1.0 INTRODUCTION The majority of a person s environmental dose is due to the natural radiation background to which humans have shown remarkable

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1 1.0 INTRODUCTION In nuclear energy the source of power production is nuclear fission and nuclear fusion processes. In the nuclear fission process, nuclear energy is produced through nuclear chain reaction. Fission is caused by absorption of neutron in heavy nucleus to form an excited compound nucleus. The compound nucleus splits into two (mainly) or three fission fragments with emission of radiations such as beta, gamma, neutron, neutrino, etc. The majority of fission fragment nuclides after fission have mass number peaks at 95 and 140. The nuclear energy per fission of a uranium nucleus is found 200 MeV. Most of fission energy (~85%) appears as kinetic energy of the fission fragments and remaining is distributed among various types of radiations. About 15% nuclear energy is released instantaneously as gamma-rays, fission neutrons and rest is released gradually by radioactive decay of fission products through the emission of gamma-ray, neutron, beta and neutrino. The energy released during the fission process is difference in binding energy between heavy nucleus and fission fragments. Average 2 to 3 neutrons are being emitted during nuclear fission, and these neutrons further continue chain reaction. The nuclear fission reaction with any energy occur with fissile nuclides (e.g. 233 U, 235 U and 239 P) whereas the fast neutron (>1 MeV) may split 238 U or 238 Th nucleus. A number of heavy nuclei undergo fission reactions, but the most important nuclei are 233 U, 235 U and 239 Pu for power production. Nuclear reactor is a device in which sustained nuclear reaction continues by chain reaction process. The nuclear reactors are being used for electricity production, nuclear submarine, isotope production, plutonium production, research and development etc. In nuclear reactors, heat and various types of radiations are being emitted during and later stage of the nuclear reaction. Later stage heat produced in the nuclear reaction is called as decay heat. Heat produced from the nuclear fission is passed to a moving fluid (e.g. water, sodium or gas) for steam production in secondary system to run the turbine. Today, there are about 450 nuclear power reactors are being used for electricity production in 30 different countries. Nuclear reactors are classified depending on the type of processes (fission/fusion), type of fuel and the type of applications. The reactors are called as fast reactor and thermal reactor based on the energy of neutron for fission process. There are two types of thermal reactors, light water reactor (LWR) and pressurized heavy water reactor (PHWR). The chain reaction in these reactors is sustained by thermal neutrons. The light water reactors are boiling water reactor (BWR), pressurized water reactor (PWR) and Voda-Voda Energy reactor (VVER). The nuclear power production from LWR is more than 90% in the world. The contribution from PHWR is about 6%. In PHWR, natural uranium (0.7% 235 U) is used as fuel while low enriched uranium typically 3 to 4% 235 U is used in LWR. In thermal reactors, moderator is required to slowdown the fast neutrons. Presently, BWR (2 nos.), PHWR (17 nos.) and VVER (1 no.) types of reactors are under operation and PHWR (4 nos.) and VVER (1 no.) are under construction in India. The PHWR are CANadian Deuterium Uranium (CANDU) type reactors; first was installed at Rawatbhata site, Rajasthan. In the fast reactor, the chain reaction is sustained by fast neutrons. Due to this, these reactors don t require any moderator system. The fast reactors use enriched fuel (>20%) compared with the thermal reactors. Sodium is used as a coolant in a typical fast reactor. India is in the advance stage of fast reactor technology. A breeder reactor is a nuclear reactor capable of generating more fissile material than it consumes. These reactors are able to achieve this because their neutron economy is high 1 The majority of a person s environmental dose is due to the natural radiation background to which humans have shown remarkable adaptability; it provides, therefore, a reasonable baseline for judging the significance of radiation exposures from other human activities. -Floyd L. Galpin 1

