REDUCING CARBON DIOXIDE EMISSIONS FROM DISTRICT HEATING IN FINLAND BY IMPLEMENTING MODULAR NUCLEAR REACTORS A LCA STUDY OF SMRs

Transkript

1 REDUCING CARBON DIOXIDE EMISSIONS FROM DISTRICT HEATING IN FINLAND BY IMPLEMENTING MODULAR NUCLEAR REACTORS A LCA STUDY OF SMRs Alex Back Master s thesis Supervisors: Prof. Margareta Björklund-Sänkiaho Examinator: Prof. Margareta Björklund-Sänkiaho Energy Technology, Vasa Study programme in Chemical Engineering Faculty of Science and Engineering Åbo Akademi University May 28, 2021

2 ABSTRAKT Ett av de största problemen som mänskligheten har för tillfället är hur koldioxidutsläppen ska minskas utan att minska energiproduktionen. Ett av de mest populära lösningsförslagen är att öka andelen förnybara källor. Dessa förnybara källor måste dock ha något slags reservkälla för tider när förnybara energiproduktionsätt inte fungerar optimalt. De lokala fjärrvärmenäten i Finland består av en eller flera kraftvärmeverk (CHP) som producera både fjärrvärme och elektricitet. Dessa anläggningar drivs med inhemska trädbränslen och torv eller i vissa fall med importerade fossila bränslen som kol eller naturgas. Många fjärrvärmeanläggningar står således framför en utmaning att sänka sina koldioxidutsläpp. Syftet med detta examensarbete är att undersöka hur små modulära kärnreaktorer (SMR) skulle påverka koldioxidutsläppen från kraftvärmeproduktionen i Finland. De specifika utsläppen från SMR och vanliga kärnkraftverk jämförs även med varandra för att se om det finns stora skillnader. Som metod valdes LCA med fokus på indirekta CO 2 utsläpp från de material som behövs för att bygga och driva ett kärnkraftverk. Analysen förenklades genom att samla in och jämföra data använd i tidigare LCA analyser, gjorda för konventionella kärnkraft samt data som hittades för bygget vid Olkiluoto och sedan tillämpa mest relevant data för livscykelanalyserna Kraftvärmeverkens CO 2 utsläpp uppskattades på basen av de bränslen som utnyttjades för elektricitet och fjärrvärmeproduktionen. Alla utsläpp allokerades på den värme och el som såldes till slutanvändaren. Resultatet visar att de specifika koldioxidutsläppen från kärnkraftverken är mycket lägre än koldioxidutsläppen från kraftvärmeanläggningarna och fjärrvärmeföretagen. Detta pekar på att SMR är ett möjligt alternativ för att minska koldioxidutsläppen i kraftvärmeproduktion. Resultaten för kärnkraftreaktorerna pekar på att SMR har marginellt mindre specifika koldioxidutsläpp än den konventionella kärnkraftreaktorn. För att utveckla studien som gjordes i detta arbete kunde investeringskostnader för bygget av SMR tas i beaktande. Ett problem som även bör beaktas är det ökade radioaktiva avfallet som ökningen av kärnkraft för med sig. Även data för kärnkraftsbränslekedjan borde uppdateras och möjligheten till kolneutralitet borde beaktas i framtiden om fordonen och maskinerna som används blir mera effektiva och eldrivna. LCA kunde även utföras för olika bränsledrivna CHP-anläggningar för att jämföra CHP-anläggningarna och kärnkraftsanläggningar på ett mer ändamålsenligt sätt. Nyckelord: Livscykelanalys, små modulära reaktorer, kraftvärmeverk, koldioxidutsläpp, kärnkraftverk

3 ABSTRACT One of humanities biggest problem that is in dire need of a solution is how to make sure that the energy demand can be met without a rise in carbon dioxide emissions. Several solutions have been proposed how to solve this problem, with renewable energy sources being one of the more popular solutions. Renewable sources need a backup source, for conditions in which they cannot produce energy. The local district heating network in Finland is made up of one or more combined heat and power (CHP) plants that produce both district heating and electricity. These power plants mainly use domestic wood fuels and peat. In some cases, imported fossil fuels, such as coal and natural gas is used as an energy source. Due to this, many CHP plants face a challenge in lowering their carbon dioxide emissions. The goal of this thesis is to investigate how small modular reactors (SMR) would affect the carbon dioxide emissions released by the CHP plants in Finland. The specific carbon dioxide emissions between an SMR and a conventional nuclear power plant are also compared, to see if there is a big difference between them. This is done by carrying out a LCA on the different indirect carbon dioxide emissions that are released from the materials that are needed to build the nuclear power plants as well as the emissions that are released from processes necessary to operate the plants. The analysis was simplified by gathering and comparing data used in previous LCAs done for conventional nuclear power plants and using data that was found for the Olkiluoto-3 construction. The CHP plant CO 2 emissions were calculated with data on how different fuels were used in the power plants. The specific emission factor was calculated by considering how much district heating and electricity was sold the end customer. The results of this thesis show that the specific emissions for the nuclear power plants are significantly lower when compared to the current CHP plants in Finland, meaning that using SMRs for cogeneration would make it possible to lower the carbon dioxide emissions. The results for the different nuclear reactor types point to SMRs having marginally smaller carbon dioxide emissions compared to the conventional nuclear power plant. For further studies the writer of this thesis recommends that life cycle analyses for CHP plants should be calculated in order to make the comparison of the different power plants fairer. The cost of building and operating SMRs should also be studied further, as well as the problems that might arise from the increased nuclear waste amounts. For the nuclear fuel chain, newer data could also be calculated and estimates on how the emissions from the whole process would change if the vehicles and machinery used would be replaced with more efficient and possibly fully electric counterparts in the future. Key words: Lifecycle analysis, small modular reactors, combined heat and power plants, nuclear power plants

4 Contents ABSTRAKT... 2 ABSTRACT... 3 ACKNOWLEDGEMENTS... 6 LIST OF SYMBOLS AND ABBREVIATIONS INTRODUCTION THEORY AN INTRODUCTION TO NUCLEAR HISTORY DEVELOPMENT OF THE NUCLEAR REACTOR NUCLEAR REACTOR BASICS PRESSURIZED WATER REACTOR BOILING WATER REACTOR PRESSURIZED HEAVY WATER REACTOR ADVANCED GAS-COOLED REACTOR NUCLEAR POWER PLANT BASICS NUCLEAR FUEL NUCLEAR FISSION AND MODERATORS STEAM GENERATOR STEAM TURBINE URANIUM MINING OPEN PIT MINING UNDERGROUND MINING IN SITU LEACH(ISL) MINING URANIUM PROCESSING MILLING AND CRUSHING ORE BENEFICIATION URANIUM EXTRACTION AND PURIFICATION CONVERSION ENRICHMENT CONVERSION OF ENRICHED URANIUM HEXAFLUORIDE INTO URANIUM DIOXIDE FUEL PELLET PRODUCTION BURNABLE ABSORBERS FUEL ROD FABRICATION SPENT FUEL STORAGE AND DISPOSAL SMALL MODULAR REACTORS ADVANTAGES OF SMRs SMR CHALLENGES SMR SAFETY SMR COGENERATION... 70

5 2.7 LIFE CYCLE ANALYSIS COMBINED HEAT AND POWER PLANTS MATERIAL AND METHODS OLKILUOTO WESTINGHOUSE SMR COMBINED HEAT AND POWER PLANTS RESULTS DISCUSSION CONCLUSIONS AND RECOMMENDATIONS SVENSK SAMMANFATTNING REFERENCES APPENDICES

6 ACKNOWLEDGEMENTS I would like to thank my supervisor Prof. Margareta Björklund-Sänkiaho, who made suggestions and helped me find data and helped me when I got stuck in the writing process. Many thanks to my family and friends, who gave my other things to think about during times when I was not working on this thesis.

7 LIST OF SYMBOLS AND ABBREVIATIONS ABWR ADU AGR AUC BWR CHP EPZ IAEA IDR ipwr ISL LCA LOCA NPP NSSS OTSG PHWR PRA PWR SG SMR UOC advanced boiling water reactor ammonium diurante route Advanced gas-cooled reactor ammonium uranyl carbonate boiling water reactor combined heat and power emergency planning zones International Atomic Energy Agency integrated dry route integral pressurized water reactor in situ leach life cycle analysis loss-of-coolant accident Nuclear power plant nuclear steam supply system once-through steam generator Pressurized Heavy Water Reactor probabilistic risk assessment pressurized water reactor steam generator small modular reactor uranium ore concentrate

8 1 INTRODUCTION As the Earth s population keeps rising, so does the energy demand and the carbon dioxide emissions. One of the biggest problems that is in dire need of a solution is how to make sure that the energy demand can be met without a rise in carbon dioxide emissions. Several solutions have been proposed with the most popular being the increase in the share of renewable energy sources in energy production. Solar-, windand bioenergy are used to replace fossil fuels in the energy production sector. The biggest problem related to solar- and wind power is the availability of energy. If there is no wind or it is a cloudy day, windmills and solar cells cannot produce their max output, or in the worst case, no power at all. This can be solved by using energy storage technologies during times when the renewable energy sources produce more energy than is needed by the grid. This method is not without limitations, as today the energy density of batteries is quite low and is expected to rise steadily in the future. Storing energy is also a short-term solution and due to the limitations of the storage capacity in the batteries, this cannot be relied upon during times when wind and sunshine are not available in a usable scale, e.g. during long and powerful storms. Another solution to this problem is increasing the use of nuclear power. Nuclear power plants produce no carbon dioxide emissions during power generation. The carbon dioxide emissions linked to nuclear power are created during different stages of the power plant s lifecycle, with the fuel manufacturing and mining process being the largest contributors. The carbon dioxide emissions created during construction and decommissioning of the power plant are distributed to the whole lifecycle of the powerplant, meaning the longer a nuclear power plant can be used, the smaller the specific emissions for said power plant will be. Small modular reactors (SMRs) are not a new technology. SMRs have been made since the beginning of the nuclear power development and have been mainly used by the militaries of the USA and Soviet Union, with the US military being a pioneer in using the technology. Due to the possibility to make SMRs portable or mobile as well as the long refueling cycle and relatively small amount of fuel needed compared to other power generation methods, the militaries in question saw the potential of the

9 SMRs, and many different types of reactors were created. These SMRs were mostly used to power remote military outposts, submarines and surface ships. These reactors were also used to create district heating, and in some cases, they were also used for desalinization. Small modular reactors have many benefits compared to conventional nuclear reactors. The most important benefit is the modularity of the reactors which gives the possibility to add reactors to power plants easily if there is a need to increase the power output of the plant. This modularity also makes it possible to construct the reactors in factories and provides the possibility to standardize the reactors which, in turn, will make it possible to keep costs lower compared to the conventional reactors. Other benefits that come with the ability to standardize SMRs is the ability to make the reactors more reliable and safer compared to large scale reactors. When the reactors are built in the same way multiple times, it will become easier to detect factors that might cause a decrease in the safety of the reactor. The size of the SMRs also means that the reactors can be safer compared to large reactors, as the fuel amount used in SMRs is lower compared to conventional reactors. Due to the possibility to use less fuel in the SMRs, the damage and pollution that can occur during a catastrophic disaster will be lower compared to large reactors. This, in turn, provides the possibility to decrease the size of the evacuation and safety zones, meaning reactors can be built closer to cities and settlements, with the possibility to utilize the heat from the reactor for district heating. The low power output of SMRs also means that cooling the reactors can be done more easily. The power output of SMRs is usually between 10 and 300 MWe. The goal of this thesis is to make calculations for the specific carbon dioxide emissions for the biggest CHP plants and district heating companies in Finland and to compare the results to LCAs done for an SMR and a conventional NPP, in order to see where it could be feasible to replace CHP plants in the future with nuclear counterparts in order to reduce carbon dioxide emissions. The CHP plants and district heating companies chosen for this thesis all use at least 700 GWh of fuels for district heating and electricity cogeneration. For the independent CHP plants, the efficiency for the net energy sold to the district heating companies compared to fuel utilization for the energy produced was also calculated. LCAs were done for the

10 nuclear reactors as the emissions from nuclear power generation are zero. The lifecycle analyses consider the carbon dioxide emissions created during construction of the power plants, the nuclear fuel chain as well as the commute emissions from construction worker and plant operators. The LCAs are compared to one another to see if there is a big difference in the specific emissions for the different nuclear options. Due to small modular reactors being a technology that has had a new rise of interest, data about emissions from these reactors might be somewhat inaccurate. For the SMR part, the Westinghouse SMR was chosen due to the available data for the reactor as well as the high-power output of the reactor. For the conventional NPP, the new reactor at Olkiluoto was chosen, seeing as it will be the most modern reactor in Finland once it is completed. Data for the SMR and NPP were retrieved from different LCAs published in scientific articles as well as different reports available from different organizations. The data used for CHP-plant calculations are from the district heating statistics for the year 2019.

11 2 THEORY This section aims to give the reader of this thesis a basic understanding of how the nuclear reactor works. The section starts with looking back at the history of how the field of nuclear technology started and what lead to the harnessing of nuclear energy to generate useful forms of electricity and what the early reactors have been used for. This chapter will also explain the biggest differences between a small modular reactor and a conventional nuclear reactor (LR, as well as what merits and possible drawbacks an SMR might have compared to a LR. Furthermore, this chapter will also briefly go through different types of nuclear reactors and the safety improvements that are associated with SMR will also be explained in this thesis. Lastly this chapter will also explain where the CO2 emissions that are associated with nuclear power plants (NPP) by going through the different parts of the life cycle of an NPP. This thesis will also explain what a CHP plant is, what they are used for and how they work briefly.

12 AN INTRODUCTION TO NUCLEAR HISTORY Nuclear technology is considered to have started in 1789, with the discovery of a new element by the German chemist Martin Klaproth, which he named uranium after the planet Uranus. After the discovery of uranium, it took nearly one hundred years before a use for the element was discovered. This discovery led to the creation of a new field of technology. (World Nuclear Association, 2021) Nuclear technology can be defined as engineering solutions that are based on radiation and nuclear reactions. Some purposes that nuclear energy can be used for are electricity generation, thermal energy generation for the population or industry, as well as propulsion energy for different vehicles, mostly for military vehicles and in some cases for civilian vehicles, such as submarines and surface vessels. In general, nuclear technology has three main directions that are being developed by most countries. These directions are nuclear weapons, nuclear energy and, finally, non-power applications such as the medical field. (Khan & Nakhabov, 2020) The nuclear energy field started in 1895, when Wilhelm Roentgen discovered a new type of radiation that he named X-rays. Roentgen discovered that passing an electric current through an evacuated glass tube produced a continuous ionizing radiation. (World Nuclear Association, 2021) This discovery was the spark that led to the start of a field that today uses nuclear technology to generate energy. (Khan & Nakhabov, 2020) A year later, in 1896, Henri Becquerel discovered that an ore which contains radium and uranium, called pitchblende, caused a photographic plate to darken. Becquerel went on to prove that this was caused by beta radiation and alpha particles emitted by the ore. (World Nuclear Association, 2021) Beta radiation is caused by high-energy and high-speed electrons that are released during radioactive decay. (Australian Radiation Protection and Nuclear Safety Agency, 2020) Alpha particles are particles that consist of two protons and two neutrons that are tightly bound together and, like beta radiation, are released during nuclear decay. (Australian Radiation Protection and Nuclear Safety Agency, 2020) Later Henry Villard discovered that pitchblende also emitted a third type of radiation that was named gamma rays. Gamma rays are very similar to X-rays. This phenomenon was given the name radioactivity by Pierre and Marie Curie. The Curies also managed to extract polonium and radium from pitchblende ore in Radium would later be

13 used in different medical treatments and Samuel Prescott proved that radiation could be used to kill bacteria in food. (World Nuclear Association, 2021) Scientists started to understand the reason behind radioactivity. Ernest Rutherford discovered that radioactivity was a spontaneous event that emits an alpha or beta particle from the nucleus and that this process creates a different element. Rutherford would go on to achieve a better understanding of atoms in 1919, when he fired alpha particles from a source of radium into nitrogen and noticed that a nuclear rearrangement was occurring, forming oxygen. In 1911, Frederick Soddy discovered that naturally radioactive elements had a different number of isotopes. In the same year, George de Hevesy proved that radionuclides could be used as tracers, as small amounts could be detected with simple equipment. (World Nuclear Association, 2021) According to the World Nuclear Association (2020), the neutron was discovered in 1932 by James Chadwick. In the same year, nuclear transformation was achieved by Cockcroft and Walton, who bombarded atoms with accelerated protons. A few years later, in 1934, Irene Curie and Frederic Joliot discovered that it was possible to create artificial radionuclides by using the methods that Cockcroft and Walton had used earlier. Enrico Fermi noticed that if neutrons were used instead of protons, it was possible to create a broader variety of artificial radionuclides. Fermi was able to create heavier elements from his test material and succeeded in creating lighter elements by using uranium as the bombarded material. The lighter materials that were produced from uranium were shown in 1938 to be barium and other materials that were around half the mass of uranium. This discovery, made by Otto Hahn and Fritz Strassmann, showed that atomic fission had occurred during the experiments with uranium. The scientists Lise Meitner and Otto Frisch, both of whom worked under Niels Bohr, explained that atomic fission occurred due to the neutron being captured by the nucleus, which causes vibration that leads to the nucleus splitting into two parts that were not exactly equal in weight. Meitner and Frisch calculated that the energy that was released during atomic fission was around 200 million electron volts. In an experiment in 1939, Frisch proved that the calculations were correct. (World Nuclear Association, 2021) The discovery of atomic fission led to an increase in activity in many laboratories

