Design of Nuclear Reactor (Module 4)

Design of Nuclear Reactor (Module 4)

BASIC PROFESSIONAL TRAINING COURSE Module IV Design of a nuclear reactor Version 1.0, May 2015 This material was prepared by the IAEA and co-funded by the European Union. 2 TYPES OF NUCLEAR REACTORS Learning objectives After completing this chapter, the trainee will be able to: 1. 2. 3. 4. 5. List basic components of nuclear reactors. List basic types of nuclear power plants. Sketch and describe Pressurized Water Reactor (PWR). Sketch and describe Boiling Water Reactor (BWR). Describe basic features of PHWR, GCR and LWGR reactors. 6. Describe Fast Breeder Reactor.

7. Describe basic features of small and medium reactors. Basic Professional Training Course; Module IV Design of a nuclear reactor 3 Basic components of a nuclear reactor A nuclear power plant can be basically divided to: a nuclear part and a conventional part; In the nuclear part the fission energy is converted into heat which is used to produce steam. Its main component is nuclear reactor. In the conventional part this steam runs the turbine connected to generator. Reactor contains nuclear fuel, which is made of uranium (sometimes mixed with plutonium). There are two kinds of uranium atoms, called isotopes: 99.3% of atoms are uranium-238 or 238U, and

0.7% is uranium-235 or 235U. Basic Professional Training Course; Module IV Design of a nuclear reactor 4 Basic components of a nuclear reactor Only 235U can sustain nuclear fission chain reaction. The chain reaction is maintained by subatomic particles called neutrons. Neutrons born during fission are very energetic and are called fast neutrons. For inducing further fissions, however, the most efficient are slow or thermal neutrons which have negligible kinetic energy. A reactor must therefore have means to slow down neutrons. Moderator Basic Professional Training Course; Module IV Design of a nuclear reactor 5

Basic components of a nuclear reactor If ordinary hydrogen 1H is replaced with its heavy isotope 2H (denoted also D), its compound with oxygen is called heavy water D2O. The third possible moderator is graphite which is form of carbon. Nuclear reactions produce large quantities of heat which must be transferred out of the fuel. Reactor coolant. The coolant has to be in liquid or gaseous form and should not absorb neutrons substantially. Basic Professional Training Course; Module IV Design of a nuclear reactor 6 Basic components of a nuclear reactor Control system is used to start-up the reactor, to shut it down, and to adjust the reactor power level. Contains materials that are strong neutron absorbers, such as: boron,

indium, cadmium). The basic distinction is defined by the type of their fuel, moderator and coolant. Nuclear reactors are used also for some other purposes. propulsion of ships, research reactors; Basic Professional Training Course; Module IV Design of a nuclear reactor 7 Power Reactor Information System The IAEA developed a comprehensive database of nuclear power plants worldwide.

Power Reactor Information System or PRIS (http://www.iaea.org/pris) The database covers: Reactor specification data and technical design characteristics. Performance data including energy production and energy loss data, outage and operational event information. Monthly production and power loss data have been recorded in PRIS since 1970 and are complemented by information on nuclearpower generated energy provided to non-electrical applications such as district heating, process heat supply or desalination. Basic Professional Training Course; Module IV Design of a nuclear reactor 8 Power Reactor Information System A set of internationally accepted performance indicators has been developed for calculations with PRIS data. The indicators can be used for:

benchmarking, international comparison or for analyses of nuclear power availability and reliability according to reactor type, country or worldwide. Two official Agency publications are produced each year using PRIS data: Nuclear Power Reactors in the World Operating Experience with Nuclear Power Stations in Member States Basic Professional Training Course; Module IV Design of a nuclear reactor 9 Pressurized Water Reactor - PWR Pressurized Water Reactor PWR

Moderated and cooled with ordinary water. The pressure in the reactor is so high that the water does not boil. The heat is transferred to secondary side in the steam generator. The steam produced there drives the turbine. Basic Professional Training Course; Module IV Design of a nuclear reactor 10 Pressurized Water Reactor - PWR A significant majority of nuclear power plants is cooled by ordinary water. Water is in liquid state at least in the surroundings of fuel vapour is less efficient in moderating neutrons and also as in heat transfer. The coolant temperature in Light Water Reactors (LWR) is always below 375 C. The first and still the most common type of light water reactor is Pressurized Water Reactor. Pressure is typically around 15.5 MPa (155 bar). Basic Professional Training Course; Module IV

Design of a nuclear reactor 11 Pressurized Water Reactor - PWR The heat from primary water is transferred to secondary water in steam generators. The temperature of the steam which is around 280 C, determines also the thermal efficiency around 34%. Reactor vessel or pressure vessel: is made of steel and is around 22 cm thick, diameter is around 4 m, and height around 12 m; The fuel is: slightly enriched uranium (3 5%),

which is in several thousand fuel pins around 1 cm thick and around 4 m long; Basic Professional Training Course; Module IV Design of a nuclear reactor 12 Pressurized Water Reactor - PWR Primary system (reactor vessel, steam generators) is placed inside the containment. The Russian version of pressurized water reactor is called VVER. Physical principles of PWR and VVER reactors are the same several important technical differences; Main advantage of PWRs radioactive coolant is effectively separated from the environment; PWR technology proved to be reliable and cost effective;

Basic Professional Training Course; Module IV Design of a nuclear reactor 13 Boiling Water Reactor - BWR Boiling Water Reactor - BWR Moderated and cooled with ordinary water. Water boils in the reactor and the resulting steam drives the turbine. Basic Professional Training Course; Module IV Design of a nuclear reactor 14 Boiling Water Reactor - BWR The second type of light water reactors is Boiling Water Reactor (BWR). In a BWR water boils already in reactor and the steam produced, with temperature around 290 C, is directly led to turbine. Pressure in the reactor vessel half of the pressure in a PWR,

consequently walls are thinner; Steam separation inside the reactor vessel; The fuel similar to PWR fuel; Boiling water reactors have containment; Basic Professional Training Course; Module IV Design of a nuclear reactor 15 Boiling Water Reactor - BWR The advantage of boiling water reactor is relatively simple design. There are no the steam generators. Disadvantage contaminated with radioactive substances turbine, condenser and other steam system parts; The cost of some other components is higher total investment and the operating costs are very much comparable with those of PWR;

Boiling water reactors are the second most common type of reactors. Basic Professional Training Course; Module IV Design of a nuclear reactor 16 Pressurized Heavy Water Reactor - PHWR Pressurized Heavy Water Reactor PHWR Moderated and cooled with heavy water. Water does not boil in the reactor. Heavy water transfers its heat to light water in the steam generators, the resulting steam drives the turbine. Basic Professional Training Course; Module IV Design of a nuclear reactor 17 Pressurized Heavy Water Reactor - PHWR If heavy water is used as moderator, fuel can be natural uranium. Canadians have developed a pressurized Heavy Water Reactor CANDU;

Fuel made of natural uranium inside a large number of pressure tubes, coolant (heavy water) under pressure flows through; The fuel is grouped into so-called 'bundles': about half meter long elements, 10 cm in diameter, made of individual fuel pins. Basic Professional Training Course; Module IV Design of a nuclear reactor 18 Pressurized Heavy Water Reactor - PHWR

The main advantage of CANDU reactors is the possibility to use natural uranium. Most economical operation is achieved by using slightly enriched (around 1%) uranium. The disadvantage of heavy water reactors is: expensive production of heavy water and replacement of its losses, relatively complex regulation system, and lower thermal efficiency (up to 30%). Besides Canada, CANDU reactors are in India, Pakistan, Argentina, South Korea, Romania and China. Their share is 11% of all reactors in the world. Basic Professional Training Course; Module IV Design of a nuclear reactor 19

Gas Cooled Reactor GCR, AGR, HTGR Gas Cooled Reactor GCR, Advanced Gas-cooled Reactor AGR Moderator is graphite, coolant is gas which in the steam generator transfers its heat to water. The resulting steam drives the turbine. Basic Professional Training Course; Module IV Design of a nuclear reactor 20 Gas Cooled Reactor GCR, AGR, HTGR Natural uranium can be used as fuel also in graphite-moderated reactors. United Kingdom developed a type of reactors called GCR which were cooled with CO2 at temperature around 400 C. An improved version called AGR uses slightly enriched uranium in stainless steel cladding which allows CO2 temperatures up to 650 C. The advantage of gas cooled reactors is: high thermal efficiency. Other costs, including the investment costs, are higher than for light water reactors.

Their share is 3% of all reactors in the world. Basic Professional Training Course; Module IV Design of a nuclear reactor Light Water Graphite moderated Reactor LWGR 21 Light Water Graphite moderated Reactor LWGR The moderator is graphite and the coolant is water that boils in pressure tubes around the fuel. Basic Professional Training Course; Module IV Design of a nuclear reactor Light Water Graphite moderated Reactor LWGR Graphite moderated reactors can be cooled also with water. In 1986, one of this type of reactors in Chernobyl (nowadays Ukraine) suffered the worst nuclear accident ever. RBMK reactor: Water boils in pressure tubes that encompass fuel rods.