2 enough to breed more fissile fuel than they use from fertile materials (e.g. 238 U or 238 Th). The fuel breeding is possible to achieve by thermal reactors as well fast reactors. In PHWR, heavy water (D 2 O) is used as moderator system and coolant in primary heat transport (PHT) system. During interaction of neutron with deuterium, tritium (low-energy beta emitter) a long-lived activation product is produced. The tritium is contained in the moderator and PHT systems. However, due to high temperature and pressure a very low amount of tritium in vapor form may be found inside reactor building. Though, tritium production is insignificant in light water reactors. During reactor operation, gamma-ray is produced due to prompt fission of fuel, delayed decay of fission fragments and activation of structural and construction materials. The gamma-rays released within Sec are called as prompt fission gamma-rays with energy up to 7.8 MeV per fission [1]. Typical corrosion products are 24 Na, 28 Al, 51 Cr, 56 Mn, 59 Fe, 60 Co, 64 Cu, 65 Zn, 65 Ni, 66 Cu, 85 Kr, 95 Zr, 95 Nb, 97 Zr [1]. In a typical reactor mostly 16 N, 19 O, 24 Na, 41 Ar, 46 Sc, 51 Cr, 54 Mn, 56 Mn, 58 C0, 59 Fe, 60 Co, 65 Ni, 65 Zn, 88 Rb, 95 Zr, 106 Ru, 108 Ru, 113 Sn, 125 Sb, 131 I, 133 Xe, 134 Cs, 137 Cs, 138 Cs, 140 La, 141 Ce, 144 Ce, 131 Te, 140 Ba, 238 U, etc. are generated during reactor operation [2]. 16 N is a major source of high energy gamma-ray photon (6 to 7 MeV) during power operation originates from the PHT and moderator system by neutron capture reaction. The capture gamma-ray photon (0 to 10 MeV) are produced by reaction of thermal neutrons with structural/construction materials, fuel elements such as aluminium, beryllium, iron, sodium, deuterium, zirconium and uranium present in the reactor core. Capture gamma-ray is major source of secondary gamma radiation in the reactors. The most significant isotopes in terms of dose are isotopes of noble gases, iodine and cesium. In a reactor, photon emission probability of energy range from 0.5 to 4.5 MeV is very large compared with ~5 to 10 MeV [3,4,5]. The prompt fission gamma-ray has continuous energy spectrum from 0.5 to 10 MeV, but the radiation intensity is negligible beyond 7 MeV [6]. It has been found that, most of the prompt neutrons have energies between 1 to 2, and some are with energies in excess of 10 MeV [6].The delayed neutron energy is found to be below 1 MeV (e.g. 87 Br and 137 I). Neutrons are also being produced by photo-neutron reaction with deuterium present in the moderator and PHT systems. Typical fission products in the discharged bundle have been investigated by Pal and Jagannathan [3] which are 36 Kr, 37Rb, 38 Sr, 39 Y, 40 Zr, 41 Nb, 42 Mo, 43 Tc, 44 Ru, 45 Rh, 46 Pd, 47 Ag, 48 Cd, 49 In, 50 Sn, 51 Sb, 52 Te, 53 I, 54Xe, 55 Cs, 56 Ba, 57 La, 58 Ce, 59 Pr, 60 Nd, 61 Pm, 62 Sm, 63 Eu, 64 Gd, 65 Tb and etc. It is found that the major contribution to the radioactivity after long time of reactor shutdown is due to longlived isotopes such as 93 Zr (t 1/2 = y), 94 Nb (t 1/2 = y), 99 Tc (t 1/2 = y), 125 Sn (t 1/2 = 10 5 y), and 135 Cs (t 1/2 = y). A large spectrum of mass numbers of radionuclides is being produced during the fission reactions which are of different half-lives; some are short and others are very long. Alpha emission is spontaneous fission from the fuel elements. Antineutrinos are also emitted from reactors and an attempt for calculation of antineutrinos in CANDU reactor has been reported by Christopher et al. [7]. The typical range of alpha and beta particles in air are few centimeters and more than 1000 centimeters, respectively. The alpha and beta particles directly ionise the medium, and ionisation density thru alpha is very large compared with other radiations. Due to limited range, most of the alpha and beta particles are absorbed in the reactor core and structural materials. Normally, the fission products are retained inside sheath of the fuel elements, and only a small fraction may reach into the PHT system which deposits on inner surface of the lines. The activation products arising in the PHT or in any other process fluid may be transported 2