14 around the world. Hahn and Strassmann showed that fission releases neutrons and theorized that it could be possible for the release of these neutrons to create a selfsustaining chain reaction that could lead to the release of an enormous amount of energy. This theory was proved by experiments done by Joliot and his coworkers in Paris, and by Leo Szilard who worked with Fermi in New York. Nils Bohr proposed soon afterwards that it could be possible that the isotope uranium-235 was more efficient than uranium-238 to be used in fission. He also proposed the theory that slow-moving neutrons could be more effective than fast neutrons for use in fission. The later theory was confirmed by Szilard and Fermi, who also suggested that a moderator could be used to slow down the neutrons that were emitted during fission. These ideas were further extended by Bohr and Wheeler in a paper that would become known as the classical analysis of the fission process. This paper was released two days before the beginning of the Second World War. (World Nuclear Association, 2021) One of the greater challenges at this point was the availability of U-235. Natural uranium ore contains 99.3% U-238 and only 0.7% U-235. Because the isotopes have similar chemical properties, there were some difficulties with separating the two isotopes into their own elements. There were some differences in the physical properties of the isotopes, and by exploiting these differences, it was possible to separate the two. The process that increases the proportion of U-235 is called enrichment. (World Nuclear Association, 2021) The final piece that was needed for fission and was later used for the atomic bomb concept was provided by Francis Perrin. Perrin introduced the concept of the critical mass of uranium that was needed to achieve a self-sustaining reaction. This concept was further researched by Rudolf Peierls, who presented calculations that were of great importance for the creation of the atomic bomb. Perrin and his group also demonstrated that a mixture of uranium and water could be used to sustain the chain reaction, if external neutrons were injected into the system. The group also discovered that by using neutron-absorbing materials it was possible to control the neutron multiplication, thus controlling the nuclear reaction. This discovery later became the corner stone for the use of fission for energy generation purposes. Peierl s former teacher, Werner Heisenberg, who presided over the German nuclear

15 energy project, had also done calculations that showed that a nuclear fission chain reaction was possible. He proved that by slowing down and controlling the fission chain reaction, the fission chain could be used to generate energy. Heisenberg s calculations also showed that if this chain reaction was not controlled and left to run loose, it would lead to a nuclear explosion that would be many times stronger than conventional explosives. Heisenberg suggested that natural uranium could be used in a uranium machine, also known as a nuclear reactor, with heavy water as the moderator. The researchers were unaware of the fact that delayed neutrons could be used to control the reactor. Heisenberg also suggested that by using pure U-235, an explosive could be created, but he rejected this idea due to the needed amount of U- 235, which at the time was a rare isotope, being too high to be practical. In 1940, a colleague and a friend of Heisenberg s, named Carl Friedrich von Weizsäcker, studied publications by scholars from Britain, Denmark, France and the United States, and came to the conclusion that if a uranium machine could be used to sustain a nuclear fission chain reaction, it could be used to transmute U-238 into element 94. Element 94, also known as plutonium, could according to von Weizäcker also be used as a powerful explosive. Weizsäcker submitted a patent for the use of a uranium reactor to create plutonium. The German military declared that this project was impractical, as it would require more resources than were available at the time. The knowledge of this idea later became the incentive for the creation of the atomic bomb and later for the creation of nuclear reactors by the Allied forces. (World Nuclear Association, 2021) The MAUD committee, established in Cambridge, consisted of some of the best scientists, made two big breakthroughs. The committee obtained experimental proof that the nuclear fission chain reaction could be sustained with slow neutrons by using a mixture of uranium oxide and heavy water, proving that the output of neutrons was greater than the input. The second breakthrough was made by Bretscher and Feather, who proved that fission in U-235 was more common than U-238. Bretscher and Feather also showed that U-238 was more likely to form the isotope U-239. U-239 was a new isotope that would emit an electron and form a new element with the atomic mass of 239 and the atomic number 93. The element would also emit an electron and form another element that would also have the atomic mass of 239, but

16 the atomic number would change to 94, in other words, plutonium. Bretscher and Feather claimed that this new element was fissionable by both fast and slow neutrons, and due to the different chemical properties compared to uranium, it could more easily be separated. In March 1941, the fission cross-section of U-235 was finally confirmed. This was the final piece of information that was needed for the creation of the nuclear reactor. With this discovery, it was confirmed that slow neutrons were much more efficient in causing fission when collided with U-235, compared to using fast neutrons. The MAUD committee released two summary reports in According to the World Nuclear Association (2020), the reports released were named Use of uranium for a bomb and Use of uranium as a source of power. (World Nuclear Association, 2021) The second report by the MAUD committee explained that by using a controlled fission of uranium, it would be possible to generate energy in the form of heat. This heat could be used by different machines. The report also concluded that this process could also be used to provide radioisotopes in a large enough quantity so that the isotopes could be used as substitutes for radium. The report stated that this fission process could be achieved by using heavy water and graphite as moderators if the fast neutrons were used in the reaction, and that if the isotope U-235 was used, regular water would be enough. The committee did not recommend pursuing this technology at the time but did state that this technology was promising and would be of great use in times of peace. This meant that the development of the nuclear boiler was put on hold until the war ended. The committee suggested that it would be more useful to focus the efforts on creating the atomic bomb, a suggestion that led to the creation of the Manhattan Project. (World Nuclear Association, 2021) DEVELOPMENT OF THE NUCLEAR REACTOR As WWII ended, the peaceful use of the nuclear fission reaction could now be developed. Thanks to development of the atomic bomb, scientist had created a range of new technologies and realized the vast potential of the heat generated from the fission reaction. This meant that there was now a new focus on utilizing the energy released from fission to generate steam and electricity. This new form of energy generation method would allow for the creation of compact power sources and these power source had the potential of being long-lasting. Due to potential compactness of

17 the power source the application range was wide. The reactors could be used to power ships and most importantly submarines. The reactor at the Argonne National Laboratory was the first nuclear reactor that produced electricity. The reactor, that was named Experimental Breeder reactor (EBR-1), was started up in December (World Nuclear Association, 2021) EBR-1 was a liquid metal-cooled fast reactor that was used to demonstrate that it was possible for a reactor to create more fuel than it consumed. Sodium-potassium alloy was used in the reactor, as liquid metal coolants have better heat transfer compared to water. The reactor was first used to power four 200-watt lights and would later be used to power the whole research facility. The reactor would later experience a partial core meltdown during a coolant flow test, an unfortunate event that lead to the advancement of nuclear technology as it gave scientists a better understating of the thermal capacity of the materials that were used during the test. Today it is possible to take a free guided tour at the facility during weekdays. The reactor was later declared a National Historic Landmark by President Lyndon Johnson in (Office of Nuclear Energy, 2019) The success of the EBR-1 was one of the factors that lead to President Eisenhower to propose the Atoms for Peace program. This program caused the scientific society to focus more research efforts towards the goal of generating electricity with the use of nuclear energy and was the spark that defined the direction of the civil nuclear energy development in the USA. Around the same time the Soviet Union had also made progress with nuclear reactors. The Soviets modified existing graphitemoderated channel-type plutonium production reactors to produce heat and electricity. The world s first nuclear powered electricity generator AM-1 was started in Obnisk in June The AM-1 had a capacity of 30 MWt or 5 MWe. This reactor would serve as the protype for future graphite channel-type reactors. Among these reactors is the infamous Chernobyl-type RBMK reactor. The AM-1 was used until 2000 as a research facility and isotope production. (World Nuclear Association, 2021) The first fully commercial PWR nuclear reactor was made by Westinghouse in The reactor had a capacity of 250 MWe and was built in Yankee Rowe. Around the same time, the first BWR developed by the Argonne National Laboratory and

18 designed by General Electric was also taken into use. The first large scale nuclear reactor was the RBMK reactor that started its operation in Sosnovy Bor near Leningrad in The RBMK had a capacity of 1000 MWe. (World Nuclear Association, 2021) According to the World Nuclear Association (2020) the nuclear power industry suffered from decline and stagnation from the late 1970s to about During this time period many of the ordered NPP were cancelled and the share of nuclearpowered electricity kept a share between 16-17%. Uranium prices also dropped during that time as a reaction to the decreased interest in building NPP. This caused many oil companies that had entered the uranium field to bail out. The NPP field started to recover in towards the end of the 1990s as the first third generation NPP was commissioned in Japan. The reactor in question was the Kashiwazaki-Kariwa 6 an advanced BWR with a capacity of 1350 MWe. As the demand of electricity began to grow worldwide and the concerns regarding climate change began to rise, the need for a power source that was able to produce electricity with low carbon dioxide emission made interest in NPP begin again. Another factor that lead to the new interest in NPP was energy security awareness. The last and most important reason was the scale of the projected increase in electricity demand worldwide. This electricity demand was especially noticeable in fast developing countries. All these factors as well as the new generation of nuclear power reactors lead to start of new NPP constructions. The first of the to be late third generation NPP was ordered by Finland in 2004 and is still under construction at the time of writing this thesis. The NPP that is being constructed in Finland is the 1600 MWe European PWR (EPR), and a similar NPP is also being built in France. In the USA two new Westinghouse AP1000 units are being built. The World Nuclear Association (2020) reports that the plans of NPP construction in Europe and the USA are overshadowed by the plans in Asia. Especially China and India are planning to increase their NPP capacity greatly. China has plans and the capacity to build more than 100 new large scale NPP. These NPP are mostly done according to western design or an adaption of said design. There are also designs done locally. (World Nuclear Association, 2021)

19 2.2 NUCLEAR REACTOR BASICS A nuclear reactor produces energy by splitting atoms in a reaction that is called nuclear fission. Nuclear fission is a process in which heavier nuclei are split into lighter nuclei by bombarding the nuclei with neutrons. The energy released from splitting atoms in a nuclear reactor is used to generate heat, which is used to heat water or gas and this heated medium is then used to turn water into steam. This steam is then used to spin turbines in order to run generators and generate electricity. In this section, a brief introduction will be given of how a nuclear reactor works and different types of nuclear reactors will be discussed. (World Nuclear Association, 2021) PRESSURIZED WATER REACTOR The most common type of nuclear reactor that is used globally is the pressurized water reactor. Around 300 of all active nuclear reactors that are used for power generation are PWRs. The reactor core of a PWR usually contains fuel assemblies, each of which contains fuel rods. PWRs use light water as both coolant and moderator. More specifically it is the primary coolant water that functions as a coolant. If any of the primary coolant starts to turn into steam in the reactor core, the decrease in density of the steam compared to water will make the fission reaction slow down. This is called a negative feedback system and is used in PWRs as a safety feature. What makes the PWR design special is that the water that is used as primary coolant is circulated through the core at a high pressure, usually at 150 times the atmospheric pressure, to allow it to be heated up to 325 C without turning into steam. In order to keep the primary coolant at high pressure, a pressurizer is used. The primary reactor coolant is then used to heat up the coolant in the secondary coolant circuit to temperatures that generate steam. A heat exchanger is used to make the heat transfer happen. The steam that is generated in the second coolant circuit is then used to drive the turbine that, in turn, runs the generator in order to generate electricity. After going through the turbine or turbines, the steam is condensed back into water and led back into the heat exchangers to repeat this process. The basic layout of a PWR can be seen in figure 1. (World Nuclear

20 Association, 2021) Figure 1 Basic PWR layout (World Nuclear Association, 2021) BOILING WATER REACTOR The boiling water reactor functions almost in the same way as the PWR. Key differences between these reactors is that in a BWR only a single coolant circuit is used. The pressure in the coolant circuit is kept at around 75 times atmospheric pressure in order to allow for the water to boil at 285 C in the core. The World Nuclear Association (2021) observes that BWRs are designed to operate with around 12 15% of the water in the core in steam phase in the top part of the core. After boiling the water to steam, the steam is flows through drier plates, after which it passes through to the turbines. It should be noted that the water is contaminated with traces of radionuclides and the turbines should be shielded and include radiological protection in order to protect the maintenance crew from radiation. The radioactivity in the water is short-lived as the radioactive material mostly consists of Nitrogen-16, an isotope with a half-life of 7 seconds, making it possible to enter the turbine hall almost immediately after reactor shutdown. A BWR can contain up 750 fuel assemblies and each of these assemblies can contain fuel rods. This means

21 that the reactor core can contain up to 140 tons of uranium. A basic BWR can be seen is figure 2. (World Nuclear Association, 2021) Figure 2 Boiling water reactor (World Nuclear Association, 2021) As mentioned earlier, the flow of steam is used to create electricity. In a BWR, this is done by having the steam that is boiled in the reactor move towards the steam turbine. In the steam turbine, steam is used to drive the turbine which converts the energy of the steam in shaft torque by expanding the steam from around 6.9 MPa to vacuum pressure. The torque is used to drive the generator that in turn generates electrical energy. The expanded water can be returned to the reactor by cooling it. This is done by using condensers to condense the steam into water. After the condensers the water can be fed into heat exchangers, after which the water can be returned to the reactor by the reactor feed water pumps. Using a direct cycle forced circulation is a standard for commercial BWRs, as it allows setting recirculation loops into the direct boiling water cycle and makes it possible to increase the coolant flow speed and power density in the reactor core as well as makes it possible to control the thermal power of the reactor by adjusting the flow rate of water recirculation. Some of the feedwater, combined with the drain from the steam

22 separator of the reactor, is dropped down between the pressure vessel s internal wall and shroud of the reactor core and is moved to the recirculation loops in which the pressure is increased by the recirculation pumps outside the vessel, and is then sent to the reactor core form the bottom of the vessel by internal jet pumps. (Tadashi, 2017) The development of the BWR led to creation of the Advanced boiling water reactor. The ABWR has managed to improve the safety, reliability, economic efficiency, and operability of the reactor. It has also reduced the amount of radiation exposure as well as the radioactive waste amounts. (Tadashi, 2017) PRESSURIZED HEAVY WATER REACTOR The pressurized heavy water reactor was first developed in Canada in the 1950s as the CANDU. PHWRs use natural uranium and heavy water (D2O). Heavy water is a more efficient moderator compared to light water, making it possible for natural uranium to be used instead of refined uranium. PHWRs produce more energy per kilogram of mined uranium when compared to other reactor types, but they also consume more fuel than the other reactors. The heavy water is kept in the calandria, which is basically a large water tank. What makes the calandria different from a water tank is the hundreds of horizontal pressure tubes that form the channels for the fuel. Using pressure tubes allows for the reactor to be refueled during operation, as individual pressure tubes can be isolated from the cooling circuit. The pressure tubes are cooled with flows of heavy water that is kept at pressures around 100 times atmospheric pressure. Due to the high pressure, the primary coolant circuit can reach temperatures of 290 C. The energy of the heated heavy water in the primary coolant circuit is used to heat up the light water that is used as a coolant in the secondary coolant circuit. The light water is turned into steam in the steam generator and is led into the turbines to power the generators. (World Nuclear Association, 2021)

23 Figure 3. CANDU pressurized heavy water reactor (World Nuclear Association, 2021) The fuel assemblies in the CANDU reactors are made up of bundles of 37 half-meter long fuel rods, with each containing fuel pellets and a zircaloy tube shell. A support structure that contains twelve of these bundles lying end to end are in a single fuel channel. PHWRs use control rods that penetrate the calandria vertically and as a secondary shutdown system gadolinium can be added to the moderator. The CANDU design can use different forms of uranium, unlike the light water reactors. It is possible to use recycled uranium that has been reprocessed after being used in light water reactors. CANDU reactors can also use depleted uranium from enrichment plants, if reprocessed uranium is mixed into it. It is also possible to use thorium as fuel for CANDU reactors. Figure 3 shows the layout for the CANDU reactor. (World Nuclear Association, 2021) ADVANCED GAS-COOLED REACTOR Advanced gas-cooled reactors are the second generation of gas-cooled nuclear reactors used by the British. What makes gas-cooled reactors different from the reactors mentioned so far in this thesis is the fact that water is not used as a primary coolant nor as a moderator. In AGRs, graphite is used as a moderator and carbon

24 dioxide is used as the primary coolant. The carbon dioxide circulates in the core and can reach temperatures up to 650 C. The high primary coolant temperature makes it possible to reach high thermal efficiency number up to 41%. The carbon dioxide is then used to generate steam in the steam generator and the steam is used to run the turbines. The control rods in an AGR are used to penetrate the moderator and, as a secondary shutdown system, nitrogen can be injected into the coolant. Figure 4 gives a better understanding of the layout of an AGR. (World Nuclear Association, 2021) Figure 4. Advanced gas-cooled reactor (World Nuclear Association, 2021) The World Nuclear Association (2021) states that the fuel used in AGRs consists of uranium pellets that are usually enriched to % and are incased in stainless steel tubes. (World Nuclear Association, 2021) 2.3 NUCLEAR POWER PLANT BASICS The nuclear power plant can crudely split into two areas, the nuclear power area and the rest of the cycle. The nuclear power area is the part the of cycle that provides the steam for the process and the equipment that is in this area is subject to much stricter safety requirements than the rest of the cycle. This section will discuss what parts

25 these different areas contain and briefly explain their functions in the nuclear fission process. (Belyakov, 2020) NUCLEAR FUEL As an energy source, nuclear fuel differs greatly from other forms of fuel. The energy density of the fuel is extremely high, which means that small amounts of uranium are needed to generate power. According to Letcher (2014), the volume of uranium that is needed to provide energy for a single human s lifetime is roughly around the size of a golf ball. The energy contained in this amount is almost equal to the energy that was released by the atomic bomb dropped in Hiroshima near the end of WW2. This energy density makes nuclear power a viable energy source for humanity for millennia. Another benefit that is provided by the energy density of uranium is the possibility for countries to stockpile uranium, to provide energy for tens or even possibly for hundreds of years. This benefit is one of the reasons that have led countries to pursue nuclear power plant development. The reason for the high energy density is the nature of the atomic nucleus. In elements that are heavier than hydrogen, the number of protons crammed into the atoms give the atoms a positive charge. The positive charge caused by the protons gives rise to extremely high electrostatic repulsion in the nucleus. For the nucleus to remain intact, a stronger force called nuclear force is needed. Later in this thesis, we will investigate further the steps that are needed in order to turn natural uranium into a useable nuclear fuel, as this section only gives basic information about some of the properties of nuclear fuel. (Letcher, 2014) Uranium is found in nature and is almost as common as tin or zinc. Natural uranium contains the isotopes uranium-235 as well as uranium-238. Out of these two isotopes, U-235 is the most common one to use. However, natural uranium ore does not contain big concentrations of U-235, as only 0.1 1% of the ore consists of U-235. (Letcher, 2014) The most common composition of natural uranium according to the World Nuclear Association (2021) is 0.7% U-235, while the rest of uranium consists of U-238. Uranium reserves are also not available in every country, as the reserves are scattered around the world. The biggest producers of uranium are Kazakhstan, Canada, Australia, Nigeria and Namibia, and these five countries produced 79% of all the uranium in Compared to other fuel forms, uranium is not sold on the