Fuel is uranium enriched to around 2 %. Pressure tubes are distributed in a large graphite structure. Steam that is produced in the fuel area, is collected in large vessels. The reactor core is quite large and has a complex control system. There is no containment. Refuelling can be made during operation of reactor. Basic Professional Training Course; Module IV Design of a nuclear reactor 22 Light Water Graphite moderated Reactor LWGR

An important feature of RBMK reactors is that they are unstable at low power. This was, besides lack of safety culture, the main cause for the accident that happened in Chernobyl on April 26, 1986. After the accident, there were several modifications in the remaining RBMK reactors. About 3 % of all nuclear power plants today are RBMK reactors. Basic Professional Training Course; Module IV Design of a nuclear reactor 23 24 Fast Breeder Reactor - FBR Fast Breeder Reactor FBR There is no moderator. The primary and the secondary coolant is liquid metal, usually sodium. The secondary coolant transfers its heat to water in steam generators. The resulting steam drives the turbine. Basic Professional Training Course; Module IV Design of a nuclear reactor 25

Fast Breeder Reactor - FBR In all reactors described so far the chain reaction was maintained by slow, so called thermal neutrons. But also fast neutrons can sustain chain reaction. An important feature of fast neutron-induced fission is that a higher number of new neutrons is born. To sustain the chain reaction, on average one neutron born in fission is required. The majority of the neutrons can be absorbed in non-fissile isotope of uranium, 238U. This absorption reaction leads to production of artificial element plutonium. Basic Professional Training Course; Module IV Design of a nuclear reactor 26 Fast Breeder Reactor - FBR Fast reactors are therefore also often called breeder reactors. Paradox can be explained with the fact that breeder reactors produce fuel from normally non-fissile 238

U. In principle, fast breeder reactors could therefore cover world electricity demand for several thousand years. Fast breeder reactor: Fuel is 10-30% enriched uranium or is mixed with 10-30% of plutonium. There is no moderator and reactor is cooled with liquid sodium. The boiling point of sodium is quite high. Sodium becomes very radioactive heat is transferred to a secondary sodium loop; Basic Professional Training Course; Module IV Design of a nuclear reactor 27

Small and Medium Reactors SMR Reactor classification small reactors [an equivalent electric power of less than 300 MW(e)], medium sized [between 300 and 700 MW(e)]. Worldwide, 131 Small and Medium Reactors (SMR) are in operation in 26 Member States, with a capacity of 59 GWe. The considerable development work on small to medium sized designs generally aims to provide increased benefits in the areas of: safety and security, non-proliferation, waste management, and

resource utilization and economy, as well as to offer a variety of energy products and flexibility in design, siting and fuel cycle options. Basic Professional Training Course; Module IV Design of a nuclear reactor 28 Questions 1. Which subatomic particle sustains the nuclear fission chain reaction? 2. List 4 basic components of a nuclear reactor! 3. Which of uranium isotopes is fissile and what is its abundance in natural uranium? 4. What is the name of artificial element that is (besides uranium) fissile? 5. List two type of reactors that are moderated with ordinary (light) water! 6. Which type of reactors is most common in the world? Basic Professional Training Course; Module IV

Design of a nuclear reactor 29 Questions 7. State the moderator for each of the reactors listed: a) PWR b) CANDU c) Chernobyl d) Fukushima e) Fast breeder reactor 8. In which types of NPPs the reactor coolant runs the turbine? 9. List 2 types of nuclear power plants that are moderated with graphite! Basic Professional Training Course; Module IV Design of a nuclear reactor 30 DESIGN OF RESEARCH REACTORS Learning objectives After completing this chapter, the trainee will be able to: 1. Briefly describe the research reactors history and statistics. 2. List main types of research reactors. 3. Distinguish the main types of research reactor fuel. 4. Recognize the importance of research reactors for

nuclear safety in power reactors. Basic Professional Training Course; Module IV Design of a nuclear reactor 31 DESIGN OF RESEARCH REACTORS Research reactors have played an important role in the development of nuclear science and technology. Research reactors have many and varied missions leading to many and varied designs and operating modes; Research reactors are smaller in power rating (than typical power reactors) the inventory of radioactive materials in their cores is also much smaller smaller hazard potential; Safe siting, design and operation are essential maintain the excellent safety record; Basic Professional Training Course; Module IV

Design of a nuclear reactor 32 Research reactor statistics The IAEA maintains the Research Reactor Database ( http://nucleus.iaea.org/RRDB/RR/ReactorSearch.aspx?rf=1). The following statistical information can be derived from this database: 747 reactors were built in 69 countries (246 reactors in 55 countries are still classified as operational). Of the 55 countries having operational research reactors, 29 countries having 45 operational research reactors do not have an operating power reactor. The RRDB lists a total of 501 reactors as either shut down or decommissioned. There are 18 reactors listed as either under construction or planned.

However, the viability of several of these projects is uncertain. Basic Professional Training Course; Module IV Design of a nuclear reactor 33 Research reactor statistics Research reactors were and remain very widespread around the world. Number of operational research reactors located in countries that do not have an operating power reactor; Issues include lack of financial and human resources, aging of the facility, lack of utilization and inadequate regulatory oversight continued safe operation of the reactors a significant challenge to the owners, their governments and the international community; Basic Professional Training Course; Module IV Design of a nuclear reactor 34

Research reactor utilization Research reactors and the neutrons they produce have a very wide variety of uses in nuclear science and technology. These include: applications in education and training, biology, agriculture, medicine, materials science, geochronology,

industry and safety research. Many research reactors are located at universities and serve as important tools in education and training. Basic Professional Training Course; Module IV Design of a nuclear reactor 35 Research reactor utilization Neutron activation analysis detecting the presence of various trace elements; Production of radioactive isotopes medical diagnostics and treatment; Such radioactive isotopes are: Mo (99mTc), 131I, 60Co 99 Boron-neutron capture therapy (BNCT) a technique for treatment of certain cancers;

Research reactors produce neutron beams for use in scattering experiments for determination of material structures and properties. Some industrial applications neutron radiography as a complement to X-ray and other nondestructive evaluation techniques; Basic Professional Training Course; Module IV Design of a nuclear reactor 36 Research reactor utilization Experiments done in research reactors have made significant contributions to safety of current and future power reactors. Research reactors have made major contributions to the nuclear industry and to the well-being of humanity. Need for research reactor services and products remains strong there are many challenges to be met in an increasingly economically competitive and safety-conscious environment; Approach to meeting these challenges

consolidation of the functions regional research reactor facilities services are provided at a minimum cost and with maximum safety; networks and coalitions Basic Professional Training Course; Module IV Design of a nuclear reactor 37 Types of research reactors There are many design variations in research reactors, influenced by the primary purpose of the reactor: materials testing; neutron source; multi-purpose; pulsed; critical experiments; or training. These variations include: The cooling system design, The moderator,

The reflector, The fuel, The power level; Basic Professional Training Course; Module IV Design of a nuclear reactor 38 Types of research reactors Research reactors of low and medium power the open pool design; Reactors are cooled and moderated by light water cooled by natural circulation;

Reflectors of beryllium or tanks of heavy water enhance core neutron flux; The open pool design is suitable for in-core and in-reflector irradiations. Basic Professional Training Course; Module IV Design of a nuclear reactor 39 Types of research reactors Open pool reactors suitable for installation of in-core loops for safety testing of fuel elements, for power reactors under prototypic conditions pressure and temperature. TRIGA is a widely used example of an open pool reactor. Steady-state power range from 100 kW to 14 MW,

A pulsing capability, U-ZrH fuel, A complete tutorial on the TRIGA reactor may be found in the IAEAs safety training material; Basic Professional Training Course; Module IV Design of a nuclear reactor 40 Types of research reactors Another variation on the open pool design is the so-called tank-in pool: The reactor core is enclosed in a closed tank through which the coolant is pumped.