3 from the core to outside of it. The activation products could also arise in the structural components of the reactor and the reactivity control systems. Therefore, main types of radiation sources taken into consideration in the shielding are neutron and gamma-ray. The gamma-rays and neutrons are indirectly ionising radiations. Fusion power is another type, where the nuclear energy is generated by nuclear fusion process. In the fusion reactions, two light nuclei fuse to form heavier nucleus. In the nuclear fusion process major isotopes of hydrogen, lithium and helium are being used as fuel. The D-T, H-D and He-He reactions produce amount of energy 17.59, 5.5 and 12.8 MeV, respectively. Fusion reactor has the potential to provide sufficiently high energy compared with fission reactor. The deuterium can be extracted using present technologies and tritium is produced in PHWRs or would be bred from a lithium blanket using neutrons produced in the fusion reaction itself. When deuterium and tritium fuse, a high-energy neutron (~14 MeV) is being released. This high energy neutron irradiates the structural components of the fusion reactor which emit gamma-rays. In a fusion reactor, typical radionulides are 7 Be, 14 C, 24 Na, 35 S, 36 Cl, 39 Ar, 55 Fe, 59 Ni, 60 Co, 53 Ni, 93 Mo, 94 Nb, 99 Te, 113 Cd, 121 Sn, 205 Pb, 210 Po, etc. Therefore, the fusion reactors are also source of tritium, gamma-rays and neutrons. The nuclear fission reaction doesn t require any additional source of energy except neutron to excite nucleus to form a compound nucleus. However, the nuclear fusion reaction requires additional energy to overcome the Colombian repulsive force between two nuclei. Therefore, the fusion reaction occurs at very large temperature at around 10 7 K. Presently International Thermonuclear Experimental Reactor (ITER) is a fusion reactor with the collaboration of different countries. Reactor accidents are rare in nature due to advance safety features in nuclear technology and continual review and improvements processes. However during nuclear reactor accidents of Three Mile Island and Chernobyl, a mixture of short-and long-lived radionuclides released due to the failure of fuel bundles [8,9]. It is found to be various types of radionuclides released during nuclear reactor accident [10].The short-lived nuclides released in large amount (in terms of activity) compared with long-lived. The radionuclides were halogen ( 129 I, 131 I, 132 I, and 133 I), telluride ( 132 Te), alkali metal ( 134 Cs, 136 Cs and 137 Cs), noble metals ( 103 Ru, 106 Ru and 99 Mo), refractory oxides ( 95 Zr, 95 Nb), alkaline earth ( 140 Ba), rare earth ( 104 La, 141 Ce and 144 Ce), transuranics ( 147 Nd, 239 Np), inert gases ( 133 Xe, 85 Kr, 88 Kr). The radionuclides 129m Te, 132 Te, 134 Cs, 136 Cs, 137 Cs, 131 I, 132 Iand 133 I are volatile in nature. The energy of the radionuclides released during accident is in the range of MeV ( 133 Xe) to MeV ( 136 Cs). Similar types of radionuclides are also reported during recent Fukushima accident in Japan. The noble gases are more prone to escape from reactor containment during initial phase of the accident. Therefore, reactor accident management may require confinement of the radioactivity. The neutrons are categorised based on energy as thermal ( ev), epithermal (1 ev), slow (1 kev), fast (100 kev E 10 MeV), very fast (10 MeV<E<50 MeV) and ultra-fast (E>50 MeV). Neutrons are produced in various types of nuclear reactions. Also the neutron can be produced using alpha particle interaction with light elements (e.g. beryllium, boron, or lithium). When an alpha emitting nucleus such as 226 Ra is mixed with beryllium, neutrons with an energy spectrum up to 13 MeV with most probable about 5 MeV are released by (α, n) reaction. Photo-neutron is another category of neutron production similar to the (α, n) source, (γ, n) reaction is used to produce the neutrons. The photo-neutron emission is possible for high energy gamma-ray (>2 MeV). The photo-neutrons are found mono-energetic if the gammaray source is mono-energetic such as 24 Na emits photon of energy 2.76 MeV. The strong 3