26 open market, as buyers and sellers negotiate supply contracts privately. (Letcher, 2014) Uranium oxide (UO2) is the most common type of fuel used in nuclear reactors. The melting point for uranium oxide is 2800 C. The fuel is mostly enriched and is in the form of pellets that usually have a diameter of 1 cm and the length can reach up 1.5 meters. The pellet is usually surrounded by a zirconium alloy tube to form the fuel rod. Zirconium has excellent properties that make it useful as an alloy. In order to use zirconium as alloy material for the fuel pellets, it first needs to be purified. Zirconium naturally contains hafnium, which is a neutron absorber and must be removed before the zirconium can be used. Very pure zirconium is hard, corrosionresistant and transparent to neutrons, all of which are properties desired for an alloy. (World Nuclear Association, 2021) To even out the performance of new and old fuel in reactors, burnable poisons are used in either the fuel or coolant in reactors. Burnable poisons are neuron absorbers that decay under neutron exposure and compensate for the gradual buildup of neutron absorbers in spent fuel and, in doing so, allow for more fuel to be burned. For naval reactors, the use of burnable poison is almost a must, as refueling these reactors can be difficult and thus the refueling cycle is often designed to be very long compared to other reactors. The most efficient burnable poison that is used today is gadolinium. Gadolinium (GdO2) is added to ceramic fuel pellets during manufacturing and is used at up to 3 grams per kilogram of fuel. The World Nuclear Association (2021) mentions that when using gadolinium as a burnable poison, the nuclear fuel must be enriched to a higher degree to compensate for the added poison. (World Nuclear Association, 2021) The refueling intervals for nuclear reactors vary between different reactors. For large nuclear reactors the interval is usually 12, 18 or 24 months. During refueling a third of the fuel rods are replaced with new ones. For small modular reactors this refueling interval can be much longer. Some reactors need to be shut down for refueling, however, this is not the case for the CANDU or RBMK reactor types, as these reactors use pressure tubes instead of a pressure vessel to enclose the reactor core. A single pressure tube can be removed from the core while the reactor is active. (World Nuclear Association, 2021)

27 2.3.2 NUCLEAR FISSION AND MODERATORS Letcher (2014) claim that nuclear energy is derived from the competition between electrostatic and the strong nuclear force. He further states that for light elements, such as hydrogen, the more energetical favorable option is to fuse together in a process that is called nuclear fusion. By fusing together with each other, the lighter elements form heavier nuclei. This process does not occur in heavier elements such as uranium. In heavier materials another process that is called nuclear fission takes place. Nuclear fission is a process in which elements split into lighter elements. This process can occur naturally, however most elements can only go through nuclear fission by absorbing neutrons. By absorbing a neutron, the balance between the strong and electrostatic forces is disbalanced. Due to the disruption in balance, the nucleus begins to split. When the nucleus splits, the fission fragments are repelled by the electrostatic force. In some cases, this process can occur at different neutron energies. In nuclides that this happens in are called fissile. Neutrons are also released as fission fragments in nuclides that are fissile and can cause the reaction to become self-sustaining. When a nuclear fission reaction is self-sustaining the reaction is called a chain reaction. For a chain reaction to be useful it essential that it can be controlled. These controlled chain reactions are used by nuclear power plants to generate energy. (Letcher, 2014) The fission fragments that are released during chain reactions are called fission products and have high kinetic energies. Due the high kinetic energy, fission products that collide with the surrounding material, causing the material to heat up. This is due to the collision of atoms in the material. In nuclear power plants this is used to heat up water and generate steam, which can be used to run turbine to generate electricity. The nuclear reactor can either use steam directly generated in the reactor to run the turbine, or the heat from the water can be transferred to another coolant. It should be noted that the fission products that are released during a nuclear fission reaction are highly radioactive. This radioactivity generates heat, which can cause damage to the reactors if proper care is not taken. The accident in Fukushima in 2011 was caused in part due to the fission products damaging the reactor. (Letcher, 2014)

28 As mentioned before, nuclear power plants use controls chain reactions to generate energy. According to Letcher (2014) the possibility for a nuclear fission to occur depends greatly on the energy of the neutron that is used to start the reaction. Fission is more unlikely to happen if fast neutrons are used compared to slow neutrons. The terms fast and slow are used to describe neutrons that have high and respectively low neutron energies. If neutrons with high energies are used to try to initiate fission, the likelihood for the process to start is lower, as the possibility for the high energy neutron to be absorbed is low. If fission is started successfully with fast neutrons the number of neutrons released will usually be much higher compared to fissions that are started with slow neutrons. Due to the quantum nature of the neutron, fission is much more likely to occur at lower neutrons energies. In most nuclear reactors, fission is achieved by moderating neutrons, in other words slowing down neutrons to energies that are in thermal equilibrium with the moderating material. When choosing the material for moderators that are going to be used it is imperative, that the moderator material does not absorb too many neutrons. If too many neutrons are caught by the moderator, the possibility for fission might decrease. If a neutron is absorbed and no fission takes place this is called parasitic neutron loss. Parasitic neutron loss can be avoided by choosing moderator materials that only require a few collisions to take place in order to slow down the neutrons. This usually means that the moderator material is required to have a nucleus that has a mass close to neutrons mass, as well as a sufficient density. Moderator materials that have these properties are quite few and the most common materials that are used as are carbon, in the form of graphite, and hydrogen and deuterium as light water and respectively heavy water. Heavy water is the most troublesome moderator to manufacture, as the hydrogen in water only consist of 0.016% of deuterium. (Letcher, 2014) Out of these three moderators, light water is the one that is used most. This is due to different reasons, but the main reasons being the availability of light water and the cost-effectiveness. Letcher (2014) also mentions that another factor that makes light water an attractive choice is that light water is more effective at slowing down neutrons when compared to heavy water. This is due to the mass of hydrogen being closer to the mass of a neutron than deuterium. The lower mass of light water also makes it possible to make smaller reactor cores. One big demerit that light water has

29 is that is quite parasitic and will absorb much more neutrons than heavy water. This makes using natural uranium impossible with light water as the moderator. This means that the uranium that is used with light water as a moderator must be enriched to reach U-235 concentrations that make it possible for the fission reaction to be selfsustaining. Heavy water can use natural uranium as fissile material and can be seen as the moderators biggest merit. (Letcher, 2014) Light water can also be used as a coolant in the reactor. When light water is used as a moderator and coolant at the same time, the reactor can be made to be selfregulating. This means that the reactor can automatically control the reactivity that is taking place in the reactor. When fission takes place, heat is released, and the water temperature will begin to rise. As the temperature rises, the density of the water will also decrease and this in turn will decrease the levels of moderation in the reactor, as neutrons can now pass more easily through the moderator that is also used a coolant. This means that the reactivity in the reactor will decrease which in turn stops the temperature from rising further. Modern NPP use this as a safety feature, called a negative feedback, to control the fission and prevent it from going out of control. (Letcher, 2014) Letcher (2014) also states that control rods can be used to control the fission reaction. They are mostly used to shut down the nuclear reaction as they are lowered down into the reactor. When the control rods enter the reactor, neutrons that are released during the fission reaction are captured by the control rods instead of reaching the fissile material. As the neutrons that would usually make the fission reaction selfsustaining are now absent, the reaction will thus stop the chain reaction and in doing so, shut down the reactor. (Letcher, 2014) STEAM GENERATOR Nuclear power plants use steam generators to turn water into steam. It should be noted that not all reactor types need a steam generator. BWR use the steam generated in the reactor core to run turbines and therefore does not need a steam generator. Nuclear power plants that use PWRs usually contain between two to four steam generators depending on the design and size of the plant. PWR steam generators have an element that is different from steam generators that are used in fossil-fired

30 power plants. In PWR SGs the boiling occurs in the shell side of the heat exchanger rather than in the tubes. By doing this it is possible to keep the primary reactor coolant at a high velocity and pressure, which in turn allows for efficient heat transfer from reactor coolant to the low-pressure steam system. The primary reactor coolants temperature is usually in the range of K ( C) and the pressure is kept high at around 15 MPa according to Riznic (2017). (Riznic, 2017) The most common type of SG is the vertical cylindrical vessel setup that have inverted U-tubes in the lower section. The upper section contains steam-water separators or moister separator. The primary reactor coolant flows through the U- tubes and rises to the hot leg side of the SG. From the hot leg side, the coolant is then lead into the cool leg side of the SG which it exits at a temperature around 560 K (287 C). These temperatures only apply for new NPP and in older plants these temperatures are lower. Feedwater is fed into the downcomer by leading it through a feedwater nozzle, after which is fed through a feedring. In the downcomer the feed water and recirculated water coming from the moisture separator. The water mix is then usually fed through a gooseneck pipe assembly. After the gooseneck pipe assembly, the water mix is led downward into the feedring, also called the feedwater header, to avoid risks of flow stratification. Doing this minimize the risk of high thermomechanical stressing in the feed water piping system. J-Tube vents are usually installed on top of the feedring to avoid steam plugs as well as prevent feedring drainage. After this the water from the downcomer flows to the bottom of the steam generator via the top of the tubesheet. After this the water is led to the tube bundle in which the water is turned to steam. Not all the secondary water is turned into steam, as only 20 25% of the water is turned to steam per passthrough. The remainder of the secondary water is recirculated. The method described does not use an integral preheater or economizer. It is recommended that the reader studies figure 5 for a more clear picture of the setup of a recirculating PWR steam generator. (Riznic, 2017)

31 Figure 5 Layout of a recirculating PWR steam generator (Riznic, 2017) If an integral preheater is used, then no feedrings are used in the SG. In this type of design all the feedwater is instead forced to flow through the preheater. The feedwater is fed through a nozzle into the cold leg side of the tube bundle, located near the tubesheet. The feedwater is forced to flow through the preheater with crossflow over baffle plates on the cold leg side. Auxiliary feedwater enters the upper part of the vessel via a separate nozzle. The heat from the primary fluid leaving the steam generator is used by the preheater to increase the temperature of the feedwater until it almost reaches saturation temperature. After reaching this temperature, the feedwater is mixed with the recirculating water that flows down from the tube bundles top. In certain types of plants two feedwater nozzles are used. One of these nozzles is used for the auxiliary feedwater then the operation is started,

32 while the other is used for the main feedwater during operation. The auxiliary feedwater can be taken from a separate auxiliary tank, in which the water is at ambient temperature, or directly from the feedwater tank. (Riznic, 2017) Another type of steam generator is the once-through steam generator, which is the smallest type of steam generator as it lacks moisture separators. The primary reactor coolant in OTSGs is pumped from the top to the bottom through tubes. The secondary coolant is made to flow in counter-flow direction on the outside of the tubes, going from the bottom to the top. The secondary-system water flows through a feedring above the ninth tube support plate. Here the water is mixed with steam that is drawn from the tube bundle area. After mixing the secondary-system water and steam, the mixture is then preheated until it reaches saturation temperature. The saturated water is then fed through the annulus, through the lower tubesheet until it reaches the tube bundle in which water is turned into steam. The steam that has been superheated will flow outwards in a radially fashion down the annulus into the steam outlet connection. In OTSG most of the secondary coolant is evaporated in the SG in a single pass. (Riznic, 2017) To avoid erosion, as well as residual stresses the tubing in OTSG is treated with different high temperature treatment processes. Some of these heat treatments include high mill-annealing temperatures that are done the remove the internal stresses of the material used for tubing. Temperatures for this process can reach up to 1338 K. This treatment is then followed by holding the SG at a temperature of 893 K for up to ten hours after assembly. This method makes the OTSG tubing very resistant to primary water stress corrosion cracking. It should be noted that this heat treatment method might deplete the chromium in the grain boundary, which makes the primary as well as the secondary sides vulnerable to sulfur species attacks. (Riznic, 2017) STEAM TURBINE The steam turbine has been used in nuclear power plants since the beginning of commercial use of NPPs for power generation. The steam turbine had earlier been used in fossil-fired power plants and it was clear from the start of the development of NPPs that using a turbine was the best method to generate electricity. The main

33 function of a steam turbine is to use steam to drive the generator. This is done by converting the steam flow into mechanical energy in the turbine. The mechanical energy in turn is used to drive the generator which changes the mechanical energy into electrical energy. Different types of NPPs have differences in this process, but the main concept remains the same. (Tadashi, 2017; Belyakov, 2020) The most important factor that controls how much electricity an NPP can generate is the nominal thermal power output of the power plant. The nominal thermal power output is a factor that depends on the design of the NPP itself as well as the quantity and quality of the steam that is produced by the nuclear reactor. The pressure of the steam in NPP is always subcritical and is linked to the coolant parameters that is used in the cooling loop. The temperature of steam is dependent on the type of reactor that is used in the NPP and in some cases the steam can be very saturated, and this can cause the moisture content to become too high and warrant further treatment throughout the NSSS. The process that generates hundreds of megawatts of electricity requires large flows of steam into the turbine. To ensure that the steam flow into the turbine is adequate, many NPPs choose to use one large steam turbine per nuclear reactor. The type of turbine, configuration as well as the steam parameters depend on the type of nuclear reactor that is being used in the NPP. Usually the turbines that are used, are specifically designed for said NPP and the reactor that is being used. (Belyakov, 2020) According to Belyakov(2020) turbines can generally be divided into two major types. The types are defined based on the speed of the rotating shaft. The first type is the fast or full-speed steam turbine that runs at speed of rpm. The fullspeed steam turbine is used to run a synchronous generator, and the size of the turbine increases fast if the output is increased. This turbine type suffers from dynamic loads and stresses at massive loads. The second turbine type is the slow or half-speed steam turbine. As the turbine name implies this turbine runs at half of the speed of the fast turbine, at around rpm. This type of turbine is larger and heavier than its fast counterpart. Unlike the fast turbine, the slow turbines size does not increase as much, if the capacity of the turbine is increased, which means at some point the size of the two turbines types will even out. (Belyakov, 2020)

34 There are also different turbine setups available for NPPs. The most typical configurations used in NPPs are the double- and three-pressure units respectively. In the double-pressure system the steam first goes through a high-pressure turbine section. After this section the steam is reheated and separated from the moisture content, after which the steam is led to the low-pressure turbine section. After going through the low-pressure section, the steam is lead into the condenser. In the threepressure units, steam first flow through the high-pressure section in the same way as the double-pressure system. However, after the high-pressure section the steam is first reheated and lead into an intermediate-pressure system, in where a part of the energy is transmitted to shaft. The steam is then lead into the low-pressure section. The number of sections that are used depend on the capacity of the NPP. Generally, steam flows must be organized in a way that have the least negative effects on the shaft line and support bearings. This usually means that for the low-pressure sections, an axial double-counterflow setup of cylinders is used. If a three-pressure design is used the high- and intermediate-pressure section only have one steam flow into them, and the direction of the flow is different along the shaft line to minimize the strain on the overall structure. Steam turbines use a reheater and a moisture separator to treat the steam after it exits the high-pressure section. (Belyakov, 2020) It should be noted that the type of nuclear reactor that is being used will also affect the turbine design. When comparing BWRs and PWRs there are notable difference in the way in which steam is used to generate electricity. In BWRs, the reactor pressure vessel is designed for lower pressure compared to PWRs as the steam pressure that is used in the turbines is close to the reactor pressure. The steam itself however contains radioactive materials. Due to the radioactive materials in the steam some things should be taken into consideration when designing the turbine. The turbine needs to have radiation shields for all components to avoid damage from the radiation that is released from the steam. It is also important that some type of measurement device is used in the turbine to make it possible to notice if there are any leakage in the turbine. By including this sort of safeguard against leakage, any possible damage and radiation contamination can be kept as low as possible. According to Tadashi (2017) turbines designers should also take the decontamination possibility into consideration when creating the turbine. Turbines and their

35 components that are used in BWRs need to go through decontamination before any maintenance can be done on them. (Tadashi, 2017) The turbines used in PWRs differ from turbines in BWRs in the aspect that these turbines are not in contact with steam that contains radioactive materials. This means that there is no need to include radiation shields or protection against radiation in maintenance work in turbines that are used in PWR. In short PWR uses two different coolant system. The steam that is used to generate electricity is created by the secondary coolant system. The heat of the pressurized water is transferred from the primary coolant system to the secondary coolant systems steam generator. The water in the secondary coolant system is then converted into steam in the steam generator and this steam is used to run the turbines to generate electricity. As the steam that is used to run in the turbines is not in direct contact with the water from the reactor the risk for nuclear contamination is nonexistent if a leak should occur in the turbine cycle. (Tadashi, 2017) NPP steam turbines are usually operated as base load. According to Tadashi (2017) base load operation means that steam turbines are optimized in the design stage in order to minimize the exhaust loss. This is done by applying the optimum last-stage blade and number of exhaust flows. Steam used in high pressure turbines is saturated with a wetness of 0.4%. After the high-pressure turbine, the steam is treated in moisture separators to remove the excess water fraction. The wetness of the steam increases when its energy is converted to power output in the high-pressure turbine. The steam is reheated in the reheating cycle, as the thermal efficiency can be increased with around 2% compared to cycles that don t reheat the steam before being led into the low-pressure section. Reheating the steam also reduces last-stage blade erosion that can be caused by water droplets in the saturated steam that hit the blades at high speeds. (Tadashi, 2017) Compared to fossil-fired power plants, the turbines in NPPs require more steam to generate the same amount of power. Tadashi (2017) states that is due to the total heat drop of the main steam to be much lower in an NPP compared to fossil-fired power plants. In NPP the steam consumption per power output will be times higher due to the main steam pressure being lower. This also causes the steam volume per power output to become 4 5 times larger. Due to the larger volume flow of steam,