For high power reactors coolant loop is slightly pressurized, or using heavy water as a moderator and/or coolant necessary to separate the heavy and light water (in the tank); Heavy water moderator very high thermal neutron flux for beam port experiments or high-flux irradiations; Basic Professional Training Course; Module IV Design of a nuclear reactor 41 Types of research reactors A closed tank design is used in cases where a higher power than can be accommodated with a tank in pool design is needed. These reactors generally operate at elevated pressure and temperature, and so have some similarities to power reactors. Numerous low power reactors used for: training, university research, activation analysis and applications requiring only a low neutron flux have been built;

These reactors are typically rated at a few tens of kilowatts or less. Basic Professional Training Course; Module IV Design of a nuclear reactor 42 Research reactor fuels The fuels used in research are, like the designs, very diverse. The most common form is plates, pins/rods or concentric tubes of UAl alloy or U3Si2Al dispersion, clad with aluminium. U-Al fuels are typically enriched to about 93% 235U. The silicide fuels are enriched to 19.75% 235U. Many research reactor designed in the Soviet Union now use 36% enriched fuel. TRIGA reactors use a U-ZrH or U-ZrH1.65 alloy fuel in Al or 304

stainless steel cladding. The original TRIGA fuel was 20% enriched, but some reactors converted to the FLIP (Fuel Life Improvement Program) fuel, which is 70% enriched. Basic Professional Training Course; Module IV Design of a nuclear reactor 43 Research reactor fuels Effort to reduce the civilian use of highly enriched uranium the RERTR program and GTRI. Has as its aim to conversion of as many research reactors as possible to low-enriched uranium fuel. A reduction in the enrichment by a factor of about 5 means an increase in the content of 238U in the fuel, resulting in increased neutron absorption and decreased density of fissile atoms. Qualification studies, have shown that the irradiation behaviour of the silicide fuel is satisfactory.

LEU fuel development is continuing to develop and qualify a fuel having even higher density (e.g. UMo) so that the very high power reactors that cannot use the silicide fuel successfully can be converted. Basic Professional Training Course; Module IV Design of a nuclear reactor 44 Research reactors and power reactor safety Experiments conducted in research reactors have been of great importance in developing the safety technology for power reactors and confirming our understanding of the behaviour of materials under irradiation and in accidents. Steady-state irradiation of sample fuels, cladding and structural material have been carried out in many materials testing reactors. Thus, research and development of new fuels and materials can proceed at a faster rate than would otherwise be possible. Basic Professional Training Course; Module IV Design of a nuclear reactor

45 Research reactors and power reactor safety Experiments in research reactors have contributed significantly to safety technology for both: water-cooled and sodium-cooled reactors. These experiments generally involve fuel and material samples. Measurements of interest include: inter alia, fuel failure energy in transients,

fuel relocation following failure, fission product release from failed fuel and fission product transport in the reactor cooling system. Basic Professional Training Course; Module IV Design of a nuclear reactor 46 Research reactors and power reactor safety Some of the reactors used and the types of experiments done include: The PHEBUS-FP experiments simulated a severe accident in a PWR involving meltdown of a portion of the core. The CABRI reactor was used for a series of experiments simulating accidents in fast reactors.

The TREAT has been used for many simulations of fast reactor accidents. JSRR - Experiments on failure of LWR-fuels have been conducted in this reactor. IGR reactor has been used for many transient experiments on LWR fuels and recently on a program of experiments to investigate fuel relocation in fast reactor accidents. BR-2 reactor hosted a series of experiments simulating fuel failure and meltdown in fast reactor accidents. Basic Professional Training Course; Module IV Design of a nuclear reactor 47 Questions 1. List the areas in which the research reactors are used! 2. List some of the most important medical isotopes that are produced in research reactors!

3. List some of the most important types of the research reactors! 4. Briefly describe the open pool TRIGA reactor! 5. What are most common fuels that are used in research reactors? 6. Explain how the use of research reactors contribute to the development of the safety of power reactors! Basic Professional Training Course; Module IV Design of a nuclear reactor SAFETY CONCEPTS IN THE DESIGN OF NUCLEAR REACTORS Learning objectives After completing this chapter, the trainee will be able to: 1. Describe the basic safety objective in the design of a nuclear installation. 2. Describe the term Design Basis Accident (DBA). 3. Describe the term Postulated Initiating Event (PIE). 4. List the levels of defence in the design of nuclear installation. 5. Describe the concept of a series of physical barriers. Basic Professional Training Course; Module IV Design of a nuclear reactor 48 49 Basic safety objectives

Fundamental safety objective has to be achieved without unduly limiting the operation of facilities or the conduct of activities that give rise to radiation risks. To ensure objective, measures have to be taken: To control the radiation exposure of people and the release of radioactive material to the environment; To restrict the likelihood of events that might lead to a loss of control over a nuclear reactor core, nuclear chain reaction, radioactive source or any other source of radiation; To mitigate the consequences of such events if they were to occur. The fundamental safety objective is to protect people and the environment from harmful effects of ionizing radiation. Basic Professional Training Course; Module IV Design of a nuclear reactor 50 Basic safety objectives In order to achieve the safety principles in designing a nuclear power plant, a comprehensive safety analysis is carried out.

In this context, the following definitions are important: Design basis accident (DBA) is a postulated accident condition against which a facility is designed according to established design criteria, and for which the damage to the fuel and the release of radioactive material are kept within authorized limits. Postulated initiating event (PIE) is an event identified during design as capable of leading to anticipated operational occurrences or accident conditions. Basic Professional Training Course; Module IV Design of a nuclear reactor 51 Basic safety objectives The safety analysis examines all plant states: all planned normal operational modes of the plant; plant performance in anticipated operational occurrences; design basis accidents;

event sequences that may lead to a severe accident; and severe accidents. On the basis of this analysis: the robustness of the engineering design in withstanding postulated initiating events can be established, the effectiveness of the safety systems and safety related items or systems can be demonstrated, and requirements for emergency response can be established. Basic Professional Training Course; Module IV Design of a nuclear reactor 52 Basic safety objectives

Measures are taken to control radiation exposure in all operational states and to minimize the likelihood of an accident. Measures are therefore taken to ensure that the radiological consequences are mitigated. Such measures include: engineered safety features and systems (ESF); on-site accident management procedures established by the operating organization; and possibly off-site intervention measures established by appropriate authorities in order to mitigate radiation exposure if an accident has occurred. Basic Professional Training Course; Module IV Design of a nuclear reactor 53 The concept of defence in depth Application of the concept of defence in depth in the design of a plant provides a series of levels of defence:

inherent features, equipment and procedures aimed at preventing accidents and ensuring appropriate protection in the event that prevention fails. Basic Professional Training Course; Module IV Design of a nuclear reactor 54 First level of defence Its aim is to prevent deviations from: normal operation and the failure of items important to safety. This leads to the requirement that the plant be:

soundly and conservatively sited, designed, constructed, maintained, and operated. To meet these objectives, it is important to select appropriate design codes and materials, and paid attention to the quality control of the manufacture of components and construction of the plant, as well as to its commissioning. Basic Professional Training Course; Module IV Design of a nuclear reactor 55 Second level of defence

Its aim is to: detect and control deviations from normal operational states in order to prevent anticipated operational occurrences at the plant from escalating to accident conditions. This is in recognition of the fact that some PIEs are likely to occur over the service lifetime of a nuclear power plant, despite the care taken to prevent them. Basic Professional Training Course; Module IV Design of a nuclear reactor 56 Third level of defence For this level, it is assumed that, although very unlikely, escalation of certain anticipated operational occurrences or PIEs might not be controlled at a preceding level and that an accident could develop. These unlikely events are anticipated in the design of the plant. This leads to the requirement that:

inherent and/or engineered safety features, safety systems and procedures are capable of preventing damage to the reactor core or significant off-site releases and returning the plant to a safe state. Basic Professional Training Course; Module IV Design of a nuclear reactor 57 Fourth level of defence Its aim is to mitigate the consequences of accidents that result from failure of the third level of defence in depth. The most important objective of this level is the protection of the confinement function and thus to ensure that radioactive releases are kept as low as reasonably achievable. Basic Professional Training Course; Module IV Design of a nuclear reactor 58

Fifth level of defence This is the final level of defence aimed at mitigation of the radiological consequences of potential releases of radioactive materials that may result from accident conditions. This requires: the provision of an adequately equipped emergency control centre, and plans for the on-site and off-site emergency response. A relevant aspect of the implementation of defence in depth is: the provision in the design of a series of physical barriers, as well as a combination of active, passive and inherent safety features. Basic Professional Training Course; Module IV Design of a nuclear reactor 59

Questions 1. What is the fundamental safety objective for nuclear installation? 2. Which tools are used to ensure stat safety objectives are met? 3. Describe the meaning of abbreviations DBA and PIE! 4. How many levels of defence in depth there are in the design of a nuclear installation? 5. Give an example of series of physical barriers in a nuclear power plant! 6. Give an example of series of physical barriers in a radwaste repository! Basic Professional Training Course; Module IV Design of a nuclear reactor 60 BASIC SAFETY FEATURES OF THE DESIGN Learning objectives After completing this chapter, the trainee will be able to: 1. List main organizational requirements for the design organization. 2. List main design management requirements. 3. List main design requirements for defence in depth. 4. Define main fundamental safety functions which must be performed. 5. List and briefly describe main requirements for plant design.

