. BIBLIOGRAPHY

Cassiopeia Technologies Inc. “Advanced Pressurised Water Reactor Simulator - User Manual.” Canada, 2011. Cassiopeia Technologies Inc. Generic Passive BWR Simulator. 2009. Cassiopeia Technologies Inc. IAEA Generic Pressurised Water Reactor Simulator. 2007. Galvin, Mark R. “System Model of a Natural Circulation Integral Test Facility.” PhD Dissertation, Oregon State University, 2009. Grove, E.J., and R.J. Travis. Effect of aging on the PWR chemical and volume control system, NUREG/CR-5954, BNL-NUREG-52410. Upton: Brookhaven National Laboratory, 1995. IAEA. Advances in Small Modular Reactor Technology Developments, A Supplement to: IAEA Advanced Reactors Information System (ARIS). IAEA, 2014. IAEA. Boiling Water Reactor Simulator with Passive Safety Systems - User Manual. Vienna: IAEA, 2009. IAEA. IAEA-TECDOC-1474 - Natural circulation in water cooled nuclear power plants - Phenomena, models, and methodology for system reliability assessments. Vienna: IAEA, 2005. IAEA. IAEA-TECDOC-1677 - Natural circulation phenomena and modelling for advanced water cooled reactors. Vienna: IAEA, 2012. —. Simulators for training nuclear power plant personnel, IAEA-TECDOC-685. 1993. IAEA. Training Course Series No. 22 - Pressurised Water Reactor Simulator Workshop Material. Vienna: IAEA, 2003. James E. Fisher, S. Michael Modro, Kevan D. Weaver, Jose Reyes, John Groome, Pierre Babka. “Performance and Safety Studies for Multi-Application Small Light Water Reactor (MASLWR).” n.d. Korea Atomic Energy Research Institute. Visual System Analyzer (ViSA) Two-loop Large PWR Simulator. n.d. Kuridan, R.M., and T.D. Beynon. “A linearized non steady state model for the pressuriser of the safe integral reactor concept.” Progress in Nuclear Energy, Vol. 3, No. 4., 1998: 421-438. Micro-Simulation Technology. PCTRAN - Personal Computer Transient Analyser For a Two-Loop PWR And TRIGA Reactor. User Manual, Micro-Simulation Technology, 2011. Micro-Simulation Technology. PCTran/PWR Personal Computer Transient Analyser For a Two-loop PWR Version 6.0.1. Montville, New Jersey, 2011. Modro, S. Michael, et al. “Multi-Application Small Light Water Reactor Final Report.” Idaho National Engineering and Environmental Laboratory Idaho Falls, Idaho 83415 , 2003. Moscow Engineering and Physics Institute. WWER-1000 Reactor Simulator. n.d. National Training Laboratories. Bethel, Maine, n.d. NuScale Power. NuScale Plant Design Overview. Corvallis, Oregon : NuScale Power, LLC, 2014. Ruben, Mazzi. “CAREM: An Innovative-Integrated PWR.” 18th International Conference on Structural Mechanics in Reactor Technology (SMiRT 18). Beijing, China, 2005. 4407-4415. Westinghouse. “Westinghouse Technology Systems Manual, Section 4.1, Chemical and Volume Control System.” n.d.

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