36 the different components in NPP turbines need to be made larger when compared to the turbines in fossil-fired power plants. Turbines used in NPP are operated in saturated steam and as mentioned earlier saturated steam can cause erosion in the turbine. The water droplets in the steam can cause the efficiency of the turbine to decrease if the impact erosion is not taken into consideration. This has led to the designers of turbines to incorporate anti-erosion measure for NPP turbines. Moistureextracting blades have grooves longitudinal to the blade s direction on the suction side that capture water droplets and discharge the from the blade due to the centrifugal force of the spinning blades. The stationary turbine components can also experience wire erosion. This erosion is usually caused by drain intrusion through small clearances in fastened flanges which overtime form grooves in the flanges. This can be countered by covering the flanges with anti-erosion metal. (Tadashi, 2017) 2.4 URANIUM MINING As mentioned earlier, uranium is almost as common as tin or zinc. According to World Nuclear Association (2021) traces of uranium can be found almost anywhere on Earth. Besides the uranium deposits beneath the earth, vast amounts of uranium can also be found in the oceans around the world, albeit at extremely low concentrations. The uranium ore deposits that are currently being mined have an average grade of around 0.10% of uranium. In 1960s when uranium mining was first started, this grade would have been considered adequate, however today some uranium mines in Canada contain vast amounts of uranium ore that can have grades up to 20%. Some mines can be successfully run with ore grades of 0.02% uranium. In 2019, uranium mines were operated in 20 different countries. It should be noted that around 50% of all uranium produced worldwide comes from only nine mines spread across four different countries. (World Nuclear Association, 2021) In some cases, uranium is produced as a by-product when mining for other minerals. One of these types of operations is the copper mine Olympic Dam in Australia. Uranium can also be acquired as by-product when treating other ores such as gold or phosphate. In cases where uranium is produced as by-product the grade of uranium usually is quite low, with grades that may not even be a tenth of what the grade

37 would be in mines that mine uranium exclusively. In cases like these, uranium can be mined at competitive costs when compared to cases where only uranium orebodies with a low grade was to be mined. The World Nuclear Association (2021) defines an orebody as a mineral deposit from which minerals can be acquired from and that it is economically viable to do so at the current market conditions. They also state that mining uranium does not differ from mining other minerals, unless the grade of uranium is high. In cases where the grade is high, special care must be taken to ensure that the health of the mine workers can be protected, and radiation exposure is kept as low as possible. Measures that should be taken are dust suppression and in some extreme cases remote handling techniques. Surveying for uranium deposits is easier when compared to other minerals as the radiation signatures of uranium decay products can be measured from the air. (World Nuclear Association, 2021) OPEN PIT MINING An open pit mine is viable in cases where access to the ore can be achieved by removing barren or waste rock. In some rarer cases, open pit mining can be done if the ore deposit is at ground surface. In order to access the open mine, one or more ramps are built on the sides of the mine. The ramp design that is chosen depends on the geotechnical strength of the soil, sediment or the rock material. The steeper the ramp can be made, the more economically viable it is. The design must consider the possibility of wall failure, so most open pits usually choose to build more ramps into the mine. Another safety aspect that should be considered is flooding. In some cases, control of groundwater inflows may be necessary if the risk of flooding exists. In general, uranium grade and the deposit size are the factors that affects how large the open pit will become. High grade uranium deposits or large quantities of ore, usually means that the size of the open pit can become a kilometer wide and hundreds of meters deep. These factors also take part in determining the optimal size of the mining, rock and ore hauling equipment. Other factors that affect the size of equipment is the expected mine life length and the distance that the ore must be transported in order to be treated. (Hore-Lacy, 2016) According to Hore-Lacy (2016) the mining methods that are used in mines in general are decided by the rock properties. In cases where the rocks in the mine have low geotechnical strength, in other words are soft, the rock material can be easily

38 removed by using scraper technology. This means that the material can be removed by excavating equipment, after which the material is transported out of the mine and dumped. During excavation only a thin layer of material is removed at a time. Another possible mining method for soft rocks is to use bulldozer to loosen and move the material in thin layers and piled into heaps. Bulldozer are usually used if the rocks are too strong for excavators. The heaps can then be transported from the mine by loading it into trucks or by using conveyor belts. The material can also be mixed with water to form a slurry which can be pumped out of the pit. (Hore-Lacy, 2016) If the rocks geotechnical strength is too high and using bulldozers or excavators is not possible, the drill-and-blast method is generally used. As the name itself implies, in this method holes are drilled into the ground. The holes drilled into the ground are done in predetermined patterns and spacing. Explosives are then loaded into the holes and detonated in a specific order that is designed so that the rock material fragments in the most efficient way. After detonation, the rock material can be excavated and loaded into trucks or onto conveyor belts, and removed from the open pit. (Hore-Lacy, 2016) UNDERGROUND MINING Uranium ore can also be found deep underground or at depths that open pit mining would not be a viable option. In cases like these, underground mining is the usual mining method with which the ore is mined. When choosing the design of the underground mine, many different factors are considered, including the topography of the ground surface, the depth of the orebody, rock strength condition as well as road access needs. The most common access methods that are used for underground mines are shafts, adits and declines. (Hore-Lacy, 2016) Shafts are vertical openings that are built into the underground mines. Shafts are usually only a few meters wide and they are usually used as both entrances into mines and to connect different levels in mines together. The length of a shaft can vary from tens of meters to over a thousand meters according to Hore-Lacy (2016). In order to strengthen the shafts and to minimize the risk of collapse, shafts are usually lined with concrete. In cases where the rock material in which the shaft is

39 built is strong, the shafts may be left unlined. Often the rock strength changes at different depths making a combination of both lined and unlined shaft sections common. In some cases, groundwater control may be necessary, and lining or grouting must be used in order to makes sure that the risk of flooding can be minimized or eliminated. Personal and materials are transported in larger containers or cages with the help of a winch beneath the headframe. (Hore-Lacy, 2016) Access to orebodies can also done by using adits. Adits are horizontal openings that have a small inclination and can be used in cases where the orebody is at the same elevation as the ground. Like shafts, adits are commonly only a few meters wide. If adits are used to gain access to the mine, transportation for personal and materials can be done by using a railway, and in some cases by using wheeled vehicles. Conveyor belts can be used for the transportation of ore and waste rock. Lining of adits can be done in a similar way to shafts, with the difference that steel and timber may also be used instead of concrete. An adit can also function as drainage for groundwater. (Hore-Lacy, 2016) The final way that may be used to gain access to orebodies in underground mining is the use of declines, which are inclined ramps that are spiral formed. This allows vehicles to be used for transportation of ore and waste rock. Hore-Lacy (2016) points out that widths of the declines are usually built according the width of the widest vehicle that is used in the underground mine. Passing spaces must be built at regular distances to allow the mining vehicles to pass each other, in order to make two-way traffic possible. As the cases was for the two earlier access methods, lining for declines should also be done if a risk for collapse exists. In cases where only a single access point exists into the mine, measures should be taken to allow the miners to escape the mine in cases where the access point might be blocked. This is done by building emergency access routes into ventilation shafts. Multiple access points are preferred in most mines, as this allows for more efficient mining operations, as well as enhanced safety for the mining personal. (Hore-Lacy, 2016) According to Hore-Lacy (2016) the most common methods for mining underground narrow ore veins is using the drill and blast method. Depending on the location of the orebody, this type of mining may be done by starting the blasting process from the bottom of the orebody and moving upwards. It is also possible to start from the top of

40 the orebody and continue towards the bottom. A combination can also be used for higher efficiency. The empty space created after this mining method can be handled in different ways. In some cases, the empty void is allowed to collapse by itself and in some cases, waste rocks can be used to fill up the space. (Hore-Lacy, 2016) In cases where the orebody is one the larger side, the preferred method for mining is stoping. Stoping is used if the orebodies can be mined in a series of vertical masses. The size of these masses can reach tens of meters wide and high. In this mining technique, access tunnels are dug towards the orebodies. Once the orebody has been reached, holes are drilled into the deposit and explosives are placed and detonated. After detonation, the now loose ore and rocks are removed from the blast site. After depleting the orebodies, the empty space can be filled with waste rocks mixed with cement. Once the mixture has hardened and deemed to be strong enough, the operation is continued by moving onto the next orebody. (Hore-Lacy, 2016) If the orebody is extensive and horizontally inclined, the room and pillar mining method can be used. This mining technique is done by removing ore in galleries and leaving pillars of orebodies at regular intervals to act as support structures. In this method the mine forms a perforated shape. After an area has been mined it is backfilled to allow the miners to remove the orebody pillars that were left as support. Another method that can be used is to remove the pillars after the orebody has been depleted and in doing so the empty pocket can be collapsed and filled easily. (Hore- Lacy, 2016) IN SITU LEACH(ISL) MINING The third mining type that is becoming more and more popular, having passed open pit mining and underground mining according to Crossland (2012), is in situ leach mining. This method is called in situ recovery mining in North America. The basic principle of ISL is dissolving orebodies that are in porous unconsolidated material, such as sand and gravel. This mining method can be used if the orebody s water body is lying vertically and ideally horizontally. This method cannot be used in cases where the waterbody near the orebody is used as a water supply for households. Compared to earlier mentioned mine types, ISL is the method that has the smallest impact on the environment. (Crossland, 2012)

41 In ISL, weakly acidified or alkaline groundwater containing vast amounts of oxygen is flown through the enclosed underground waterbody containing uranium in loose sands. The acidified or alkaline groundwater dissolves the uranium from the sand and the solution is pumped up to the treatment plant. In the treatment plant, the uranium is treated and recovered as a precipitate. Uranium production in the USA and Kazakhstan use this method. In Australia, hydrogen peroxide is used as the oxidant and sulfuric acid is used as the complexing agent to give a uranyl sulfate. In Kazakhstan, mines prefer to use higher acid concentrations instead of using an oxidant. The mines mining with ISL in USA use an alkali leach in order to get uranyl carbonate. An alkali leach is used due to the acid-consuming minerals that are present in the local waterbodies. Waterbodies that have more than a few percent carbonate minerals require alkali leach in order to function properly. (Crossland, 2012) In order to pump the fortified groundwater into aquifer a series of injection wells are used. The fortified groundwater flows slowly through the waterbody leaching the uranium from the sand. This mixture is pumped up with the help of the pumps that are in the extraction wells that have been placed at strategical distance according to Crossland (2012). The acid consumption in ISL can vary quite drastically depending on where the mining taking place. Factors that influence the acid amounts used are geological conditions, as well as the operating philosophy in the country that the mining is taking place in. A good example of this is the vast difference in the acid kilograms used to leach one kilogram of uranium in Australia and Kazakhstan. Crossland (2012) claims that the usually figure for acid use in Kazakhstan is around 80 kg of acid per kilogram of uranium, while in Australia this figure is reported to be around 3 kg/kgu. Compared to open mining and underground mining even small orebodies can be exploited quite economically efficiently by using ISL mining. It is quite common that satellite plants are set up near small orebodies. These satellite plants are used to load the ion exchange resin/polymer and transport it to the main plant for stripping. (Crossland, 2012) 2.5 URANIUM PROCESSING

42 This section will look at the process that uranium ore that is mined goes through in order to become nuclear fuel. The different steps of the process that can be seen in figure 6. Figure 6 Uranium processing chain (Hore-Lacy, 2016) MILLING AND CRUSHING Uranium ore goes through milling and extraction in order to extract the uranium from the uranium ore and remove to unnecessary coarse material that may be contained in the ore. Milling and extraction contains processes that are both physical and chemical. Generally, the first step for uranium extraction is crushing and grinding. After going through crushing, the uranium ore is broken down into fragments that are a couple of centimeters in size according to Hore-Lacy (2016). Crushing is a dry mechanical process and in order to minimize the dust that is released during this process, a water spray can be used. This is done to protect the health of the workers and ensure that visibility can be maintained. The next step after crushing is grinding. By going through grinding, the crushed uranium ore is reduced into sand or finer sized

43 particles. Uranium obtained through alkaline leaching projects generally need finer grinding compared to acid leaching. Grinding is done in large rotating drums and these drums can contain only the ore itself and steel rods, steel balls or even rock pebbles may be added to enhance the process. This process is also mechanical and, unlike crushing, is typically a wet process. Hore-Lacy (2016) points out that the size of the crushing and grinding plants are decided based on the type of ore that is processed in the plants. Factors that affect plant size are, but not limited to, hardness of the material, size of the run-of-mine ore, clay content as well as the amount of ore that is wished to be processed in the plant. It should be noted that crushing is not always necessary or even effective for certain types of ore. In cases where the ore is fine enough without crushing, the ore can instead be grinded immediately. (Hore- Lacy, 2016) ORE BENEFICIATION The term beneficiation refers to the process in which the crushed and grinded uranium is checked and high-grade uranium ore is collected for further treatment, while the uranium ore which grade is deemed to low is rejected and sent for further treatment. The collected higher-grade ore can be treated in smaller-scale processing plants and different technology can also be used when compared to the run-of-mine ore. Beneficiation is only viable if most of the mass is rejected and the greater part of uranium is secured. When uranium mining was first started, this process was done by hand and based on the visual appearance of the uranium ore or by measuring the gamma radiation that the ore radiated. Nowadays mechanical sorting is practiced, and the sorting process takes physical, mineralogical or radiometric characteristics into account when sorting the ores. (Hore-Lacy, 2016) Hore-Lacy (2016) mentions that in Canada beneficiation was carried out by feeding the uranium ore into a gravity concentration plant. The plant uses jigs and vibrating tables to sort the ore and was shown to retain the uranium ore that had a grade over 30%. Another separation method that can be utilized is flotation, in which air bubbles are blown into a mixture of water, chemical additives and minerals, causing certain types of minerals to float up with froth that is created by the air. This method is not widely used in mines focusing on only uranium mining but is used in mines that mine for copper-uranium-gold. In cases where uranium is considered a side product

44 in mines and flotation is used in order to separate minerals from each other, uranium is recovered from the reject stream of the flotation circuit. In Eastern bloc countries automated radiometric ore separation was widely used. It is still used in some of these countries, but due to the separation method having some problems with radiation, this method has fallen out of fashion. (Hore-Lacy, 2016) URANIUM EXTRACTION AND PURIFICATION The next step in the uranium refining chain is the extraction of uranium. According to Hore-Lacy (2016) uranium has four oxidation states and two of these states namely U4+(tetravalent) and U6+(hexavalent) are of importance geochemically and mineralogically. Tetravalent uranium can usually be found on the subsurface, while hexavalent uranium can be found under oxidizing conditions near the surface. Out of these two oxidation states, hexavalent uranium is the more soluble one. Hexavalent uranium can easily be dissolved with dilute acid and carbonate solutions, while tetravalent uranium is insoluble. The basic purpose of extraction is to convert tetravalent uranium into hexavalent uranium with the help of oxidation. Generally, uranium oxides and silicates can be easily dissolved in oxidizing acid or alkaline solutions. Out of these two options, oxidizing acid is the preferred method, as the dissolving process that uses alkaline solutions is slower, requires more grinding and in some cases can be less complete when compared to oxidizing acid. An exception worthy of mentioning is if the uranium ore has carbonate content of more than 1-3%, in which case using acid is not recommended due to high acid consumption of the carbonate and the use of alkaline leaching is the more efficient method. Alkaline leaching may however require elevated temperature and pressure to function properly. (Hore-Lacy, 2016) According to Hore-Lacy (2016) acidic leaching is usually carried out by using sulfuric acid, due to its low price and availability in industrial grade. Other acid options also exist, such as nitric and hydrochloric acids, but due to the higher prices, as well as the risk to causes greater damage to the environment and the increased chloride levels from hydrochloric acid that may cause problems later in the processing chain, sulfuric acid is to be preferred. The amount of sulfuric acid that is used in this process can vary drastically from a few kilograms per ton ore to tens of kilograms of acid per kilogram ore. If the uranium ore contains brannerite, which is a

45 uranium calcium titanium iron oxide, or uranium-bearing zircon or alkaline minerals, such as carbonate as mentioned earlier, higher sulfuric acid amounts are needed. Brannerite is a tricky mineral as it is very resistant and can still contain uranium after treatment especially if the plant focuses more on the economic efficiency than trying to recover all the uranium. In order to produce a high yield of uranium maintaining a suitable oxidizing environment during acid leaching is critical. In order to maintain the environment in an oxidizing state, different oxidizing agents can be added such as pyrolusite, oxygen, Caro s acid and many more. If no free oxygen gen is available in the process that will take place in the uranium ore is the reducing of the ferric ion into the ferrous ion. (Hore-Lacy, 2016) When the uranium is in an aqueous solution the normal procedure is to do additional purification and concentrate the uranium before it is packaged. As uranium is typically not present as a metal ion in the uranium ore, but as a part of an anionic complex, the uranium must be separated from this complex. According to Hore-Lacy (2016) the two preferred methods to remove uranium from the loaded solution are solvent exchange and ion exchange. Solvent exchange is a technology that is used for separation, purification and recovery. In solvent exchange the different solubility of the compounds is taken advantage of by using two different immiscible liquids, these liquids typically being water and an organic solvent like kerosene with added amines. Solvent exchange can only be used in uranium recovery in combination with acid leach solutions. The loaded solution and barren solvent are mixed which results in uranium being transferred into the solvent while leaving the other dissolved metals behind. The mixture is left to settle with the loaded solvent floating above the now barren aqueous solution. The barren aqueous solution can be cleaned and topped up with acid and oxidant and returned to the extraction stage for reuse. The next step in solvent exchange is to mix the loaded solvent with a new aqueous solution containing different compounds and repeat the process mentioned earlier multiple times in order to increase the uranium concentration and remove any unwanted purities. (Hore-Lacy, 2016) Hore-Lacy (2016) states that ion exchange is the exchange of ions from a solution with ions from solid organic resins. Ion exchange is a reversible process, allowing the target ion, in this case the uranium complex, to be recovered and the resin to be