6. List and briefly describe main requirements for design of plant systems. Basic Professional Training Course; Module IV Design of a nuclear reactor 61 Management of safety The design organization ensures that the installation is designed to meet the requirements of the operating organization. Thus, the design organization shall: implement safety policies; have a clear division of responsibilities with corresponding lines of authority and communication; ensure that it has sufficient technically qualified and appropriately trained staff at all levels; establish clear interfaces between the groups;

develop and strictly adhere to sound procedures; review, monitor and audit all safety related design matters on a regular basis; and ensure that a safety culture is maintained. Basic Professional Training Course; Module IV Design of a nuclear reactor 62 Management of safety The design management for a nuclear power plant must ensure that: all components important to safety have the appropriate characteristics; the requirements of the operating organization are met;

due account is given to the capabilities and limitations of the personnel who will eventually operate the plant; adequate safety design information is supplied; recommended practices for incorporation into the plant administrative and operational procedures are supplied; results of the deterministic and complementary probabilistic safety analyses are taken into account; generation of radioactive waste is kept to the minimum practicable; Basic Professional Training Course; Module IV Design of a nuclear reactor 63

Management of safety Wherever is possible all components important to safety must be: designed according to the latest or currently applicable approved standards, of a design proven in previous equivalent applications, selected to be consistent with the plant reliability goals necessary for safety. Where an unproven design or feature is introduced or there is a departure from an established engineering practice, safety must be demonstrated to be adequate: by appropriate supporting research programmes, or by examining operational experience from other relevant applications. Basic Professional Training Course; Module IV

Design of a nuclear reactor 64 Management of safety The design must take due account of relevant operational experience. A comprehensive safety assessment is carried out. Safety assessment part of the design process; The operating organization ensures that an independent verification of the safety assessment is performed. Basic Professional Training Course; Module IV Design of a nuclear reactor 65 Management of safety Integrated management system arrangements for the management, performance, and

assessment of the plant design; Design, including subsequent changes or safety improvements carried out in accordance with established procedures, appropriate engineering codes and standards, and incorporate applicable requirements and design bases; The adequacy of design, including design tools and design inputs and outputs verified or validated by individuals or groups; Basic Professional Training Course; Module IV Design of a nuclear reactor 66 Principal technical requirements Primary means of preventing and mitigating the consequences of

accidents defence in depth, which is incorporated in the design. The design therefore: multiple physical barriers combination of a number of consecutive and independent levels of protection; conservative, construction must be of high quality; provides for control of the plant behaviour during and following a PIE; provides for supplementing control of the plant, by the use of automatic activation of safety systems; equipment and procedures to control the course and limit the consequences of accidents; one level fails subsequent level will be available The independent effectiveness of the different levels of defence is a

necessary element of defence in depth. Basic Professional Training Course; Module IV Design of a nuclear reactor 67 Principal technical requirements Defence in depth maintained, the design must prevent: Challenges to the integrity of physical barriers; Failure of a barrier when challenged; Failure of a barrier as a consequence of failure of another barrier. Defence in depth is provided by: management system strong commitment to safety and safety culture, site selection, incorporation engineering features safety margins, diversity and redundancy

Comprehensive operational procedures and practices, accident management procedures; The design must be such that the first, or at most the second, level of defence is capable of preventing escalation to accident conditions for all but the most improbable PIEs. Basic Professional Training Course; Module IV Design of a nuclear reactor 68 Principal technical requirements To ensure safety, the following fundamental safety functions must be performed in operational states, in and following a design basis accident and, to the extent practicable, in and after the occurrence of plant conditions considered that are beyond those of the design basis accidents: Control of the reactivity; Removal of heat from the core;

Confinement of radioactive materials and control of operational discharges, as well as limitation of accidental releases. A systematic approach is followed to identify the systems, structures and components that are necessary to fulfil the safety functions at the various times following a PIE. Basic Professional Training Course; Module IV Design of a nuclear reactor 69 Principal technical requirements The plant design is such that its sensitivity to PIEs is minimized. The expected responses to any PIE are from those of the following that can reasonably be achieved (in order of importance): A PIE produces no significant safety related effect or produces only a change in the plant towards a safe condition by inherent characteristics; or Following a PIE, the plant is rendered safe by passive safety features or by the action of safety systems that are continuously operating in the state necessary to control the PIE; or

Following a PIE, the plant is rendered safe by the action of safety systems that need to be brought into service in response to the PIE; or Following a PIE, the plant is rendered safe by specified procedural actions. Basic Professional Training Course; Module IV Design of a nuclear reactor 70 Principal technical requirements In order to achieve the three safety objectives given in the design of a nuclear installation, all actual and potential sources of radiation are identified and properly considered, and provision is made. The design must have as an objective the prevention or, if this fails, the mitigation of radiation exposures resulting from design basis accidents and selected severe accidents. Plant conditions that could potentially result in high radiation doses or radioactive releases are restricted to a very low likelihood of occurrence, and it is ensured that the potential

radiological consequences of conditions with a significant likelihood of occurrence are only minor. Basic Professional Training Course; Module IV Design of a nuclear reactor Requirements for plant design Safety classification All items important to safety must be first identified and then classified. Classification: The safety function(s) to be performed by the item; The consequences of failure to perform their function; The frequency with which the item will be called upon to perform a safety function; and The time following a PIE at which, or the period for which, the item will be called upon to perform a safety function (operate).

The design ensures that any failure in a system classified in a lower class will not propagate to a system classified in a higher class. Basic Professional Training Course; Module IV Design of a nuclear reactor 71 72 General design basis For all items important to safety is in the design basis specified: the necessary capability, reliability and functionality. Over the lifetime of the nuclear power plant. If the design basis for each item important to safety is systematically justified and documented, then this documentation

could provide necessary information for safe plant operation. Basic Professional Training Course; Module IV Design of a nuclear reactor 73 Categories of plant conditions The plant conditions are identified and grouped into a limited number of categories. The categories typically cover: Normal operation; Anticipated operational occurrences, which are expected to occur over the operating lifetime of the plant; Design basis accidents; and Design extension conditions, including accidents with significant degradation of the reactor core (in old terminology: Severe accidents).

Basic Professional Training Course; Module IV Design of a nuclear reactor 74 Postulated initiating events In designing the plant, it is recognized that challenges to all levels of defence in depth may occur and design measures are provided to ensure that the necessary safety functions are accomplished and the safety objectives can be met. These challenges stem from the PIEs, which are selected on the basis of deterministic or probabilistic techniques or a combination of the two. Independent events, each having a low probability, are normally not anticipated in the design to occur simultaneously. Basic Professional Training Course; Module IV Design of a nuclear reactor 75 Internal events An analysis of the PIEs is made to establish all those internal events that may affect the safety of the plant. Fires and explosions Requirements are achieved by suitable incorporation of:

redundant parts, diverse systems, physical separation, and design for fail-safe operation. With such incorporation the following objectives are achieved: To prevent fires from starting; To detect and extinguish quickly; To prevent the spread of fires. Basic Professional Training Course; Module IV Design of a nuclear reactor

76 Internal events Other internal hazards The potential for internal hazards such as: flooding, missile generation, pipe whip, jet impact, or release of fluid from failed systems or from other installations; Appropriate preventive and mitigatory measures; Some external events may initiate internal fires or floods;

Interaction of external and internal events; Basic Professional Training Course; Module IV Design of a nuclear reactor 77 External events The design basis natural and human induced external events Significant radiological risk Combination of deterministic and probabilistic methods Natural external events: earthquakes, floods, high winds, tornadoes, tsunami (tidal waves), and

extreme meteorological conditions Human induced external identified in site characterization Basic Professional Training Course; Module IV Design of a nuclear reactor 78 Site related characteristic In determining the design basis of a nuclear power plant, various interactions between the plant and the environment are taken into account. Including such factors as: population, meteorology, hydrology,

geology and seismology. Basic Professional Training Course; Module IV Design of a nuclear reactor 79 Combinations of events Randomly occurring individual events Certain events may be the consequences of other events, such as: a flood following an earthquake. Such consequential effects are considered to be part of the original PIE. Design limits A set of design limits consistent with the key physical parameters for each structure, system or component are specified for operational states and design basis accidents.

Basic Professional Training Course; Module IV Design of a nuclear reactor 80 Operational states The plant is designed to operate safely within a defined range of parameters, and a minimum set of specified support features for safety systems are assumed to be available. The design is such that the response of the plant to a wide range of anticipated operational occurrences will allow safe operation or shutdown The potential for accidents to occur in low power and shutdown states are addressed in the design Basic Professional Training Course; Module IV Design of a nuclear reactor 81 Design basis accidents

A set of design basis accidents is derived from the listing of PIEs Provision is made to initiate the necessary safety system actions automatically Manual initiation of systems or other operator actions administrative, operational and emergency procedures; Basic Professional Training Course; Module IV Design of a nuclear reactor 82 Severe accidents Plant conditions may jeopardize the integrity of barriers Beyond design basis accidents Severe accidents Combination of engineering judgement and probabilistic methods Realistic or best estimate assumptions, methods and analytical criteria Basic Professional Training Course; Module IV Design of a nuclear reactor 83

Severe accidents Design activities for addressing severe accidents take into account the following: Important event sequences; Event sequences are reviewed; Potential design changes or procedural changes; Plants full design capabilities; Multiunit plants; Accident management procedures; Basic Professional Training Course; Module IV Design of a nuclear reactor

Design for reliability of systems and components All components important to safety are designed to be capable of withstanding all identified PIEs with sufficient reliability. Basic Professional Training Course; Module IV Design of a nuclear reactor 84 85 Common cause failures The potential for common cause failures of items important to safety is considered to determine where the principles of: diversity, redundancy and independence should be applied to achieve the necessary reliability.