4 neutron source in nuclear reactors is (γ, n) due to presence of hard gamma-rays from 16 N. In general the neutrons are produced by common reactions 9 Be(α, n) 12 C, 2 H(γ, n) 1 H, 9 Be(γ, n) 4 He, 2 H( 2 H, n) 3 He, 2 H( 3 H, n) 4 He etc. A fission takes place when a neutron interacts with a fissile materials such as 235 U, is called as induced fission reaction. However, a nucleus undergoes spontaneous fission due to instability of heavy nucleus without any external source. Nevertheless, the spontaneous fission probability is quite low. The spontaneous neutron emission rate from 235 U, 238 U, 239 Pu, 240 Pu and 252 Cf are , , 0.022, 920 and neutron/g/s, respectively. The neutron emission per fission is in the range of 1.86 to The energy of neutron in spontaneous fissions is found in the range of 1 to 3 MeV. The neutron flux near the core of a fission reactor is very high typically neutron/cm 2 /s [11].The energy spectrum of the neutrons is found 5 to 7 MeV with peak at 1 to 2 MeV. The energy of these neutrons is reduced to thermal energy within reactor core, but there are fast neutrons present in the core also. It has been found that the prompt neutron energy spectrum ranges 0.18 to 12 MeV with most of neutron having energies between 1 to 2 MeV during the uranium fission [6]. Tritium is a beta emitter with maximum energy of 18.6 kev and an average energy of 5.7 kev. The penetration of 5.7 kev beta in water and soft tissues is 0.42 µm which varies with the beta energy. Tritium atom is chemically analogous to hydrogen atom and exhibits similar chemical properties. The tritium transform into helium nucleus after fission of beta particle. When tritium reacts with water, form HTO (tritiated water) is most abundant form. The tritium atom replaces hydrogen by isotopic exchange process. This HTO is most probable tritiated form in the environment for exposure to human body. Tritium is found in the forms of HTO, DTO, TH [12], [13] T 2 and vapor or gaseous. The oxide form of tritium (HTO/DTO) is more prevalent for internal exposure than the other forms. The diffusion coefficient of HTO at 25 0 C is 2.44± cm 2 /sec [14]. Tritium closely follows the reactions similar to ordinary hydrogen. However, the relatively large mass difference between hydrogen and tritium, the isotopic exchange process changes. The properties of tritium and tritium compounds are discussed by Jacobs [12]. The dominant tritium production reactions are 2 D(n, γ)t, 10 B(n, 2α)T, 11 B(n, 11 Be)T, 14 N(n, 12 C)T, 9 Be(n, 8 Be)T, 6 Li(n, α)t and 3 He(n, p)t. Tritium is being produced in nuclear fission reaction at a rate of 1 atom per to fissions in the natural uranium, enriched uranium and a mixture of transuranium nuclides [15,16]. The naturally occurring tritium is produced by primary cosmic-ray reactions [17] through the reactions as 14 N(n, C)T and 2 D(d, p)t. Tritium is a byproduct in a nuclear fission process and is an activation product of deuterium nucleus. Sources of tritium in the nuclear power plants are 2 D(n, γ)t, 10 B(n, 2α)T, 7 Li(n, α), 3 He(n, p)t and 235 U(n, f)t. CANDU based heavy water reactors use D 2 O as a moderator, reflectors and coolant. Tritium production rate by 2 D(n, γ)t in PHWR is very large due to large quantity of heavy water in moderator and PHT systems. The amount of tritium production in thermonuclear reactions is several orders of magnitudes higher than the amount of tritium produced in fissions reactors. The workers are exposed by tritium in reactors through inhalation, ingestion and skin absorption pathways. Radiological protection requires in-depth study of the radiation sources, their interaction with the medium/shielding and the receptor/dosimeter. The medium is shielding materials or the air; receptor is human body and dosimeters are radiation monitors and detectors. The radiation is being measured using different types of measuring instruments and dosimeters. The gamma-ray and neutron expose structural material, equipment, radiation monitors, dosimeters and human body. 4