46 regenerated. The process can be used for both acid and alkaline leaching in order to recover the uranium. Uranium is present in a loaded solution in anionic forms, complexed with carbonate and sulfate ions instead of metallic cations. The resins are usually small polymeric beads with a diameter around 0.5 mm. The loaded solution is flown through a tank filled with resin in which the uranium is absorbed onto the resin as an anionic complex. Different variants of ion exchange exist, and one variant that has seen an increase of interest, is the resin-in-pulp technique, in which the finely grounded ore and the loaded solution is mixed into a pulp to which resin is mixed into. The resin is then separated out of the mix with the help of screening. After separating the loaded resin, the next step is to wash the resin with a different aqueous solution. Doing this causes the uranium to leave the resin and enter the new solution in a process called elution. The resin is regenerated with a third aqueous solution and reused to absorb more uranium. The uranium in the eluate will have a much higher concentration and contain fewer impurities when compared to the original loaded solution. (Hore-Lacy, 2016) Uranium can be precipitated from the solution by using different precipitants depending on the solution type. The solution type as well as the precipitant that is used will affect the ph range. Common precipitants that are used are hydrogen peroxide, ammonia, magnesia, magnesium hydroxide and sodium hydroxide. Factors that are considered when choosing the precipitant are the purity of the feed solution, reagent cost, product specifications as well as the environmental impact of the chosen reagent. The last-mentioned factor has made ammonia a less popular choice for a precipitant as it can cause severe environmental damage and is tricky to remove from the wastewater stream. (Hore-Lacy, 2016) The final step is to remove the moisture and drying of the uranium precipitate. This process varies among operations, but the uranium precipitate usually first washed with good quality water after which the precipitate is dried. Equipment that is commonly used in this process are horizontal belt filter, batch filter or centrifuges. The final drying equipment and the temperature used also varies from cases to case, with the drying temperatures in larger mines usually reach up to 700 C. The final product that is achieved at high temperatures is usually U3O8. U3O8 has a theoretical uranium content around 85%. In smaller mines the final product can contain

47 peroxide, magnesium, sodium or ammonium in addition to oxygen meaning that the uranium metal content can be much lower when compared to the final products that have been roasted at high temperatures. The final product is often called yellowcake, mostly due to the possible yellow color, although the correct technical term for the product is uranium ore concentrate or UOC. The final product is packaged into drums, volume depending on the customers desires. The drums usually have a volume of 400 liters and can weigh up to 300 kg. The drums are marked clearly and can be transported by road, rail and sea. The drums that are being transported must be accounted for and a strict record must be kept in order to comply with national and international safeguard requirements. (Hore-Lacy, 2016) CONVERSION According to Crossland (2012), uranium conversion is the process in which pure uranium hexafluoride (UF6) is manufactured from the yellowcake from the mining and refining process. UF6 is needed in the enrichment process as uranium in metal and compound form have very high boiling points and as such are very difficult to transform into gas phase. The exception to this is UF6 that is a volatile solid at room temperature and has a vapor pressure of 10.6 kpa at 20 C and a sublimation point of 56 C. The critical point for UF6 occurs at 64 C and kpa. The different states for UF6 can be seen in figure 7. Figure 7 Uranium hexafluoride phase diagram (Crossland, 2012)

48 The high vapor pressure allows for UF6 to be handled as a gas in low pressure and room temperature or at a slightly elevated temperature. Crossland (2012) mentions that there are five commercial conversion services operated. These services provided by Rosatom/JSC TVEL (Russia), Honeywell/Converdyn (USA), AREVANC/Comurhex (France), Cameco (Canada) and Westinghouse/Springfields Fuels Limited (SFL, UK). (Crossland, 2012) Every step that takes place during conversion can be seen in figure 8 and the most important steps of this process will be looked at in this section. Figure 8 Uranium hexafluoride conversion process (Crossland, 2012) The first step in the conversion process is dissolving the yellowcake in concentrated nitric acid in order to form uranyl nitrate. After this the uranyl nitrate is mixed with a

49 solution consisting of tributylphosphate, or TBP for short, and a hydrocarbon dilute, in most cases kerosene, in order to extract uranium into the solvent phase. Next the uranium is separated from this aqueous phase, leaving impurities behind, and washed out from the solvent using fresh dilute nitric acid or water in order to get a solution of purified uranyl nitrate. The kerosene used in this process does not have a chemical role and is used exclusively to lower the density of the solvent in order to make it easier to separate the aqueous phase and solvent phases. The solvent can be recycled after going through an alkaline wash as the solvent is not consumed. (Crossland, 2012) After solvent extraction, the uranyl nitrate is boiled down in order to increase the concentration of uranium in the solution to around kg/m 3. After boiling, this concentrated solution is fed into a high temperature denitration unit in order to get rid of the water and decompose the nitrate in the solution in order to get a purified uranium trioxide (UO3) product. The equipment for this step of the process can vary from the use of pot denitrators to fluidized bed reactors. The next step of the conversion of natural uranium into UF6 is to turn the pure UO3 into the intermediary product, uranium tetrafluoride (UF4). The initial reaction in order to produce UF4 is to add hydrogen and reduce UO3 into uranium dioxide (UO2) at high temperature. Depending on the facility in which the conversion is carried out at, different equipment is used. Cameco does this step by using a fluidized bed reactor, while SFL uses a rotary kiln and a furnace is used by AREVA. Crossland (2012) does point out that equipment that is used in this step does not actually affect how the reaction is done. (Crossland, 2012) Crossland (2012) mentions that the next step in conversion is to turn UO2 into uranium tetrafluoride. This can be done in different ways depending on the equipment that is used. If a rotary kiln is used, the common method to produce UF4 is to make the uranium dioxide react with hydrogen fluoride gas at a high temperature in order to produce UF4. Another option is to make UO2 react with a hydrofluoric acid at a temperature of 100 C, after which the solution is dried and calcined in order to remove water of crystallization before additional fluorination. (Crossland, 2012)

50 The final step in converting yellowcake into uranium hexafluoride is to convert UF4 into uranium hexafluoride. The process is quite simple according to Crossland (2012) as UF4 is made to react with fluorine gas at a high temperature in order to produce UF6. The only tricky part in the final step is the production of the fluorine gas that is needed in the last step. Fluorine gas is an extremely reactive gas and is usually produced at the conversion facility itself. This is done by using electrochemical cells with graphite anodes. The cells themselves contain molten potassium bifluoride salt as the electrolyte that is fed anhydrous hydrogen fluoride gas at a constant flow. The hydrogen fluoride that is fed into the electrochemical cell are split into hydrogen and fluorine component elements and the fluorine component is fed into the UF6 production reactor. Depending on the type of reactor that is used to carry out the last step in this process the temperatures can vary quite drastically. The temperature in flame reactors usually reach C, while much lower temperature (450 C) are used in fluidized bed reactors. Once the reaction has been completed and UF6 has been produced, the UF6 gas is collected and condensed into a solid state. The solid UF6 is then heated until it reaches liquid form in order to drive out any remaining gas impurities, making it possible to sample the liquid and provides a form that is easy to dispense into transport containers. The liquid UF6 is then dispensed into containers for transport and left to cool and solidify for around five days, after which the uranium hexafluoride is ready to be transported to the enrichment facility. (Crossland, 2012) Crossland (2012) mentions that other ways to produce UF6 also exist, but the writer of this thesis has chosen not go through these techniques any more than to mention that the process are different in the way that uranium dioxide is produced in the beginning of the process after which the process are generally the same as described earlier. The exception to this being the conversion process that is used in Russia, in which the yellowcake is made to react with fluorine gas in a flame reactor from the start in order to produce UF6. (Crossland, 2012) ENRICHMENT According to Hore-Lacy (2016) all the major commercial civil enrichment facilities use some form of gas centrifuge to achieve isotope separation of 235 U to 238 U. In order to use a gas centrifuge, the uranium that is being enriched must be in a gaseous

51 state. The process that is done to turn natural uranium into uranium hexafluoride was described earlier. Isotope separation is a purely physical process and no chemical transformation occurs. The feedstock that is used in separation in gas centrifuges is UF6 and the product as well as the tail also contain UF6. The product can be converted into uranium dioxide later in the nuclear fuel manufacturing process. Ever since uranium has been enriched on an industrial scale, uranium hexafluoride has been used, and this has led to the global industrial infrastructure that supports uranium enrichment to be built to support the process that uses uranium hexafluoride in order to enrich uranium. This is also due to previous enrichment method called gas diffusion that was used earlier but it has mostly been replaced by gas centrifuges in modern times. It is possible that some form of enrichment technology could emerge that could challenge the current norm, however if such as technology did emerge, it would require large investments and changes to the current infrastructure. (Hore- Lacy, 2016) Gas centrifuge technology take advantage of the mass difference of 235 UF6 and 238 UF6 to promote separation. This process is done by feeding UF6 gas into a centrifuge unit that spins at a very high speed. The centrifuge must be vacuumed before feeding the UF6 into the centrifuge. The centrifuge is kept at a constant temperature and protected from impact and other physical interference that might interfere with the process. The wall of the centrifuge acts as the rotor. Crossland (2012) explains that rotation of the gas applies an acceleration to the gas molecules in the direction of the walls of the centrifuges, and due to the higher mass of some of the gas molecules, they are affected by greater forces and makes the heavier gas molecules concentrate at the centrifuge walls while the lighter gas molecules are concentrated closer to the central axis of the centrifuge. The next step in gas centrifuge enrichment process is to use a thermal gradient to make the partial separated gas circulate along the axis of the centrifuge, after which scoops are used to draw off an enriched product stream and another stream for the depleted tails. A simplified gas centrifuge can be seen in figure 9 below.

52 Figure 9 Simplified gas centrifuge (Crossland, 2012) The centrifuge itself is connected to a motor that is used to drive the rotor and in the top part of the centrifuge there are magnetic bearings. The inlet and outlet pipes go through the magnetic bearings. A casing surrounds the centrifuge and the space between the centrifuge and casing is evacuated in order to minimize friction. It should be noted that modern centrifuges can be run for long periods of time without maintenance. The failure rate for a gas centrifuge is less than 1% per year and some centrifuges can run for years without failure. Thanks to the high reliability of centrifuges, they can be kept running 24/7. (Crossland, 2012) According to both Crossland (2012) and Hore-Lacy (2016) the degree of separation that is achieved in a gas centrifuge is small and the process must be repeated hundreds of times in order to achieve the needed the required concentrations that are need to enrich the uranium in order for it to be useable by nuclear power plants. In order to speed this process multiple gas centrifuges are connected in series and parallel in order to reach the desired outputs and reduce the 235 U levels in the tails which will increase the economic viability. The configuration of machines that are

53 used to perform repeated isotope separations is known as a cascade. The steps that are included in a cascade can be seen in figure 10. Figure 10 Casacde process (Hore-Lacy, 2016) Crossland (2012) mentions that commercial scale gas centrifuge plants can possess tens of thousands of gas centrifuges and many cascades in order to do this. (Crossland, 2012; Hore-Lacy, 2016) The tails that is produced during the enrichment process is usually kept by the enricher as the contracts made with customers usually state that the customer provides the enricher with a certain amount of feedstock and will receive a certain amount of enriched product. The degree of 235 U that the tail may contain is usually stated in the contract and tails that fall under that limit is kept by the enricher. As the tail still contains 235 U it is possible to further treat and enrich the tail in order to produce more product by using the tail as feedstock. Different factors will decide if it is economically viable to further enrich the tails. Some of the factors that affect this are the price of natural uranium feedstock, the price of enrichment services as well as

54 the marginal costs of operating an enrichment facility. In cases where the enrichment facilities possess a surplus of enrichment capacity, the tails are usually treated as the enriched tails can be sold on the uranium market. Other options for treating the produced tails is to operate plants in a mode of operating that is known as underfeeding. In this mode, the enricher will use more separate work units than agreed in the customer contract to provide the agreed upon enriched product and strip the tails to a lower assay than agreed and by doing so, less feedstock is used. The unused feedstock can then be sold back to the market. Underfeeding is usually done in times when the demand for primary enrichment is suppressed. In cases where uranium feedstock prices are low, enrichers can choose to overfeed enrichment plants. This mode is the opposite of underfeeding, meaning that the enricher uses less separate work units and more feedstock than agreed upon. This has only been done in cases where the price of natural uranium has been much lower than enrichment services or when enrichment services have been in short supply compared to the demand. (Hore-Lacy, 2016) CONVERSION OF ENRICHED URANIUM HEXAFLUORIDE INTO URANIUM DIOXIDE The next step in the uranium processing chain is to convert the uranium hexafluoride into uranium dioxide. Uranium hexafluoride is received in a solid form at the fuel fabrication facility in 30B transport cylinders. The first step in converting UF6 into UO2 is to heat the transport cylinder until the uranium hexafluoride is vaporized. This is done by using steam in an autoclave, electrically heated hot air chamber or an electrically heated blanket. Once the UF6 is liquified and enough UF6 gas pressure is generated above the liquid, the gaseous UF6 is drained from the cylinder. It is also possible to use vacuum at ambient temperature to drain uranium hexafluoride gas from the cylinder; however, heating is the more common way of doing this step of the process. Many of the processes and chemical reaction that the place when converting the enriched uranium hexafluoride into uranium oxide are similar to those that take place the natural uranium is processed. There are several conversion processes that can be used in order to produce uranium hexafluoride and the most common processes will be explained briefly in this section. (Piro, 2020)

55 The first process that can be used for conversion is the ammonium diuranate route, which is a wet process in which UF6 is hydrolyzed in order to produce a solution of uranyl fluoride (UO2F2) and hydrofluoric acid (HF(aq)). This process can be seen from figure 11. Figure 11 ADU conversion (Piro, 2020) The reaction itself is exothermic and takes place instantly. Uranyl fluoride can by itself be easily dissolved in water, but as the solution that is formed when UF6 is mixed into water contains HF, this will decrease the solubility of uranyl fluoride. The reactors that are used to carry out are designed with a few important features that are needed for this process to be carried out smoothly. It is important that the reactors include at least the following features to operate well: nozzle designs that prevent clogging from occurring by the hydrolysis products, reactor materials that are corrosion resistant, suitable reactor size for the intended volumes as well as the abilities to control the concentration of uranyl fluoride and suitable mixing capabilities. Hore-Lacy (2016) mentions that in the next step ammonia is added to the uranyl fluoride solution in order to produce precipitate of ammonia diuranate (ADU). After this the ADU is filtered, dried and calcined in the presence of steam and reduction is done with the help of hydrogen in order to produce uranium dioxide powder. (Hore-Lacy, 2016; Piro, 2020)

56 The second process that can be in order to get uranium dioxide from the uranium hexafluoride is the ammonium uranyl carbonate route. This process steps can be seen as a whole from figure 12. Figure 12 AUC route (Piro, 2020) The first step in this process is the feeding of UF6 gas into a stirred aqueous system at the same time as gaseous carbon dioxide and ammonia is fed into the reactor. The reaction between these elements form an ammonium uranyl carbonate precipitate with ammonium fluoride (NH4F) in solution. The AUC precipitate is filtered from the mother liquor and washed in order to reduce the fluoride content in the uranium dioxide product. The filter cake is washed with methanol in order to get rid of some of the moisture content. In order to convert the solid AUC into uranium dioxide, thermal decomposition and reduction is done in one or multiple fluidized-bed reactors at temperatures between 520 C and 650 C. This will produce uranium trioxide, which is reduced by using hydrogen in the fluidizing gas stream in order to convert it into uranium dioxide. The fluoride content is also reduced and removed from the reactor as hydrogen fluoride. The uranium dioxide that is produce by using this process will have surface activity and will have to be stabilized in order to keep a relatively constant oxygen/uranium ration once it is exposed to air. (Piro, 2020) The final process that is used in order to turn uranium hexafluoride into uranium dioxide is a dry technique that according to Crossland (2012) is the most environmentally friendly conversion process when compared to earlier two processes. This is due to the process not creating any uranium contaminated waste

57 and the clean water that is left by this process as well as the hydrogen fluoride can easily be disposed. Many different dry technique processes exist however only the integrated dry route will be explained in this section. The IDR was originally developed and used as a commercial process in the UK in The process has later been implemented by other producers as the main benefit of the process is that the whole reaction chain can be done in one reaction vessel as can be seen in figure 13. Figure 13 IDR route (Piro, 2020) In IDR, the vaporized uranium hexafluoride is injected into the reaction chamber that is located in the front of the reaction vessel at temperature ranging between 150 C to 250 C. The UF6 is then contacted with a jet of superheated steam in order to form a product plume of UO2F2 particulate as well as HF in an exothermic reaction that takes place instantly. Pori (2020) explains that the UO2F2 disengages from the HF in the reaction chamber and is fed into the rotary kiln with the help of a screw feeder. Sintered metal filters are used to remove the hydrogen fluoride from the reaction chamber after which it is sent over to a HF recovery process. Once the uranyl fluoride enters the rotary kiln it is converted to uranium dioxide with the help of a countercurrent flow of steam and hydrogen at a temperature interval of C. The mechanism of the reaction is seen continue by separate pyrohydrolysis and reduction steps, however in reality the mechanism can be more complex with different competing reactions. This step in the process is endothermic. (Piro, 2020) FUEL PELLET PRODUCTION

58 The next step in the nuclear fuel production chain is turning the uranium dioxide powder into pellets. In order to convert the uranium dioxide powder into pellets the following steps are done; mixing the UO2 powder with a binding and lubrication agents, compaction or cold pressing, sintering and finally precision grinding. (Hore- Lacy, 2016) The binding and lubricating agents that are added in the first step of this process are typically organic compounds such as aluminum or zinc stearate and stearic acid. By adding these agents, it is possible to enhance pore formation and in doing so increase the density during the sintering process. Hore-Lacy (2016) points out that binding agents also increases the strength of the powder once it is cold pressed and also helps in reducing dust hazards that can be associated with the use of uranium dioxide powder. After these additives have been added to the powder, the next step is to cold press the powder into so called green pellets. The green pellets have a density of around 55 60% of the theoretical density. In order to increase the density of the green pellets as well as to get rid of the additives, sintering is done in a hightemperature furnace. Once sintering is complete, the green pellets will have been converted into a stable ceramic that has the necessary heat transfer properties that are required for nuclear fuel. The final pellet density will be between 96 97% of the theoretical density. After sintering the pellets are grounded down to the final dimension, inspected and stored for fuel rod loading. (Hore-Lacy, 2016) BURNABLE ABSORBERS Burnable absorbers are according to Hore-Lacy (2016) neutron-absorbing materials that included in light water reactor fuel as means of power shaping, local powerpeaking control as well as long-term reactivity control. Burnable absorbers are sometimes called burnable poisons. Burnable absorbers that are used for long-term reactivity control are depleted as the fuel assembly is used in the reactor core and provides more reactivity control when the assembly is taken into use, decreasing over time. Whether burnable absorbers are used or not depends on the fuel vendor as well as the fuel design, but in general absorbers can be grouped into three different groups; discrete absorbers, intimately mix absorbers and surface coating absorbers. (Hore-Lacy, 2016)