Basic Professional Training Course; Module IV Design of a nuclear reactor 86 Single failure criterion The single failure criterion is applied to each safety group Spurious action Safety function must be performed under the following conditions: Any potentially harmful consequences of the PIE for the safety group are assumed to occur; and The worst permissible configuration of safety systems; Non-compliance with the single failure criterion is exceptional Not be necessary to assume the failure Basic Professional Training Course; Module IV Design of a nuclear reactor 87 Fail-safe design The principle of fail-safe design is considered and incorporated into the design of systems and components important to safety for the

plant as appropriate. If a system or component fails, plant systems are designed to pass into a safe state with no necessity for any action to be initiated. Basic Professional Training Course; Module IV Design of a nuclear reactor 88 Auxiliary services Auxiliary services that support equipment that forms part of a system important to safety are considered part of that system and are classified accordingly. Auxiliary services necessary to maintain the plant in a safe state may include the supply of: electricity, cooling water and compressed air or other gases, and means of lubrication.

Basic Professional Training Course; Module IV Design of a nuclear reactor In-service testing, maintenance, repair and inspection All components important to safety are designed to be: calibrated, tested, maintained, repaired or replaced, inspected and monitored; Plant layout Basic Professional Training Course; Module IV

Design of a nuclear reactor 89 90 Ageing Appropriate margins are provided in the design for all components important to safety so as to take into account relevant ageing and wear-out mechanisms and potential age related degradation, in order to ensure the capability of the: structure, system or component to perform the necessary safety function throughout its design life. Basic Professional Training Course; Module IV Design of a nuclear reactor 91

Human factors The design must be operator friendly and is aimed at limiting the effects of human error. Systematic consideration of human factors and the human-machine interface is included in the design process The human-machine interface The design is aimed at promoting the success of operator actions Need for intervention Following an event, the physical environment in the control room or in the supplementary control room, and the access route to that supplementary control room is acceptable Basic Professional Training Course; Module IV Design of a nuclear reactor Other design considerations Sharing of safety systems between multiple units of a nuclear power plant Safety systems must not be shared between two or more nuclear power plants unless, if this mean enhance of safety. In exceptional cases it is permitted that, safety system support features and safety related items are shared between two or more units

If such sharing dont contribute to safety, then must not be permitted. Basic Professional Training Course; Module IV Design of a nuclear reactor 92 Systems containing fissile or radioactive materials All systems within a nuclear power plant that may contain fissile or radioactive materials must be designed: To prevent the occurrence of events; To prevent accidental criticality and overheating; To ensure that radioactive releases of material are kept below authorized limits on discharges; and To facilitate mitigation of radiological consequences of accidents.

Basic Professional Training Course; Module IV Design of a nuclear reactor 93 94 Escape routes from the plant Sufficient number of safe escape routes, clearly and durably marked; Escape routes must meet: the national and international requirements for radiation zoning and fire protection the national requirements for industrial safety and plant security Available at least one escape route from workspace and other occupied areas; Basic Professional Training Course; Module IV Design of a nuclear reactor 95

Communication systems at the plant Effective means of communication; Communication is available for use after events; Suitable alarm systems and means of communication are provided for: warning instructions; Suitable and diverse means of communication within the nuclear power plant, in the immediate vicinity and with off-site agencies; Basic Professional Training Course; Module IV Design of a nuclear reactor

96 Control of access The plant is isolated from the surroundings layout of the structural elements access permanently controlled The design of the buildings and the layout of the site provision is made for operating personnel and/or equipment, attention is paid to guarding against the unauthorized entry; Must be prevented unauthorized access, interference for any reason; Basic Professional Training Course; Module IV Design of a nuclear reactor

Prevention of harmful interactions of systems important to safety Simultaneous operation systems important to safety possible interaction is evaluated, effects of interactions prevented; Analysis physical interconnections possible effects of one systems operation, malfunction failure Basic Professional Training Course; Module IV Design of a nuclear reactor 97 Interactions between the electrical power grid

and the plant The functionality of items important to safety is not compromised by: disturbances in the electrical power grid, anticipated variations in the voltage frequency Decommissioning Special consideration is given to the incorporation of features that will facilitate the decommissioning and dismantling; Basic Professional Training Course; Module IV Design of a nuclear reactor 98 99 Safety analysis A safety analysis of the plant design

deterministic probabilistic analysis; The design basis for items important to safety established confirmed meeting the prescribed and acceptable limits defence in depth achieved; Basic Professional Training Course; Module IV Design of a nuclear reactor 100

Deterministic approach Analysis includes the following: Operational limits and conditions are in compliance; Characterization of the PIEs; Analysis and evaluation of event sequences; Comparison of the results; Establishment and confirmation of the design basis; Demonstration; Verification

analytical assumptions, methods and degree of conservatism; Updating significant changes operational experience, Basic Professional Training Course; Module IV Design of a nuclear reactor 101 Probabilistic approach Is carried out in order to do the following: To give confidence; To demonstrate; To provide confidence;

To provide assessments of the probabilities of occurrence of severe core damage states; To provide assessments of the probabilities of occurrence and the consequences of external hazards; To identify systems; To assess the adequacy; To verify compliance; Basic Professional Training Course; Module IV Design of a nuclear reactor 102 Requirements for design of plant systems Safety recommendations for the design plant systems are given in

several Safety guides: NS-G-1.3, Instrumentation and Control Systems Important to Safety in Nuclear Power Plants NS-G-1.4, Design of Fuel Handling and Storage Systems in Nuclear Power Plants NS-G-1.5, External Events Excluding Earthquakes in the Design of Nuclear Power Plants NS-G-1.6, Seismic Design and Qualification for Nuclear Power Plants NS-G-1.7, Protection Against Internal Fires and Explosions in the Design of Nuclear Power Plants Basic Professional Training Course; Module IV Design of a nuclear reactor 103

Requirements for design of plant systems NS-G-1.8, Design of Emergency Power Systems for Nuclear Power Plants NS-G-1.9, Design of the Reactor Coolant System and Associated Systems in Nuclear Power Plants NS-G-1.10, Design of Reactor Containment Systems for Nuclear Power Plants NS-G-1.11, Protection against Internal Hazards other than Fires and Explosions in the Design of Nuclear Power Plants NS-G-1.12, Design of the Reactor Core for Nuclear Power Plants NS-G-1.13, Radiation Protection Aspects of Design for Nuclear Power

Plants Basic Professional Training Course; Module IV Design of a nuclear reactor 104 Reactor core and associated features General design The reactor core and associated coolant, control and protection systems are designed with appropriate margins in all operational states and in design basis accidents; The maximum degree of positive reactivity and its maximum rate of increase limited; Recriticality or reactivity excursion minimized; The reactor core and associated coolant, control and protection systems inspection and testing; Basic Professional Training Course; Module IV Design of a nuclear reactor 105

Reactor core and associated features Fuel elements and assemblies Are designed to withstand satisfactorily the anticipated irradiation and environmental conditions notwithstanding all processes of deterioration; Permit adequate inspection of structure and component parts; In design basis accidents, the fuel elements remain in position and dont suffer distortion core cooling insufficiently effective, specified limits arent exceeded; Basic Professional Training Course; Module IV Design of a nuclear reactor 106

Reactor core and associated features Control of the reactor core The provisions for fuel for all levels and distributions of neutron flux in all states of the core, after shutdown and during or after refuelling; Adequate means of detecting flux distributions are provided no regions of the core the provisions breached without being detected; Provision is made for removal of non-radioactive substances, corrosion products; Basic Professional Training Course; Module IV Design of a nuclear reactor 107

Reactor core and associated features Reactor shutdown Means are provided to ensure that there is a capability to shut down the reactor in operational states and design basis accidents, the shutdown condition can be maintained; Specified limits are not exceeded effectiveness, speed of action and shutdown margin; Reactivity control and flux shaping in normal power operation a part of the means of shutdown may be used; Basic Professional Training Course; Module IV Design of a nuclear reactor 108

Reactor core and associated features The means for shutting down the reactor at least two different systems; At least one of the two systems capable of quickly rendering the nuclear reactor subcritical single failure, capable of rendering the reactor subcritical and maintaining the reactor subcritical (even for the most reactive conditions of the core); Judging the adequacy of the means of shutdown part of the means inoperative, a common cause failure; Basic Professional Training Course; Module IV Design of a nuclear reactor