59 Discrete absorbers are rarely used today, only found in PWR fuel designs and are usually used in the form of absorber filled rods that are inserted into an empty rod control cluster assembly guide tube. Materials that are used in discrete absorbers are Boron-10 dope Pyrex glass or aluminum oxide-boron carbide pellets. Hore-Lacy (2016) points out that reactivity control is done by varying the absorber that is loaded per rod as well as the number of absorbers rods that are sued per fuel assembly and the total fuel assemblies that contain absorbers. (Hore-Lacy, 2016) The next type of burnable absorbers, intimately mixed absorbers, are made by combining uranium dioxide with gadolinium (GdO2) and erbium (ErO2) before palletization. Intimately mixed absorbers can be used by both PWR and BWR fuel designs and most fuel manufacturers use this type of absorber. It should be pointed out that BWRs exclusively use gadolinium as an absorber, while PWRs use both gadolinium and erbium. According to Hore-Lacy (2016) reactivity control is achieved in a similar way that was mentioned earlier with the added method of varying the distribution of the absorber in the fuel rods. (Hore-Lacy, 2016) The final group of burnable absorbers is the surface coating absorbers. Surface coating absorbers are applied as thin coating on every fuel pellet in the fuel rod. The coatings that are used for this are boron compounds such as zirconium diboride (ZrB2). These coated pellets can then be distributed in different ways in the fuels rods in order to achieve specific power-peaking control. (Hore-Lacy, 2016) FUEL ROD FABRICATION The specifics of fuel rods vary between fabricators and fuel type, in general a light water reactor fuel rod is composed of a zirconium alloy cladding tube, a column of uranium oxide pellets, two end plugs as well as an internal plenum spring. Each manufactured fuel rod possesses a unique identification number that makes it possible to trace the history of every rod. According to Hore-Lacy (2016) the plenum springs task is to prevent pellet movement inside the fuel rod and to protect the rod from any possible damage that might occur during handling operations. (Hore-Lacy, 2016) The manufacturing process usually consists of the following steps that will be explained briefly in this section. In the first step the bottom end plug is inserted into

60 the cladding tube. Once the bottom plug is in the correct position, it is welded in place. The fuel pellets are loaded into the cladding tube as a column. This is done either by pushing the pellets into the cladding tube horizontally or gravity loaded at an angle. The method with which the fuel pellets are loaded depends on the fabricator. Once the pellets have been loaded into the tube, the length of the pellet column is measured and verified that it is the correct length and follows the manufacturing specification. The plenum spring is inserted and placed at the top of the pellet column, once the tube contains the correct number of pellets and the length of the column fulfills the specifications. Once the spring is in place, the top of the cladding tube is placed into a sealed chamber for pressurization. Hore-Lacy (2016) points out that helium is added into the fuel rod until the specified internal pressure is reached. The end plug can be put into place and welded once the pressure is correct, however some fabricators weld the end plug into place before pressurization, in which cases the end plug includes a hole that is used to achieve pressurization that must be welded shut at this point. At this point the fuel rod is complete and different inspections are carried out in order to verify that the quality of the fuel rod is up to standards and no manufacturing errors have slipped by into the final product. The inspections that are done are non-destructive techniques such as visual, x-ray, and ultrasonic inspections. (Hore-Lacy, 2016) SPENT FUEL STORAGE AND DISPOSAL Once the nuclear fuel has been spent in the reactor it needs to be disposed or stored properly. According to The Radiation Act (592/1991) the waste generated in the use of radiation and other radiation practices shall be kept to the minimum which is reasonably practicable without compromising the general principles of radiation protection. According to the report by the Ministry of Employment and the Economy, Energy Department (2016) the objective of final disposal is to process spent fuel in a way so that it will not cause damage or waste management obligations in the future. This should be done in a way in which future generations would not be required to know about the existence or location of the disposal facilities. In order to minimize the risk for accidental discover or misuse of nuclear waste, the repositories should be placed deep in bedrock and records should be kept of where and how the

61 canisters containing the waste are stored for future generations. (Ministry of Employment and the Economy, Energy Department, 2016) In Finland, spent nuclear fuel assemblies are cooled down for a few years in water pools in the reactor hall or fuel building. Once the assemblies have cooled down, they transferred in a transport cask to the on-site interim storage water-filled pools, in which they will be kept under water for around 40 years. After this time the radioactivity levels as well as heat generation have decreased to levels that are acceptable for final storage. Finland s disposal is to pack the fuel assemblies into tight metal containers that have a cast-iron insert and a copper outer casing in order to make them as corrosion resistant as possible. These containers are then transported to the final disposal site at Olkiluoto island. (Ministry of Employment and the Economy, Energy Department, 2016) 2.6 SMALL MODULAR REACTORS The interest in modular low-power reactors has risen in the world in recent years. According to the International Atomic Energy Agency, small modular reactors are classified to have an electric capacity of up to 300 MW, and the components and systems can be shop fabricated. The modules can be transported to the sites for installation once a demand arises. (Khan & Nakhabov, 2020) According to Carelli & Ingersoll (2014), no definitive power rating exists that decides if a reactor is small or not, however, the interval of MWe is generally used. This range includes a minimum amount to ensure that the reactors deliver enough power for them to be suitable for practical industrial application. The modules include the NSSS which, when coupled with a power conversion system like a turbine and generator or a process heat system, can be used to produce energy in a form that is needed. As mentioned in the definition by IAEA, these modules are designed and manufactured at a factory and transported to the installation site. The whole unit can consist of one or several modules, which can be built at the same time or installed later, once there is an increased need in power generation. As the modules are built in a factory and not on the installation site, costs can be kept lower as compared to conventional nuclear power plant construction. The reactor term in

62 this case refers to the nuclear reactor in which a controlled nuclear fission reaction is carried out in order to generate energy. (Carelli & Ingersoll, 2020) Currently, various countries have started projects in order to develop and build SMRs. According to Budnitz, Rogner & Shihab-Eldin (2018), nearly 40 different SMR design concepts are being developed worldwide, yet none of these are for sale currently. Only a few of these reactors are under construction now. Most of the reactors under construction are PWRs. (Budnitz, Rogner & Shihab-Eldin, 2018) ADVANTAGES OF SMRs Small modular reactors have many advantages when compared to conventional nuclear reactors. Budnitz, Rogner & Shihab-Eldin (2018), claim that the initial investment cost of building SMRs when compared to large reactors will naturally be lower and, thus, present much lower investment risks to project sponsors. They also mention that even though the initial investment cost will be lower, the specific investment costs per kwe installed are expected to be much higher than large reactors at the beginning of SMR building. This will most likely lead to higher costs per generated kilowatt electricity when compared to current NPPs, with some estimations reaching up to 50%, or in worst cases, even more cost per kwe. Other experts claim that the costs of kwe will be lower when compared to new NPPs, but most experts agree that the disadvantages that may occur at the beginning of SMR construction will over time turn into an advantage for SMR, as the process is developed and cost can be lowered over time. Budnitz, Rogner & Shihab-Eldin (2018), also observe that some economics of SMRs, such as capital costs, operating, maintenance and fuel cost, are still unknown. Other factors that might be in favor of SMRs are the cost reductions that will come from standardization, shop fabrication as well as the reduced levels of complexity of SMR designs that can be associated with the modular nature of the reactor types. Modularity also allows for improved quality control during construction, shorter construction time which will decrease the risks of cost escalation and the cost reductions that occur during learning by doing. Learning while doing means that while SMRs or any other technology is being built, the workers involved in the process will learn more about the technology as they gain experience during the work process. The increased experience can and will, in most cases, lead to performance improvements of a new technology in both production

63 and deployment. Learning while doing will also lead to cost reductions, as the manufacturing process will improve over time and the construction times will become shorter. The fabrication of SMRs also means that the risks for delays in the construction schedule are less likely to occur when compared to conventional nuclear power plant construction sites, as the process is standardized. (Budnitz, Rogner & Shihab-Eldin, 2018) Another major benefit of SMRs when compared to large reactors is grid compatibility. Customers who might be interested in nuclear power plants usually have constraints regarding the size and need of the power capacity additions made, and these constraints can make building conventional nuclear power plants problematic, as the power capacity is too large. The power capacity of SMRs is, in many cases, closer to the need and the mentioned lower prices of constructing SMRs make it a more attractive option. The shorter construction times are also a major benefit as well as the possibility to use the SMR power plants for other purposes than electricity generation also favors SMRs. The different uses for SMRs will be discussed later in this thesis. (Carelli & Ingersoll, 2020) SMR CHALLENGES According to Ingersoll (2016), SMRs also face many challenges in addition to all the advantages that the reactor type has. Ingersoll (2016) mentions that the first and biggest challenge that SMRs must overcome, is for the SMR designers to finish their designs. Licensing does not require the designs to be complete before a license can be given, however, as the nuclear industry has learned, a complete design will help avoid construction delays as well as huge cost overruns. A perfect example of this is the Olkiluoto nuclear power plant that is being built in Finland, which has seen multiple delays and the budget has been exceeded by billions of euros. Ingersoll (2016) notes that for SMRs to succeed, it is important that the reactor vendors do not give in under pressure from investors and customers and introduce new technologies that have not matured and been tested thoroughly. Some experts believe that SMRs have high standards that it must reach and, hopefully surpass, if the technology is to compete with and possibly replace LRs. (Ingersoll, 2016) Ingersoll (2016) stresses that experience with LWR technology has been gained over

64 the years and mentions that according to an assessment conducted by Oak Ridge National Laboratory, LWRs have been in operation for over 20,000 reactor years. When compared to other types of nuclear reactors the operational time for LWRs is by far the longest, as the second longest operational time for a reactor is the gascooled type and this reactor type has only run for 2,000 reactor years. This can lead to many different challenges when designing and engineering non-lwrs, as experience with some of these types of reactors is still low. Depending on the coolant that is used in a reactor, the operational temperatures of the reactor can vary quite drastically compared to LWRs, and because of this, the materials used in the reactor must be chosen carefully in order to ensure that they are suitable for the high temperatures. Another factor that designers and engineers must take into account is the type of nuclear fuel that is used in the reactor. Different coolants can require different types of nuclear fuel, and these fuel types can utilize high- or low-energy fission reactions which, in turn, impacts the rate and type of radiation damage that it inflicts on the surrounding materials. Fuel cycle length also brings challenges, as some SMR designs claim that the cycle can be times longer when compared to current LWR fuels. Designers must make sure that the protective cladding, and the fuel material are able to withstand the conditions in the reactor and maintain their integrity for a much longer time. (Ingersoll, 2016) Designers of light water SMRs choose to use as much of regular NPP components as possible to keep costs down, as well as to hasten the development and licensing processes. Using old and known parts in the design also helps minimize testing and keep the qualification costs down when compared to a first-of-a-kind component. It should be noted that it is not always possible to use old parts and sometimes it is necessary to design new components. As most of SMR designs utilize ipwr technology, Ingersoll (2016) points out that some new technological development is needed for parts that are unconventional such as helical coil steam generators, internal control rod drive mechanisms and internal coolant pumps. As these parts are completely new, they need to go through thorough testing and be qualified for the operational environment in order to fulfill the needs and demands of regulators as well as customers. Ingersoll (2016) also mentions that, in addition to these parts, sensors, instrumentation and control systems may need to be developed especially

65 for SMRs. Due to the lack of external primary coolant loops in integral SMRs, new methods must be developed in order to carry out in-vessel measurements. For SMRs to be used remotely in remote locations, implementation of additional sensors and instrumentation is necessary to allow the reactor operators to monitor plant health, run diagnostics and prognostics. Cogeneration may also need new control systems to be developed in order to properly control the multiproduct load balancing. (Ingersoll, 2016) Ingersoll (2016) also notes that computational analysis methods need to be developed and validated in order to enhance the methods with which safety and operational performance of reactor components and systems can be predicted. Codes for analysis exist, especially for LWR-based designs; however, the differences in system configuration can create new data and validation needs. As SMRs use natural circulation flow for the primary coolant during normal operational conditions, thorough validation of the methods that are used to simulate and predict thermalhydraulic performance during all operational conditions is needed. (Ingersoll, 2016) Another computational method that must be modified and further developed for it to be used reliably with SMRs is called probabilistic risk assessment. PRA is a method for analyzing the safety and reliability of a nuclear power plant, with a focus on assessing the possibility of a severe accident occurring. According to Ingersoll (2016), PRA has been developed for single reactor units and has not been used for units that include multiple reactors and, thus, does not consider how the reactors might cross-unit interact. The accident at the Fukushima Daiichi plant involved four out of the six reactors cross-unit reacting with each other, with hydrogen generated in one reactor leaking into another reactor and causing an explosion. This accident has given designers valuable information about cross-unit accidents and gave birth to the idea that multimodule plants need multimodule analyses to be developed. It is possible to modify single-unit PRA and adapt it for multi-unit analysis, however experts agree that new analyses should be developed for multimodule plants. (Ingersoll, 2016) The last challenge that Ingersoll (2016) mentions can be counted as both a social and technical challenge. He claims that for SMRs to remain economically successful it is important to optimize the economies of SMRs and to keep the designs as simple as

66 possible. This can be done by staying on guard against creeping complexities when designing the reactors as new challenges rise. (Ingersoll, 2016) SMR SAFETY The primary concern that most common people have about nuclear reactors is the possibility of an accident occurring that will release radiation and the possible damages radiation might cause. The safety of NPPs is improving fast and SMRs are no exception to this. According to Budnitz, Rogner & Shihab-Eldin (2018), the primary reason for why SMRs being less prone to accidents compared to LRs is that the thermal energy production per reactor is lower compared to their larger counterparts during both operation and during shutdown. This makes cooling a reactor much easier during an accident or during a regular shutdown. Since cooling is easier, the risk of a core meltdown, as well as the possibility of radioactive material release, is drastically reduced. In some SMR designs, passive approaches are used to remove decay heat, extending the time before overheating becomes an issue, and increasing the stability and safety of said reactor designs. This section will focus on the safety features of water-based SMRs. (Budnitz, Rogner & Shihab-Eldin, 2018) Active safety systems require some sort of external power, force, action or signal in order to work, according to Carelli & Ingersoll (2014). Systems that can be put into this category are pumps and motor-driven or manually operated valves. As these safety features require external power, they are vulnerable and if they are damaged in a natural disaster or terrorist attack, they can make the NPP vulnerable to accidents. Passive safety systems, in contrast, work based on the laws and forces of nature and require no external power, making them less vulnerable to external impacts. It should be noted that building a reactor that only uses passive safety systems is difficult and somewhat risky, even though the risk of the system s failing is significantly lower compared to active systems. Passive systems can be destroyed in accidents. Carelli & Ingersoll (2014) mention examples, such as pipes in the natural circulation loop which can be crushed during an earthquake or a wall that is used as a safety separation barrier, can collapse. For this reason, most modern nuclear power plants use both active and passive safety systems in order to make the plants as safe as possible. Another approach that is used is safety by design. Safety by design is achieved by making design and engineering choices that can stop certain accidents

67 and types of accidents from occurring. Doing this does not only remove the risk of said accidents from occurring but also removes the need of the safety systems that are in place to counter those accidents. Design choices that can be made include avoiding large primary piping and instead using multiple small ones, in order to avoid loss-of-coolant accidents from occurring, if the piping ruptures or is otherwise damaged. Another example is to use shaftless coolant pumps in order to avoid shaft seizure or shaft break, thus reducing the risk of core meltdown. (Carelli & Ingersoll, 2020) In the case of SMRs, many of the PWR designs obtain their inherent safety features from their integral design. Due to the high pressure that is used in PWR, the reactor type is sensitive to leaks or breaches of the primary boundary. However, ipwrs are not as vulnerable to leaks as the integral design makes external piping redundant which, in turn, makes the risk of leaks much lower. The integral design of SMRs also reduces the risk of LOCA from occurring. The primary coolant is also contained in the reactor vessel and as the steam generators and primary heat exchangers are also internal in SMRs, the risk of steam pipe ruptures is lower as the pipes are not external and not as vulnerable to external factors that might damage them. Since the reactor type uses natural circulation in order to remove heat in both normal and offnormal condition, the need for auxiliary pumps and external power is eliminated as well as the risk of either failing. Carelli & Ingersoll (2014) also stress that due to the longer core life of SMRs, the chance of refueling accidents is reduced, as refueling is done rarely. Since SMRs also contain less fuel compared to LRs, a debate regarding whatever the EPZ could be reduced for SMRs has started. Reducing or removing the EPZ would mean that SMRs could be placed near cities, or even in cities, making the district heating with SMRs more viable. (Carelli & Ingersoll, 2020) According to Liu & Fan (2014), SMRs have an increased relative coolant inventory compared to LRs. This means that the water volume per power unit is much higher and, in the case of an accident in which the flow of coolant is stopped or slowed down, the reactor operators have more time to react and get the situation under control. SMRs also have an increased passive cooling capability, since the vessel s height-to-diameter ratio is 2 3 times larger than an LRs. This factor also means that the gravity-driven natural convection circulation capability is increased and, in some

68 reactor, designs the natural circulation can be enough to cool the core at full power operation. Liu & Fan (2014) mention that as SMRs are smaller in size, the economic viability for building the plant underground is increased. Building an SMR underground will protect it from outside attacks and certain accidents from occurring. Underground construction also helps to protect the reactor from seismic activity and reduces the risk of the fission product being released during an accident. (Liu & Fan, 2014) According to the International Atomic Energy Agency (2016), in some ipwr SMR designs the control rod drive mechanism is placed inside the vessel in order to remove the possibility of rod ejection accidents and the LOCA that is tied to this type of accident, as the penetrations in the reactor vessel closure head are eliminated. As mentioned earlier, decay heat is removed passively in some SMR designs. The International Atomic Energy Agency (2016) mentions that this can be done in different ways. One method of removing decay heat is to pair the steam generators with heat exchangers that are immersed in a water pool. In this design, the decay heat is released into the water surrounding the parts. Another design that is used to deal with the decay heat is to have a passively cooled condenser connected to the upper dome of the vessel, creating a natural circulation loop that cools down the primary system. As the valve to the condenser is opened, steam flows through the condenser tubes and, as it is turned into water, it returns to the vessel with the help of gravity. The decay heat can also be removed with the use of a system that utilizes pumps, valves and heat exchangers in a similar way that is used in modern NPPs. (International Atomic Energy Agency, 2016) SMRs also include a safety injection system that will feed borated water into the reactor in case of a LOCA. This can be done in different ways by using gravity, machinery or pressurized tanks. The injection systems are divided into low- and high-pressure injection systems. Low-pressure injection systems feed water into the reactor once the reactor has been depressurized, whereas high-pressure injection systems can do this while there is still pressure in the reactor. If the cooling water in the reactor starts disappearing, the systems will feed additional borated water into the reactor. SMRs also have containment systems that are designed to have three functions. According to the International Atomic Energy Agency (2016), these