109 Reactor core and associated features The means of shutdown are adequate to: prevent, withstand inadvertent increases in reactivity; Deliberate actions that increase reactivity in the shutdown state and a single failure taken into account; Instrumentation is provided and tests are specified shutdown means are always in the state stipulated for the given plant condition; Basic Professional Training Course; Module IV Design of a nuclear reactor 110

Reactor coolant system designed with sufficient margin to ensure that reactor coolant pressure boundary are not exceeded in operational states; adequate isolation devices to limit any loss of radioactive fluid The component parts: reactor pressure vessel or the pressure tubes, piping and connections, valves, fittings, pumps,

circulators and heat exchangers, together with the devices by which such parts are held in place; Basic Professional Training Course; Module IV Design of a nuclear reactor 111 Reactor coolant system Materials for the component parts selected minimize activation of the material; The reactor pressure vessel and the pressure tubes designed and constructed highest quality, respect to, materials, design standards,

capability of inspection, fabrication; Basic Professional Training Course; Module IV Design of a nuclear reactor 112 Reactor coolant system The design of the components, such as pump impellers or valve parts minimize the likelihood of failure and associated consequential damage, in all operational states and in design-basis accidents, with deterioration that may occur; In-service inspection of the reactor coolant pressure boundary Components are designed, manufactured and arranged

possible to carry out inspections and tests; Material surveillance programme (for determining) metallurgical effects of factors irradiation, stress corrosion cracking, thermal embrittlement and ageing; Basic Professional Training Course; Module IV Design of a nuclear reactor 113 Reactor coolant system Indicators for the integrity monitored; Results of measurements determination which inspections are necessary for safety; Safety analysis indicates failures in the secondary cooling system serious consequences inspection of the secondary cooling system;

Inventory of reactor coolant Control of the inventory and pressure design limits are not exceeded in any operational state; Basic Professional Training Course; Module IV Design of a nuclear reactor 114 Reactor coolant system Clean-up of the reactor coolant Adequate facilities removal of radioactive substances including activated corrosion products and fission products leaking; Removal of residual heat from the core Means for removing residual heat Safety function transfer fission product decay heat and other residual heat

rate design limits are not exceeded; Basic Professional Training Course; Module IV Design of a nuclear reactor 115 Reactor coolant system Interconnections and isolation capabilities and other appropriate design features provided on the assumptions single failure and the loss of off-site power incorporation of redundancy, diversity and independence; Emergency core cooling Provided in the event of a loss of coolant accident Limiting parameters for the cladding or fuel integrity not exceed; Chemical reactions limited;

Alterations in the fuel and internal structural alterations not significantly reduce the effectiveness of cooling; Cooling ensured for a sufficient time. Basic Professional Training Course; Module IV Design of a nuclear reactor 116 Reactor coolant system Design features and suitable redundancy and diversity provided to fulfil requirements for each PIE (assumption of a single failure); Extending the capability to remove heat from the core following a severe accident; Inspection and testing of the emergency core cooling system Designed to permit periodic inspection and testing to confirm: Structural integrity and leak tight integrity;

The operability and performance of the active components; and The operability under the conditions specified in the design basis. Basic Professional Training Course; Module IV Design of a nuclear reactor 117 Reactor coolant system Heat transfer to an ultimate heat sink Systems provided transfer residual heat an ultimate heat sink; Function carried out very high levels of reliability operational states and DBAs; Reliability achieved by use of proven components,

redundancy, diversity, physical separation, interconnection, and isolation; Basic Professional Training Course; Module IV Design of a nuclear reactor 118 Containment system Design of the containment system Containment system provided

release of radioactive materials to the environment below specified limit design-basis accident; This system includes: leaktight structures, associated systems for the control of pressures and temperatures, features for isolation, management and removal of fission products, hydrogen, oxygen and other substances; Basic Professional Training Course; Module IV Design of a nuclear reactor 119 Containment system Strength of the containment structure Strength of the containment structure

calculated with sufficient margins of safety (on the basis of) internal overpressures, underpressures and temperatures, dynamic effects, and reaction forces; Provision for maintaining the integrity the effects of any predicted combustion of flammable gases; Capability for containment pressure tests Designed and constructed pressure test to demonstrate structural integrity; Basic Professional Training Course; Module IV Design of a nuclear reactor 120 Containment system Containment leakage Design maximum leakage rate not exceeded Containment structure and equipment and components

designed and constructed leak rate can be tested (design pressure); Determination of the leakage rate at the containment design pressure, or at reduced pressures permit estimation of the leakage rate; Control any leakage of radioactive materials event of a severe accident; Basic Professional Training Course; Module IV Design of a nuclear reactor 121 Containment system Containment penetrations Number of penetrations kept to a practical minimum;

Penetrations meet same design requirements as the containment; Protected against reaction forces pipe movement, or accidental loads missiles, jet forces and pipe whip; Capability penetrations remain functional severe accident; Basic Professional Training Course; Module IV Design of a nuclear reactor 122 Containment system Containment isolation Line that penetrates containment (part of the reactor coolant

pressure boundary) automatically, and reliably sealable; in the event of a design-basis accident Lines are fitted with two containment isolation valves, arranged in series; reliable and independent actuation Line that penetrates containment ( not part of the reactor coolant pressure boundary)

at least one containment isolation valve, valve is outside the containment; Basic Professional Training Course; Module IV Design of a nuclear reactor 123 Containment system Containment air locks Personnel access to the containment airlocks equipped with doors doors interlocked; Internal structures of the containment Ample flow routes between separate compartments; Cross-sections of openings (between compartments)

pressure differentials occurring during pressure equalization not result in damage to the pressure bearing structure; Capability of the internal structures withstand the effects of a severe accident; Basic Professional Training Course; Module IV Design of a nuclear reactor 124 Containment system Removal of heat from the containment Capability to remove heat from the containment Safety function is fulfilled by reducing pressure and temperature; System performing the function has adequate reliability and

redundancy; Capability to remove heat from the containment in the event of a severe accident; Basic Professional Training Course; Module IV Design of a nuclear reactor 125 Containment system Control and clean-up of the containment atmosphere Systems to control fission products, hydrogen, oxygen and other substances provided; Systems for cleaning up the containment atmosphere suitable redundancy in components and features fulfil the safety function; Control of fission products, hydrogen and other substances

also considered in the event of a severe accident; Basic Professional Training Course; Module IV Design of a nuclear reactor 126 Instrumentation and control General requirements for instrumentation and control systems important to safety Instrumentation to monitor variables and systems normal operation, anticipated operational occurrences, design-basis accidents, and severe accidents;

Measuring all main variables that can affect: the fission process, the integrity of the reactor core, the reactor cooling systems, and the containment; Basic Professional Training Course; Module IV Design of a nuclear reactor 127 Instrumentation and control Instrumentation and recording equipment provided to ensure essential information, and

for predicting the locations and quantities of radioactive materials; The instrumentation and recording equipment adequate to provide information, for determining the status of the plant, and for taking decisions; Appropriate and reliable controls provided to maintain the variables; Basic Professional Training Course; Module IV Design of a nuclear reactor 128 Instrumentation and control Control room A control room

safe operation in all its operational states, and measures can be taken to maintain the plant in a safe state; Identifying events which may pose a direct threat to its continued operation internal and external events; Layout of the instrumentation and mode of presentation adequate overall picture of the status and performance of the plant; Devices visual, and audible indications; Basic Professional Training Course; Module IV Design of a nuclear reactor

129 Instrumentation and control Supplementary control room Instrumentation and control equipment at a single location, physically and electrically separate; Reasons for supplementary control room reactor placed and maintained in a shutdown state, residual heat removing, and monitoring of essential plant variables; Use of computer based systems in systems important to safety

appropriate standards and practices for development and testing computer hardware, software; Basic Professional Training Course; Module IV Design of a nuclear reactor 130 Instrumentation and control Development subject integrated management system; Level of reliability commensurate with safety importance; Reliability is achieved means of a comprehensive strategy various complementary means at each phase of development,

a validation strategy confirm the design requirements; Safety analysis includes conservatism; Basic Professional Training Course; Module IV Design of a nuclear reactor 131 Instrumentation and control Automatic control Various safety actions automated operator action is not necessary within a justified period of time; Appropriate information for operator to monitor the effects of the automatic actions; Functions of the protection system

Automatically initiate the operation of appropriate systems; Detect design-basis accidents; Overriding unsafe actions; Basic Professional Training Course; Module IV Design of a nuclear reactor 132 Instrumentation and control Reliability and testability of the protection system High functional reliability and periodic testability; Redundancy and independence designed into the protection system: Single failure protection function remains; The removal from service no loss of redundancy; Design techniques to prevent loss of a protection function: testability,

self-checking capability, fail-safe, diversity, Basic Professional Training Course; Module IV Design of a nuclear reactor 133 Instrumentation and control The protection system designed to permit: periodic testing, testing channels independently; The design permits tests during operation

functionality from the sensor to the input signal to the final actuator; The design minimizes the influence of operator action possible defeating the effectiveness of the protection system, but not to negate correct operator actions; Basic Professional Training Course; Module IV Design of a nuclear reactor 134 Instrumentation and control Use of computer based systems in protection Where is used in a protection system, requirements are taken into account: The highest quality of and best practices; Systematic documentation and reviewing;