69 functions are confinement of radioactive substances in operational conditions or accidents, protecting the plant from natural disasters or human induced events and shielding from radiation in operational states and in accidents. The integrity of the containment must be protected and maintained. This is done by keeping the pressure and temperature inside the containment vessel within and below the design limits. This can be done in different ways. One method is to use a water pool for pressure suppression. In this solution, the high temperature steam that is released from the reactor vessel is led into a suppression pool or tank, in which the steam will condense and, in doing so, the pressure in the containment can be kept from rapidly increasing. Another method to control containment pressure is to spray water into the containment s atmosphere. This is done in most PWR designs where the containment structure is made of concrete. Pumps are used to inject water from the reservoirs, which are located outside the containment, to the sprayer that is in the reactor building dome. The sprayed water condenses the steam and reduces pressure. A more passive method for controlling the containment pressure is to use a large volume metal chamber as a containment unit. This metal chamber is surrounded by a concrete building. The metal chamber is cooled by air flow or by spraying water into it, condensing the steam that might be released during a LOCA on the inside of the chamber. (International Atomic Energy Agency, 2016) The International Atomic Energy Agency (2016) also mentions that SMRs have safety features that are created to deal with severe accident conditions. One of the methods that the agency mentions is to use an in-vessel corium retention feature. This is done by flooding the reactor vessel cavity in a way that the lower part of the vessel is submerged in water during severe accidents. Doing this allows for the integrity of the vessel to be protected and the corium can be retained inside the vessel in the case of an accident. This method is usually combined with either a passive or active containment cooling system. SMRs also contain core catchers that are used to prevent the molten core from escaping the containment building. The core catcher is a specific structure or device that is placed under the reactor vessel and it is made of a heat-resistant concrete ceramic that can withstand the temperatures of the molten core and prevent it from melting through the containment shield. The core catcher is designed to spread the molten core and vessel material in order to decrease the

70 thermal density and includes systems to keep the core stable for extended periods. (International Atomic Energy Agency, 2016) In accidents that cause nuclear fuel damage, the heat released from the fuel will overheat the zirconium alloy in the fuel clad. The overheated zirconium alloy will react with steam at high temperatures and generate hydrogen gas. This reaction is exothermic and can cause the generated hydrogen to explode, if it reaches stoichiometric proportions. This explosion will cause a pressure spike that can exceed the designed pressure of the containment vessel. In order to avoid this kind of accident, it is important that the hydrogen concentration can be reduced and controlled. This is done using passive auto-catalytic hydrogen re-combiners that recombine the hydrogen with oxygen when it is produced. (International Atomic Energy Agency, 2016) SMRs also use filtered containment venting systems in order to maintain containment integrity by controlling the excessive pressure that can be generated inside the containment in the case of an accident. The system vents the gases that might be generated in the reactor, like hydrogen, and by venting these gases the accidents associated with them can be reduced or even eliminated. Decay heat can also be released by venting, further reducing hydrogen generation that can occur at elevated temperatures. Using a filtered containment venting system will also minimize the on- and off-site contamination and help save human lives as well as reduce the impact that an accident might cause on the environment. (Bal, Jose & Meikap, 2019) SMR COGENERATION The power of SMRs can be changed rapidly to ensure that load following can be done easily. Load following means changing the power output of a power plant to adjust the demand and price of electricity as it fluctuates during the day. As an example, in the SMR that Westinghouse is developing the power levels can be changed at a rate of 5 % per minute from the interval of 100% to 20% power during daily load follow and in continuous load follow the power rate can be changed ±10% power at a rate of 2% per minute. (Westinghouse, 2021) Comparing these numbers to regular NPPs where the output variation rate must be at least 3% for the NPPs to be

71 considered suitable for load following, it becomes clear that SMRs can be adjusted much faster and are more suitable for load following. Locatelli et al. (2017) point out that modern NPPs, such as the European Pressurized Reactor, can be adjusted at similar speeds as SMRs, and that in France many NPPs use different load following programs that usually contain one or two large power changes per day. For PWRs load following is done by inserting the control rods in order to reduce the reactivity in the reactor core. Doing this might cause problems by introducing thermomechanical stresses in the reactor fuel and components, a problem that has been considered and mitigated in the design of newer reactors. (Locatelli, et al., 2017) According to Locatelli et al. (2017) the best way to carry out load following with small modular reactors is to use the reactor for cogeneration during times when the electricity demand is lower. They point out that reducing the power in the primary circuit is not ideal, as the operating cost of an NPP does not decrease significantly if the power output is lowered, however if cogeneration is carried out the economic viability might not suffer as much as if the power plant is only used to generate electricity at a lower capacity. Load following by cogeneration is done by running the NPP at nominal power all the time, without changing the conditions of the primary circuit. The amount of electricity generated is changed to fulfill the high load peaks, during which all the power generated in the power plant is converted into electricity, while during low load, the excess power is used to power external systems that are used for cogeneration. For cogeneration heat supply is preferred over electricity as most of the cogeneration processes need heat to function and heatto-electricity efficiency conversion losses can be avoided. (Locatelli, et al., 2017) Cogeneration can be done to carry out seawater desalination, as it has proven itself to be a reliable method to deliver large quantities of fresh water from seawater. Locatelli et al. (2017) claim that about 2.3 billion humans live in water-stressed areas and that in some of the countries that are in these areas, desalination has been used for nearly three decades. Desalination can be done by using two different process: thermal processes and membrane processes. Thermal processes require thermal energy, usually delivered with low-temperature steam, while membrane processes need electricity. Out of these two processes, the membrane process, more specifically

72 osmosis has become the more popular method, due to its lower overnight construction cost and total produced water costs when compared to other processes. NPPs have been used for over 200 reactor-years to carry out nuclear desalination and has proven itself a reliable method to generate fresh water. (Locatelli, et al., 2017) Some researchers have studied if SMRs could be used for cogeneration to produce gasoline as the refining process needs large of amounts of thermal energy to separate the different fractions during fractional distillation. According to Locatelli et al. (2017) the results of these studies showed that while it is technically possible to use SMRs to provide thermal energy for the refining process, the economic feasibility of doing so however is unclear as the cost competitiveness is heavily linked to the cost of gas and CO2 taxes, seeing as gas turbines are the standard today for this process. Another possible use for cogeneration is shale oil extraction. Shale oil extraction requires vast amounts of heat as the oil shale must be heated to around 330 C in order to convert the kerogen into light high-value oil, natural gas and char. Locatelli et al. (2017) point out that in non-nuclear processes around 25% of the shale oil is burned in order to generate the needed heat. SMRs would be used to heat steam that is sent into subsoil to heat the soil to temperatures between C, after which the steam would be further heated to around 370 C with the help of electric heaters. Researchers point out that for this to be viable the reactor would need to have a lifetime of 60 years and be placed within a 2 km radius of the shale oil extraction operation. Locatelli et al. (2017) also mention that some experts think that the NPP should be capable of moving between sites if some of the sites have less operation time than expected. This would give rise to different problems, especially with licensing and is therefore an extremely controversial subject. (Locatelli, et al., 2017) An option that is being investigated locally is the possibility to use SMRs for district heating. District heating is used as a method to heat buildings and the system consists of a thermal power plant, that is usually used to generate both electricity and heat, and a network of distribution and return pipes. According to Locatelli et al. (2017) many central, northern and former soviet countries have used district heating for many decades and the system has proved itself a viable heating method. The most common fuels that are used to generate heat and electricity in power are coal and gas. Locatelli et al. (2017) point out that some of the countries that use district heating

73 have experimented using nuclear cogeneration for district heating, showing that it is possible. District heating has many requirements for it to be used efficiently. Some of requirements are that the steam or hot water that is transported in the heat distribution network should have a temperature between C, a relatively close location to the customer, typically within km, and a suitable capacity, typically between MWt in large cities. It should also be noted that district heating is only used during the colder periods of the year and that a backup capacity is required to ensure that the city has access to reliable source of heat, even in cases where the power plant is not able to produce heat. (Locatelli, et al., 2017) According to Teräsvirta, Syri & Hiltunen (2020) around 50% of all heat that is produced for district heating in Finland comes from fossil fuels and high-emission fuel peat. Practically all large Finnish cities use district heating and have one or more CHP power plants. In addition to these CHP plants, heat-only boilers are also used in cases where the demand peaks are at their highest. As the price of fossil fuels keep rising the profitability of CHP plants keep going down and the demand for new ways of generating electricity and district heat have risen, sparking an interest in SMRs. According to some researches, the manufacturing costs for SMRs that only produce heat is lower when compared to cogeneration SMRs and the safety design is also simpler. As mentioned earlier, district heating with nuclear power is not a new concept and in the 1970s, a Swedish-Finnish collaboration was carried out. The goal of that project was to develop a light water reactor, called SECURE, that would have a core outlet temperature of 115 C and a power output of 200 MW. The project was discontinued, but the concept has been revisited multiple times over the years. Teräsvirta, Syri & Hiltunen (2020) mention that there are currently many MWt reactors that will become available in the current decade. The first heat-only reactor that is expected to become commercially available is the Chinese DHR400, a reactor with a 400 MWt capacity. The reactor is expected to produce hot water at atmospheric pressure and a temperature of around 90 C, making the reactor unusable in Finnish conditions as the capacity of the reactor is oversized and the output temperature too low for colder conditions. Teräsvirta, Syri & Hiltunen (2020) point out that the economics of the reactor does show promise. (Teräsvirta, Syri & Hiltunen, 2020)

74 SMRs as well as LR could also be used to produce hydrogen by water electrolysis. Hydrogen is an option to be used as a fuel in vehicles in the future, as it is one of the cleanest fuels currently available. Hydrogen is also a product that is used in many chemical processes, mainly in ammonia synthesis and in the petroleum industry. According to Locatelli et al. (2017) 17 NPPs with a capacity of 1 GW in load following mode would be needed if only 1% of the total hydrogen production was done in this way. They point out that electrolysis is not cost competitive with the production of hydrogen from natural gas, unless the cost of electricity during the night becomes much lower than it is currently. The more attractive method for generating hydrogen is the Sulphur-Iodine Thermochemical cycle, in which sulfuric acid is heated until 900 C, at which a series of reaction take place, producing hydrogen. This method is still under development, capable of producing hydrogen with an overall efficiency of around 45%, using only heat. Sulfuric acid and halogen are both extremely corrosive and the materials used for both the reactor and other equipment must be chosen carefully. (Locatelli, et al., 2017) Out of the possible cogeneration options that have been mentioned in this thesis, the most promising ones are according to Locatelli et al. (2017) district heating, desalination and hydrogen production. In the article written by Locatelli et al. (2017) more options for cogeneration were mentioned. These processes were not considered suitable for cogeneration by Locatelli et al. (2017) and due to this, the writer of this thesis has chosen not include them. (Locatelli, et al., 2017) 2.7 LIFE CYCLE ANALYSIS A life cycle analysis is a method that is used to analyze the effect that a product has on the environment during its whole life cycle. In general, an LCA is done from the cradle to the grave, meaning that the analysis is done from resource extraction until the disposal of the final product. This product can anything from a sock to a nuclear power plant, and depending on the product, the life cycle analysis can be very different. Another name for a life cycle analysis is a life cycle assessment. An LCA is carried out in four parts, in which the first part is to decide how big a part of the products life cycle will be included in the LCA. The next step is to make an inventory analysis in which a description of all materials and energy flows within the

75 product system is given, with a focus on its interaction with the environment, raw materials that are need and emissions that are released into the environment. The third step is to make an impact assessment, in which the indicator results are detailed, and the importance of every impact category is assessed by normalization and in some cases by weighing. The final step is to do a critical review of the results, determine data sensitivity and present the results. (Muralikrishna & Manickam, 2017) According to Muralikrishna & Manickam (2017) the concept of a LCA is generally easily understood and appreciated, however carrying out an LCA can sometimes be impractical as access to needed data, time, expense as well as the uncertainty regarding the dependability of the results can lead to problems. This has led to the creation and development of the streamlined LCA, in which the basic concepts of LCA are simplified to make the process more efficient and straightforward. When doing an LCA it is important that the boundaries of the analysis are chosen carefully. (Muralikrishna & Manickam, 2017) In an LCA for NPPs, the whole cycle starts with the nuclear fuel cycle, in which all the stages from mining until the finished nuclear fuel rod are included. Nuclear reactors need to be refueled on different timescales depending on the design of the reactor, meaning that fuel cycle is one of the factors that impacts the results of the LCA the most. Construction of the power plant itself is also one of the bigger parts of an LCA, as a normal NPP require many different materials, some of which release large amounts of carbon dioxide when they are manufactured, while other release less. Materials that are used in large amounts in NPP construction are concrete and steel. The work equipment like excavators, trucks etc. also release carbon dioxide when they are used during the power plant construction. In some cases, the distances that the construction workers have to travel to the power plant construction site are also taken into account when carrying out an LCA as the emissions from these trips can become considerable due to high amount of workers and possible distances that they need to travel. Operation and maintenance are also included in the LCA, with most of the emissions coming from the power plant workers commuting and plant repairs, replacements as well as refurbishments. Diesel generators are also used during outages and in some cases during maintenance, resulting in some CO2

76 emissions. The final part of an NPP LCA is decommissioning, in which the emissions from the deconstruction of the facility and building, radioactivity measurements, cutting and decontamination, and finally interim storage are taken into consideration. (Carless, Griffin & Fischbeck, 2016) 2.8 COMBINED HEAT AND POWER PLANTS CHP is a process in which both electricity and heat are generated. Traditionally CHP production is done by using the heat that is left after the steam has gone through the turbines in a power plant for district heating or cooling networks. By doing this it is possible to decrease the fuel consumption of the power plants with around 25 35% when compared to plants that generate electricity and heat separately. Fuel sources that can be used in CHP plants is quite wide with the possible fuels that the plants can use include natural gas, coal, light fuel oils, biomass and waste fuels can all be used for energy generation. (Sipilä, et al., 2005) In Finland, CHP is used for heating communities as well for the heat and power needs of the industrial section. The use of CHP for heating the largest cities in Finland was started in 1950s and 60s, while the first CHP plants for industrial use were built in the 1920s. Using CHP as an energy generation method allows Finland to produce 11% more energy when compared to a case if the heat and power were generated separately. The efficiency of CHP plants can reach the range of 80 90% according to Alakangas & Flyktman (2001), while conventional condensing power plants only reach an efficiency around 40% when producing electricity only. (Alakangas & Flyktman, 2001) Other reports claim that the efficiency of CHP plants can be higher when using certain fuel types. According to Flyktman & Helynen (2004) the efficiency for CHP plants are used to produce both heat and power can reach over 93%. The highest efficiencies are reach when using different oils and natural gas. The efficiency for bigger CHP plants will suffer a little when compared to smaller plants, however the decrease is insignificant. The increased efficiency of the power plants lowers the emissions per produced unit of power in the plants as well. (Flyktman & Helynen, 2004)

77 Different technologies can be used for combustion in the CHP plants. All CHP plants use some sort of heat engine to generate electrical power, such as steam turbines or gas turbines as examples. The only CHP technology that does not use a combustion chamber are fuel cells as electricity is generated electrochemically. CHP technology can be roughly divided into two different types of systems, namely the topping cycle and bottoming cycle. The topping cycles primary function is to generate electricity and the heat that is left after generating electricity is used for different applications. The heat can be used for district heating and hot water production or in cases where the much heat is retained, it can be used for industrial processes. In the bottoming cycle the generation of heat is the primary function. The generated heat is used for industrial processes such as iron smelting or other processes that require high temperatures. The leftover heat energy is often used to generate electricity. This cycle is also called waste energy recovery or waste heat recovery. (Breeze, 2018) The α-value of a CHP plant is a performance indicator that tells how much electricity is produced compared to the heat produced. A CHP plant usually operates according to the heat demand in the network, meaning that a plant with a high α-value produces more electricity to the grid with the same heat demand when compared to a plant that has a lower α-value. The α-value can be increased by improving the turbine process, by swapping to a newer and efficient turbine. (Sipilä, et al., 2005) According to Aalto et al. (2012), the α-value of larger steam CHP plants is usually over 0.5, while for smaller plants the value can be much lower, usually around 0.3. For CHP plants that use natural gas, the α-value can be close to 1, meaning that an equal amount of heat and electricity can be generated. Aalto et al. (2012) mention that it is necessary for CHP plants to be able to operate at different power outputs in order to be able to follow the distrcit heat demand. In some cases the heat demand is so low that the CHP plants only produce electricity, or in some cases shutdown for the summer as the power output of the power plant is too high. (Aalto, et al., 2012)

78 3 MATERIAL AND METHODS The data and numbers that have been used in these calculations have been gotten from different sources. Data needed for the Olkiluoto-3 NPP has been retrieved from the power plants company websites. For the chosen Westinghouse SMR data for this reactor has been taken from the article written by Carless, Griffin & Fischbeck (2016). The data that was needed for the construction materials, such as density and specific carbon emissions per kilogram of produced material, have been taken from companies in the field. Data that is used for the nuclear fuel chain has been retrieved from Taylor (1996) as well as Beerten et al. (2009). In the NPP section data for distance from Rauma and Pori to the Olkiluoto NPP site has been taken from Google Maps. This section of the thesis is split into different section for the different power plant types in order to make the calculations more easily understandable. 3.1 OLKILUOTO-3 Olkiluoto-3 was chosen to be reference for modern NPPs and to see how the carbon dioxide emissions of the chosen SMR relate to the larger plants. Most of the data for the NPP was retrieved from company that owns the Olkiluoto NPP as was mentioned earlier. The used data can be seen in appendix 1. The LCA that was calculated includes the emissions from the most common construction materials, emissions from the construction workers and plant workers commute trips as well as the emissions from the nuclear fuel production chain, from mining until fuel fabrication. In order to carry out an LCA, it is necessary to know how much of the different construction materials are needed in the NPP. The data available on TVO (2021) gave the concrete amount for the NPP as cubic meters and had to be converted into kilograms in order to calculate the carbon dioxide emissions. This was done for both regular concrete as well as for rebar concrete with the help of the equations 1 and 2 below. m C = V C ρ C (1) m C = Mass of concrete [kg] V C = Concrete volume [m 3 ] ρ C = Density of concrete [kg/m 3 ]