Assessment by expert personnel; A diverse means of ensuring fulfilment of the protection functions provided; Basic Professional Training Course; Module IV Design of a nuclear reactor 135 Instrumentation and control Separation of protection and control systems Interference between the protection system and the control systems Prevented by avoiding interconnections, or suitable functional isolation; Signals used in common by both systems separation ensured,

demonstrated safety requirements fulfilled; Basic Professional Training Course; Module IV Design of a nuclear reactor 136 Emergency control centre An on-site emergency control centre provided separated, serve as meeting place for the emergency staff; Information available there parameters and radiological conditions in the plant, and immediate surroundings; The room provides means for

communication (with control room,); Measures taken protect the occupants for a protracted time; Basic Professional Training Course; Module IV Design of a nuclear reactor 137 Emergency power supply After PIEs, emergency power is needed various systems and components important to safety; Ensured emergency power supply in any operational state, in a design basis accident,

assumption of the coincidental loss of off-site power; Basic Professional Training Course; Module IV Design of a nuclear reactor 138 Waste treatment and control systems Systems to treat radioactive liquid and gaseous effluents radioactive discharges within prescribed limits; ALARA principle applied; Systems for handling and safely storing on the site for a period of time; Transport of solid wastes decision of competent authorities; Control of releases of radioactive liquids to the environment Means to control the release of radioactive liquids

to the environment emissions and concentrations remain within limits; Basic Professional Training Course; Module IV Design of a nuclear reactor 139 Waste treatment and control systems Control of airborne radioactive material Ventilation system provided to do the following: Prevent unacceptable dispersion; Reduce the concentration; Keep the level below prescribed limits; and Ventilate rooms;

Control of releases of gaseous radioactive material to the environment Ventilation system filtration system control the release to the environment (within prescribed limits); Basic Professional Training Course; Module IV Design of a nuclear reactor 140 Fuel handling and storage systems Handling and storage of non-irradiated fuel Handling and storage systems for non-irradiated fuel do the following: Prevent criticality by physical means or processes; Permit maintenance, periodic inspection and testing; and

Minimize the probability of loss of or damage; Handling and storage of irradiated fuel Handling and storage systems for irradiated fuel designed: prevent criticality, permit adequate heat removal, permit inspection; Basic Professional Training Course; Module IV Design of a nuclear reactor 141 Fuel handling and storage systems Handling and storage of irradiated fuel (cont.) Permit periodic inspection and testing,

Prevent the dropping of spent fuel, Prevent unacceptable handling stresses, Prevent the inadvertent dropping of heavy objects, Permit safe storage of suspect or damaged fuel, Proper means for radiation protection, Identify individual fuel modules, Control soluble absorber levels, Facilitate maintenance and decommissioning,

Facilitate decontamination, and Operating and accounting procedures, Basic Professional Training Course; Module IV Design of a nuclear reactor 142 Radiation protection General requirements Preventing any avoidable radiation exposure and to keeping any unavoidable exposures; Objective accomplished: Layout and shielding, Design of the plant and equipment minimize the number and duration of human activities and reduce contamination,

Treatment of radioactive materials, and Reduce the quantity and concentration of radioactive materials; Basic Professional Training Course; Module IV Design of a nuclear reactor 143 Radiation protection Design for radiation protection Provision made in the design and layout minimize exposure and contamination; Provision includes design of systems and components in terms of: minimizing exposure maintenance and inspection,

shielding, ventilation and filtration, limiting the activation, means of monitoring, control of access to the plant, and decontamination facilities; Shielding design radiation levels do not exceed limits; (ALARA) Basic Professional Training Course; Module IV Design of a nuclear reactor 144 Radiation protection

Layout and procedures control of access to radiation and contamination areas, minimize contamination; Means of radiation monitoring Equipment is provided radiation monitoring operational states, design-basis accidents, and severe accidents; Basic Professional Training Course; Module IV Design of a nuclear reactor 145

Questions 1. What are requirements for the design organization? 2. What is ensured with design management? 3. What must be done in case of an unproven design or feature? 4. What are the fundamental safety functions that must be performed to ensure safety in all operational states and in case of accident? 5. List factors that are taken into account when the classifying of the SSC is made! 6. List categories of plant condition! 7. Briefly describe meaning of the: Common cause failure, Single failure criterion, Fail-safe design, Auxiliary service! Basic Professional Training Course; Module IV Design of a nuclear reactor 146 Questions 8. Briefly describe two methods of safety analysis: deterministic and probabilistic approach (what does include)! 9. List requirements for reactor core and associated features! 10. List requirements for reactor coolant system! 11. List requirements for containment system! 12. List requirements for instrumentation and control! 13. What is function of the protection system and for what is designed? 14. List requirements for fuel handling and storage systems!

15. List requirements for radiation protection! Basic Professional Training Course; Module IV Design of a nuclear reactor SAFETY REQUIREMENTS AND GUIDANCE FOR RESEARCH REACTORS DESIGN Learning objectives After completing this chapter, the trainee will be able to: 1. 2. 3. 4. List main safety issues of research reactors. Recognize the important points of the contents of NS-R-4. List other IAEA publications for safety in research reactors. List serious research reactor incidents and accidents. Basic Professional Training Course; Module IV Design of a nuclear reactor 147 148 IAEA Safety Requirements NS-R-4 Requirements for research reactors NS-R-4 Safety of Research Reactors;

Comprehensive collection of the safety requirements: Regulatory supervision; Management and verification of safety; Site evaluation; Design; Operation; Decommissioning; Appendix and Annexes;

Guidance on applying requirements is provided in SSGs Supporting documents compliment the Safety Standards and include Safety Reports and TECDOC Basic Professional Training Course; Module IV Design of a nuclear reactor 149 Factors to be considered in a graded approach Research reactors wide variety of sizes and designs, used for many varied purposes; A graded approach application of requirements; Requirements applied to research reactors limited potential for hazard public,

environment; Research reactors may pose a greater hazard to the operators and facility personnel. Basic Professional Training Course; Module IV Design of a nuclear reactor 150 Factors to be considered in a graded approach The scope, extent and detail of the safety analysis may be significantly less accident scenarios may not apply or need only a limited analysis; Extensive analysis, including standards for power reactors and/or additional special safety measures: research reactors with power level in excess of several tens of megawatts,

fast reactors, and reactors with experimental devices; Flexible approach to achieving and managing safety; Basic Professional Training Course; Module IV Design of a nuclear reactor 151 Factors to be considered in a graded approach All safety requirements applied cannot be graded to zero; Factors considered: reactor power, radiological source term,

amount and enrichment, presence of various systems and materials, design of the reactor, amount and rate of reactivity addition, reactivity control mechanisms,, containment or confinement structure, utilization factors, siting factors; Basic Professional Training Course; Module IV Design of a nuclear reactor

152 Factors to be considered in a graded approach Many factors must be considered; Factors are established at the design stage some may change as utilization of the reactor, its operating mode changes or site parameters change; Managers aware of changes during the lifetime, make changes; Basic Professional Training Course; Module IV Design of a nuclear reactor 153 Design philosophy Top-level design philosophy does not differ from power reactors, and

satisfy similar safety objectives; Defence-in depth concept protection against uncontrolled release; Design proven technology and conservative margins, a integrated management system, surveillance and inspection; Physical barriers provided, number and strength may be less;

Basic Professional Training Course; Module IV Design of a nuclear reactor 154 Design philosophy The three basic safety functions must be satisfied: Control of reactivity, Heat removal after shutdown, Confining radioactive material; Safety functions incorporating combination of inherent and passive safety features, engineered safety systems, and

administrative procedures; Design of the safety systems single failure criterion, high reliability, and provisions for regular inspection, testing and maintenance; Basic Professional Training Course; Module IV Design of a nuclear reactor 155 Design philosophy General design requirements NS-R-4 includes design requirements summarized here,

very brief shall statements, consult the source document; Classification of structures, systems and components (SSCs): function and significance to safety, consequences of failure; Classification to grade the design and quality requirements; Codes and standards identified and applied safety classification; Basic Professional Training Course; Module IV Design of a nuclear reactor 156 Design philosophy Design

considers all challenges (during lifetime); Imposed demands determine the design basis of the research reactor facility; Challenges may arise normal operations, site-related characteristics, internal events, or external events; To ensure safety set of PIEs and DBAs formulated, and

inherent, passive or engineered safety features provided; Basic Professional Training Course; Module IV Design of a nuclear reactor 157 Design philosophy Design applies principles of: redundancy and the single failure criterion, diversity, independence and fail-safe design; Radiation protection important design consideration; Design provisions

shielding, ventilation, filtration, and decay systems; Monitoring instruments for radiation and airborne radioactive material; Basic Professional Training Course; Module IV Design of a nuclear reactor 158 Design philosophy Structural materials chosen limit doses,

inspection, maintenance and decommissioning; Neutron activation is considered in the design for radiation protection. Special consideration to experimental equipment since: cause hazards directly (if it fails or affecting safe operation), increase the hazard from a PIE; Modification designed to standards equivalent to the reactor, fully compatible with the reactor; Design of experimental devices; Basic Professional Training Course; Module IV Design of a nuclear reactor