79 m RC = V RC ρ RC (2) m RC = Rebar concrete mass [kg] V RC = Rebar concrete volume [m 3 ] ρ RC = Density of rebar concrete [kg/m 3 ] The specific emissions factor for concrete was calculated from the values from the Ali, Saidur & Hossain, (2011) according to equation 3. u Conc = M Conc m ConcCO2 (3) u Conc = Concrete specific emission factor [kg CO2/kg concrete] M Conc = Cement production, EU [kg] m ConcCO2 = Cement production CO2 emissions [kg] Once the masses for the different concretes were known, the were added to together according to the equation 4 and the carbon dioxide emission for concrete was calculated with the help of equation 5, both of which can be seen below. m TotCon = m C + m RC (4) m TotCon = Total concrete mass [kg] m C = Mass of concrete [kg] m RC = Rebar concrete mass [kg] M Conc = m TotCon u Conc (5) M Conc = Total concrete CO2 emissions [kg CO2] m TotCon = Total concrete mass [kg] u Conc = Concrete specific emission factor [kg CO2/kg concrete] Similar calculations were done for the steel that is needed in the NPP. Here an educated guess had to done in order to figure out how much steel was used in the Olkiluoto-3 NPP as no data for this was given on the power plant owners website. An assumption that 1,6 times more steel is needed for OL-3 than for an AP1000 NPP is needed, as the power output for OL-3 is 1,6 times bigger. The data for how much steel is needed for an AP1000 was taken from Carless, Griffin & Fischbeck (2016). The scale up calculations were done with the equation 6 that can be seen below. m Steel = m AP1000 f (6)

80 m Steel = mass of steel in OL3 [kg] m AP1000 = mass of steel in AP1000 [kg] f = Upscale factor, based on 1,6 higher output The carbon dioxide emissions for the steel was done in a similar way as was done for concrete and the equation 7 was used for this. The specific emission factor for steel was taken from the report by World Steel Association (2021). M Steel = m Steel u Steel (7) M Steel = Total CO2 emissions [kg CO2] m Steel = mass of steel [kg] u Steel = Steel specific emission factor [kg CO2/kg steel] The next step in the LCA for OL3 is to calculate the emissions that are born from the NPP construction site workers commute trips. TVO (2021) mentioned the amount of construction workers on their homepage, and as mentioned earlier the distances from Rauma and Pori to the NPP site was measured with the help of Google Maps. An assumption that the construction workers were from Rauma and Pori, with an even split, was done in order to make calculations more manageable and it should be noted that in reality, the distance that the workers actually travelled might be very different as it is unlikely that all the workers where from these two cities. An average distance was calculated with this assumption with the help of the distances from Rauma and Pori to the Olkiluoto NPP site. According to data from TVO (2021), the NPP would be complete in 2022 and as the construction started in 2004, the construction time for the NPP in the calculations was set to 18 years. The amount of construction workers was also mentioned on TVOs webpage, and an assumption was made that 50% of the workers work every day for those 18 years, a quite optimistic assumption, as it is unlikely that this happened in reality. In order to be able to calculate the emissions from the commute trips and in order to simplify the calculations, the specific emissions from the most common car used in Finland around the middle of the construction timeline was used by the construction workers. This data was retrieved from Auto-Data.net (2021). In order to calculate the emissions from the commute trips, equation 8 was used.

81 M ConWor = W OL3 d l 0,5 average U car (8) M ConWor = CO2 emissions, construction workers [kg] W OL3 = Construction workers, OL3 d = Workdays laverage = average distance to OL [km] UCar = specific emissions car [kg CO2/km] Similar calculations were done for the plant operators with the same assumptions as above, with the same assumptions as above. Only difference in these calculations is that the carbon emissions for the construction workers is for the whole construction time, while the calculations for the plant operators were done for only one year. Equation 9 was used for this, with the only difference being the different worker numbers as well as the number of workdays. It should be noted that emissions for the plant workers can in reality be quite different as the specific emissions from cars will most likely decrease in the future. M OpeWor = P OL3 d 0,5 l average u car (9) M OpeWor = CO2 emissions, construction workers [kg] P OL3 = Plant operators, OL3 d = Workdays laverage = average distance to OL [km] ucar = specific emissions car [kg CO2/km] In this thesis the emissions for the nuclear fuel cycle were calculated with three different values. The first set of data for the nuclear fuel cycle that was used in the calculations is the values from the report by Taylor (1996) for the IAEA. As the unit for the emissions in this source was in the form of kg CO2/kg natural uranium, it is necessary to convert the amount of enriched uranium that is used in the reactor annually into the corresponding amount of natural uranium. This was done by using the data available in the article written by Carless, Griffin & Fischbeck (2016). In this article data of how much natural uranium is needed in order to produce the needed enriched uranium for different types of nuclear reactors. With the help of this data an enrichment factor could be calculated for the Olkiluoto-3 reactor, which

82 made it possible to calculate how much natural uranium said reactor consumes per year. The equation 10 was used in order to calculate the enrichment ratio. R = m UNatR m UEn (10) R = Enrichment factor m UNat R = Required natural uranium for AP1000 reactor [kg] m UEn = Required enriched uranium for AP1000 reactor [kg] The enrichment factor could now be used to convert the annual enriched uranium consumption of the OL-3 reactor into a corresponding natural uranium mass. This was done by using equation 11 which can be seen below. m UNat = AC R (11) m UNat = Natural uranium mass needed, per year [kg] AC = Annual enriched uranium consumption [kg] R = Enrichment factor The next step in the LCA is to add up the different phases of the uranium fuel chain in order to get a value for the whole chain. Equation 12 below is used for this. U u = u MinMil + u Con + u Enr + u FFab (12) U u = Total emission factor, nuclear fuel chain [kg CO2/kg Unat] u MinMil = Emission factor, Mining and milling [kg CO2/kg Unat] u Con = Emission factor, Conversion [kg CO2/kg Unat] u Enr = Emission factor, Enrichment [kg CO2/kg Unat] u FFab = Emission factor, Fuel fabrication [kg CO2/kg Unat] With the help of equation 11 and 12, the total CO2 emissions resulting from fuel manufacturing can be calculated by multiplying the answer with each other according to equation 13. M U = m UNat U u (13) M U = Emissions Uranium fuel chain, per year [kg CO2]

83 m UNat = Natural uranium mass needed, per year [kg] U u = Total emission factor, nuclear fuel chain [kg CO2/kg Unat] After this step all the CO2 emissions where add together. M tot = M Conc + M Steel + M ConWor + (M Ope 60) + (M U 60) (14) M tot = Total CO2 emission [kg CO2] M Conc = Concrete CO2 emissions [kg CO2] M Steel = Steel CO2 emissions [kg CO2] M ConWor = Emissions construction workers [kg CO2] M Ope = Emissions operators, per year [kg CO2] M U = Emissions Uranium fuel chain, per year [kg CO2] After this the electricity produced during the NPP lifetime was calculated with the help of equation 15. E L = E A 60 (15) E L = Lifetime produced electricity [kwh] E A = Annually produced electricity [kwh] From the specific emission factor for the electricity production was calculated with equation 16. U IAEA = M tot E L (16) U IAEA = Specific emission factor, IAEA [g CO2/kWh] M tot = Total CO2 emission [g CO2] E L = Lifetime produced electricity [kwh] Similar calculations were done for the specific emission factor with regards to the produced heat in the NPP during its life. In order to calculate the produced heat, the thermal efficiency of the reactor had to considered. This was done in equation 17 below. E H = E L η E H = Lifetime produced heat [kwh] E L = Lifetime produced electricity [kwh] (17)

84 η = Thermal efficiency Once the lifetime produced heat was known, the specific emission factor could be calculated in a similar fashion as earlier. UH IAEA = (M tot 1000) E H (18) UH IAEA = Specific emission factor, IAEA, heat [g CO2/kWh] M tot = Total CO2 emission [g CO2] E H = Lifetime produced heat [kwh] For the two other sets of data from the article written by Beerten et al. (2009) the calculations were done with mostly the same equations as with the data from Taylor (1996), with the differing equations mentioned here. The first difference in the calculations is the way in which the emissions from the nuclear fuel chain. In this case the unit for the data is in g CO2/kWhe, meaning the equation 12 can be used but it should be noted that the unit is different. The total carbon dioxide emission equation is different than earlier and the new equation 19 below is used in this case. M tot = M conc + M steel + M ConWor + (M Ope 60) + (U U E L ) (19) M tot = Total CO2 emission [kg CO2] M Conc = Concrete CO2 emissions [kg CO2] M Steel = Steel CO2 emissions [kg CO2] M ConWor = Emissions construction workers [kg CO2] M Ope = Emissions operators, per year [kg CO2] U U = Total emission factor, nuclear fuel chain [kg CO2/kWh] E L = Lifetime produced electricity [kwh] Equation 16 to 18 are used to calculate the rest of the results. 3.2 WESTINGHOUSE SMR The LCA calculations for the Westinghouse SMR were done mostly with the same equations that are used for OL-3, with a few differences, relating to the decrease in few areas due to the smaller size of the reactor. The data used for the calculations can be seen in appendix 2. The first notable difference is the annual consumption of fuel in the reactor. As the consumption is not directly stated, it was calculated by dividing

85 the amount of nuclear fuel in the reactor with the planned refueling time according to equation 20. AC WSMR = U reactor t refuel (20) AC WSMR = Westinghouse SMR annual fuel consumption [kg] U reactor = Total fuel in reactor [kg] t refuel = Refueling cycle [years] After this the emissions from the construction materials were calculated with equations 4,5, and 7. The number of workers and operators that are needed for W- SMR are also reduced when compared to a conventional NPP. According to Carless, Griffin & Fischbeck (2016) the reduction of workers and operators in SMRs can be between the interval of 0 73%. For the calculations done in this thesis the max value was chosen due to comparison being done to the OL-3 workers as no data was given in the source material for how many workers were assumed to be needed. For these calculations equations 21 and 22 were used. W WSMR = W OL3 ΔW (21) W WSMR = Workers W-SMR W OL3 = Workers OL-3 ΔW = Reduction factor for workers O WSMR = O OL3 ΔO (22) O WSMR = Operators W-SMR O OL3 = Operators OL-3 ΔO= Reduction factor for operators After these corrections for the workers and operators the rest of LCA calculations were done according to the equations mentioned in the OL-3 section. 3.3 COMBINED HEAT AND POWER PLANTS The data used for the calculations in this section have been retrieved from Energiateollisuus (2020) and from Tilastokeskus (2021). The districit heat companies and independent power plants for this thesis consumed atleast 700 GWh of fuel for cogeneration production. Equation 23 is used to calculate the delivery of district heating of the different plants.

86 E DHdeliv = E NETfuel E RECheat (23) E DHdeliv = Delivery of district heating [GWh] E NETfuel = Net production with fuels [GWh] E RECheat = Recovery of heat [GWh] Once the amount of delivered district heating is known, the electricity that is produced at the same time as the delivered district heating is added together according to equation 24. E tot = E DHdeliv + E el (24) E tot = District heating + produced electricity [GWh] E DHdeliv = Delivery of district heating [GWh] E el = Electricity production associated with district heating [GWh] After this the emission factors provided by Tilastokeskus (2021) have to converted from ton CO2/TJ to ton CO2/GWh. This is done with the help of equation 25. t CO 2 /GWh = t CO 2 /TJ 3,6 (25) The conversion must be done in order to be able to calculate the emissions from the different fuel source that the different power plants use. The different fuels were compiled into tables that can be seen in appendix 3. It should be noted here that for fuels that are specified as complete biofuels, the emissions are regarded as zero and are not considered in the calculations. The composition of wastes used for energy generation are assumed to have a bio-content of 50%. The emissions from each fuel source were calculated with the help of equation 26. M fuel = u fuel E fuel (26) M fuel = Total Emissions per fuel [t CO2] u fuel = Specific emission for different fuels [t CO2/GWh] E fuel = Amount of energy from fuel [GWh] Once these calculations were done for every type of fuel that were used in the different CHP plants, the emissions from the different fuels were added up per plant. Equation 27 was used for this. M tot = M fuel M fueln (27) M tot = Total Emissions [t CO2]

87 M fuel1 = Total Emissions per fuel [t CO2] Once the total emissions per power plant were known, the final step was to calculate the specific emissions per power plant in regard to the district heating and electricity generated. This was done only for the independent CHP plants. Equation 28 was used to carry out the calculations for this. It should be noted that if a fuel is considered a biofuel, the emission factor for that fuel is zero in the calculations. The emissions factors for different fuels can be seen in appendix 3. k = M tot E tot (28) k = Specific emission factor [g CO2/kWh] M tot = Total Emissions [t CO2] E tot = District heating + produced electricity [GWh]

88 4 RESULTS In this section of the thesis the results for the different power plant types are presented. The results for Olkiluoto-3 and the Westinghouse SMR are presented in table 1 that contain the results for the specific carbon dioxide emissions for both electricity and heat generation. Three sets of data were used for the nuclear fuel chain as mentioned earlier in this thesis. The calculations for the nuclear options can be seen in appendix 1 for Olkiluoto-3 and appendix 2 for the Westinghouse SMR. Table 1. Results of the LCA for OL-3 and W-SMR Energy production Unit Plant Dataset 1 Dataset 2 Dataset 3 Electricity only g CO 2/kWhe OL Heat production only g CO 2/kWhheat OL Electricity only g CO 2/kWhe W-SMR Heat production only g CO 2/kWhheat W-SMR Dataset 1: Taylor (1996); dataset 2: Beerten et al. (2009), Lenzen; dataset 3: Beerten et al. (2009), Torfs; The results for the specific CO2 emissions for the district heating companies can be seen in table 2. The results take into consideration the cogeneration of district heat and electricity produced in the companies. Table 2. Specific carbon dioxide emissions for district heating companies Company g CO 2/kWh Fortum Power and Heat Oy Fortum Power and Heat Oy, Espoo Fortum Power and Heat Oy, Joensuu Helen Oy Kuopion Energia Oy Lahti Energia Oy Lahti Energia Oy, Lahti Loimua Oy Nevel Oy Oulun Energia Oy Pori Energia Oy Pori Energia Oy, Pori Savon Voima Oyj Tampereen Sähkölaitos Tampereen Sähkölaitos, Tampere Vantaan Energia Oy 254.1

89 Table 3 presents the results for the independent CHP plants with both district heat and electricity production included when calculating the results for the emissions and efficiency. Table 3. Specific carbon dioxide emissions for independent CHP plants CHP plant g CO 2/kWh Efficiency Jyväskylän Energiantuotanto Oy Jyväskylän Voima Oy Kaukaan Voima Oy, Lappeenranta Turun Seudun Energiantuotanto Oy Vaskiluodon Voima Oy, Vaasa

90 5 DISCUSSION The results that were obtained in this thesis show that the specific emissions from nuclear power plants are much lower when compared to CHP plants and district heating companies, making the nuclear option a viable solution for reduced carbon dioxide emissions in the power generation field. The specific emissions from the Westinghouse SMR and OL-3 reactor are quite close to each other, with W-SMR having lower specific emissions in 2/3 of the datasets used in the LCAs. The results that were calculated with the data from Taylor (1996) report are much lower, almost half, for the Olkiluoto NPP when compared with the results for the Westinghouse SMR. The reader should keep in mind that most of the emissions that are linked the nuclear life cycle are not made in Finland. The only direct emissions that are released in Finland, are the emission from construction as well as the plant workers commute trips. The nuclear power plant emissions differ quite considerably from the results achieved by Sovacool (2008). The author of that report calculated an average value of different LCAs done over the years. The biggest reason for the difference is that Sovacool (2008) calculated the emissions for the whole lifecycle, from the start to the end, while this thesis only took into account the emissions created during construction, fuel manufacturing as well as from the commute trips done by workers and plant operators. Graph 1 shows the average results that Sovacool (2008) achieved. From this it becomes apparent that the exclusion of the different stages after making the NPP operational has led to the results becoming somewhat smaller than they would have been, if decommissioning and fuel storage had been considered. It should be noted that the results for the Westinghouse SMR cannot be directly compared to Sovacool s (2008) result as the difference in technology is certain to affect the result to some degree, seeing as the construction materials needed, required construction time etc. are lower for the SMRs.

91 Figure 14. Specific carbon emissions for an NPP (Sovacool, 2008) Looking at the source materials that Sovacool (2008) used for his calculations, the results achieved in this thesis came close to the lower end of the material used by Sovacool (2008). The range of the values calculated by Sovacool (2008) varies from 1.36 g CO2/kWh up to CO2/kWh. It should be noted that the specific emission factor calculated by Sovacool (2008) is calculated by only considering the electricity produced in the NPP, meaning if the emission value was calculated with the heat produced, the value would be lower. The data used in this thesis for the nuclear fuel chain is quite old, and the values today might be drastically different, seeing as different machinery used in the nuclear fuel chain is getting more energy efficient. The data used for the Westinghouse SMR are also scaled down from conventional NPPs, meaning that once more SMRs are built and operated, the values can become lower or higher than what they are currently estimated to be. The specific emissions calculated for the independent CHP plants and district heating companies are quite close to the values reported by Fingrid (2021). The emissions reported are 330 g CO2/kWh for plants that cogenerate electricity and district heating for cities and 170 g CO2/kWh for district heating and electricity that is generated as waste energy from different industries. Most of the values in this thesis are between these two values and some of them are below. In bigger cities, coal and natural gas are used more compared to renewable sources, especially in Helsinki, while smaller cities use a bigger share of biofuels. Some form of peat is used by almost every CHPplant, with the exceptions of Helsinki and Espoo. Unlike the emissions for the