159 Safety analysis and verification of safety A safety analysis part of the design process; Analysis addresses the response to a range of PIEs that lead to AOOs or postulated accidents, some may be the DBAs; Analyses are used as the basis for the design of SSCs, and the selection of operational limits and conditions (OLCs); Basic Professional Training Course; Module IV Design of a nuclear reactor 160 Safety analysis and verification of safety Scope of the safety analysis:

Characterization of the PIEs, Analysis of event sequences and evaluation of the consequences, Comparison of the results, Demonstration the AOOs and DBAs can be managed, Determination of OLCs, Analysis of safety systems and the engineered safety features, Analysis of the means of confinement, Consideration of safety of experimental devices and their impact;

Basic Professional Training Course; Module IV Design of a nuclear reactor 161 Safety analysis and verification of safety Usually use of deterministic methods probabilistic methods complement; Data used for safety assessment; Safety assessment design process Safety assessment an ongoing process One or more safety committees independent of the reactor manager, advise on relevant safety issues of design, commissioning, operation and utilization;

Basic Professional Training Course; Module IV Design of a nuclear reactor 162 Selected postulated initiating events Starting point for a safety analysis a set of postulated initiating events; Techniques for developing a set of PIEs failure modes and effects analysis, fault trees, experience and engineering judgment NS-R-4 provides lists of PIEs. They cover the following categories: Loss of electrical power supplies,

Insertion of excess reactivity, Loss of coolant flow, Loss of coolant, Erroneous handling or failure of equipment or components, Internal and external events, Human errors; Basic Professional Training Course; Module IV Design of a nuclear reactor Examples of operational aspects of research reactors that require particular attention NS-R-4 includes an annex

discusses operational aspects that require particular attention essential differences; Core configurations frequently changed manipulations with fuel assemblies, control rods and experimental devices; Care must be exercised to ensure relevant subcriticality, and reactivity limits not exceeded; Basic Professional Training Course; Module IV Design of a nuclear reactor 163 Examples of operational aspects of research reactors that require particular attention Changes in core loading

affect the nuclear and thermal characteristics; Characteristics correctly determined, and checked; Experimental devices potential impact on safety; devices assessed for their safety implications documentation; Modifications changing requirements assessed, documented, reported, and formal approved; Basic Professional Training Course; Module IV Design of a nuclear reactor 164

Examples of operational aspects of research reactors that require particular attention In pool-type research reactor manipulating in the vicinity of the reactor core components, experimental devices and material; Manipulations strictly in accordance with procedures and restrictions; Access to the controlled area and active involvement in utilization guest scientists, trainees, students and others who visit reactor; All procedures, restrictions and controls strictly observed (for staff and the visitors); Basic Professional Training Course; Module IV Design of a nuclear reactor 165

166 Other safety guidance for research reactors SSG-20, Safety Assessment for Research Reactors and Preparation of the Safety Analysis Report; SSG-24, Safety in the Utilization and Modification of Research Reactors; NS-G-4.1, Commissioning of Research Reactors; NS-G-4.2, Maintenance, Periodic Testing and Inspection of Research Reactors; NS-G-4.3, Core Management and Fuel Handling for Research Reactors; NS-G-4.4, Operational Limits and Conditions and Operating Procedures for Research Reactors; Basic Professional Training Course; Module IV Design of a nuclear reactor 167 Other safety guidance for research reactors NS-G-4.5, The Operating Organization and the Recruitment, Training and Qualification of Personnel for Research Reactors; NS-G-4.6, Radiation Protection and Radioactive Waste

Management in the Design and Operation of Research Reactors; SSG-10, Ageing Management for Research Reactors; SSG-22, Use of a Graded Approach in the Application of the Safety Requirements for Research Reactors; WS-G-2.1, Decommissioning of Nuclear Power Plants and Research Reactors. Basic Professional Training Course; Module IV Design of a nuclear reactor The Code of Conduct on the Safety of Research Reactors 168 Safety issues have been raised, these include: aging of research reactors, lack of adequate regulatory supervision, research reactors in a status that has come to be called extended

shutdown; Concern over these issues led to the development of the Code of Conduct serves as guidance to States; Scope of this Code safety at all stages of their lives; Provides a summary of the desirable attributes for safety management for decision makers of the State, the regulatory body, and the operating organization; Basic Professional Training Course; Module IV Design of a nuclear reactor The Code of Conduct on the Safety of Research Reactors Objective of this Code achieve and maintain a high level of safety proper operating conditions, prevention of accidents,

mitigation of the radiological consequences, etc.; Application of Code accomplished through national safety regulations; First topic is the Role of the State: Establishing and maintaining a legislative and regulatory framework, The need for Regulatory body, Financing system; Basic Professional Training Course; Module IV Design of a nuclear reactor 169 The Code of Conduct on the Safety of Research Reactors Next topic role of the Regulatory body. Similar topics, which could be also applicable for the NPP:

Assessment and verification of safety; Financial and human resources; Integrated management system; Human factors; Radiation protection; Emergency preparedness; Siting;

Design, construction and commissioning; Operation, maintenance, modification and utilization; Extended shutdown; Decommissioning; Basic Professional Training Course; Module IV Design of a nuclear reactor 170 The Code of Conduct on the Safety of Research Reactors Role of the operating organization. Four major areas: General recommendations Assessment and verification of safety, Financial and human resources, Integrated management system, Human factors,

Safety of research reactors Siting, Design, Construction and Commissioning, Operation, maintenance, Modification and Utilization; Extended shutdown Decommissioning Role of the IAEA. Basic Professional Training Course; Module IV Design of a nuclear reactor 171 Some serious research reactor incidents and accidents 172 Overall safety record of research reactors excellent

several serious accidents loss of life; Here is a brief description of these accidents. 21 August 1945 - Los Alamos (USA) A criticality accident occurred when an experimenter was piling reflector blocks around a sub-critical fuel assembly and the last block fell on the fuel assembly. One person died 28 days later. 21 May 1946 - Los Alamos (USA) Similar accident to the previous one same installation cause accidentally bringing a hollow beryllium shell too close to the fuel One person died 9 days later. Basic Professional Training Course; Module IV Design of a nuclear reactor Some serious research reactor incidents and accidents 12 December 1952 - NRX - Chalk River (Canada)

A power excursion occurred due to the regulating rods being inadvertently removed and failure of the safety rods to drop. Result 4000 m3 of reactor coolant water containing about 3.71014 Bq of activity, leaked into the basement; 29 November 1955 - EBR-1 (USA) A power excursion occurred during an experiment to measure the reactivity coefficient of the reactor. Meltdown of around 40% of the core; 15 October 1955 - Vinca (Yugoslavia) A mistake by an operator led to an inadvertent increase in the level of the heavy water and an uncontrolled criticality. One person died. Basic Professional Training Course; Module IV Design of a nuclear reactor 173 Some serious research reactor incidents and accidents

03 January 1961 - SL1 - Idaho Falls (USA) Human error power excursion and steam explosion. Three people have died. 30 December 1965 - Venus - Mol (Belgium) A power excursion due to human error Result subsequent amputation of the leg; 07 November 1967 - SiloeE - Grenoble (France) Partial meltdown of a fuel element Result release of about 21015 Bq of radioactivity into the water of the pool and 7.41013 Bq through the stack (mainly noble gases); Basic Professional Training Course; Module IV Design of a nuclear reactor 174 Some serious research reactor incidents and accidents

23 September 1983 - RA-2 - Constituyentes (Argentina) A power excursion flouting of the safety rules The operator died 48 hours after the accident. Note: The accidents at VENUS and SILOE could be classed as INES Level 3. The other accidents mentioned above could be classed as INES Level 4. Basic Professional Training Course; Module IV Design of a nuclear reactor 175 176 IAEA - Safety Standards for RRs. IAEA Safety Standards homepage: http://www-ns.iaea.org/standards/default.asp?s=11&l=90 IAEA Safety Standards for RRs:

http://www-ns.iaea.org/standards/documents/default.asp?s=11&l=90&su b=20&vw=9#sf IAEA Safety Report Series: http://www-pub.iaea.org/books/IAEABooks/Series/73/Safety-Reports-Se ries IAEA TECDOCs: http://www-pub.iaea.org/books/IAEABooks/Series/34/Technical-Docume nts Basic Professional Training Course; Module IV Design of a nuclear reactor The views expressed in this document do not necessarily reflect the views of the European Commission.

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