5 In view of radiation protection in reactors, appropriate shielding materials are required. Since the gamma-rays and neutrons are emitting from the reactors, the materials should be such that they should shield human body from both types of radiations. In general, we know that high-z materials should be used to attenuate gamma-ray, and low-z materials for removal of neutrons. The radiation measuring instruments are used for ambient radiation measurement and the dosimeters for personnel dose measurement. The dosimetric materials are chosen which represent radiological characteristics equivalent to human body organs. The energy spectrum of the gamma-ray photons emitted during reactor operation and accident are quite different and large. Proper shielding is required to achieve acceptable level of ambient radiation inside the reactor building for control of exposure. All applications require detailed investigation of radiation interaction parameters for shielding, dosimetric and biological materials. Tritium is found in thermal fission reactors (mainly PHWR) and fusion reactors. Therefore, tritium removal behavior from human body is also very essential for radiation protection. When radiation interacts with any material, it ionises. The radiation received by human body is accounted in term of dose. The radiation exposure is mainly of two types, one is external and other internal. External exposure is caused mainly by gamma-rays and neutrons from the radiation sources outside the body. Alpha and beta contribute negligible or unless the sources are very close to the body. Therefore, shielding materials should be chosen based on the energy of gamma-ray and neutron emitting in the reactors to maintain well below acceptable radiation level [4]. Internal exposure can arise from alpha, beta, gamma-ray, or neutrons when radionuclide has entered into the body. Routes of intake of radionuclides resulting in internal exposure are inhalation, ingestion, injection and absorption. The radionuclides entered into the body irradiate tissues and organs until they are not excreted from the body or are completely decayed. Internal dose monitoring is done by two methods such as in-vivo and in-vitro monitoring. In-vivo monitoring is used for those radionuclides which emits penetrating radiations like X-or gamma-ray. In-vitro method is used for bio-assay sample analysis. In reactors, the exposure from alpha is negligible which can be avoided by standard radiation protection practices. The internal exposure due to intake of radionuclides can be estimated using different types of bio-assay data such as counter measurements of lung and body, radioisotope concentration in urine, etc. Internal radiation doses are not possible to be measured; they are calculated based on estimated/measured nuclide intake, an estimated/measured quantity in an organ, or an amount of nuclide eliminated from the body. Detail discussion on internal dosimetry are provided by Cember [18] and Martin [19]. The reactor core is shielded by biological shielding, and the entire assembly is inside the single or double containment. The biological shielding and containment are made of different types of concretes. Initially the biological shielding was made of ordinary concretes. But nowadays, heavy concretes are being used for better shielding effectiveness. The structural materials for the reactor systems (e.g. calandria shell, end-shield and piping) are made of various types of alloys. Lead equivalent shielding glasses are essential for radiation shielding to visualize the nuclear reactions or proper function of various parts of nuclear reactor under the harsh conditions of nuclear radiation exposures. The control rods in nuclear reactors are made of neutron absorbing materials (e.g. boron, cadmium, gadolinium and etc.). These materials have large neutron absorption crosssection to remove the neutron and discontinue nuclear chain reaction. In general neutron 5

6 shielding requires low-z (hydrogen, lithium, beryllium, carbon, boron and etc.) materials as the energy transfer is large for them. Compounds of boron are also being used as a supplementary in neutron absorbing rods. The boron containing materials are also being used in the mixed field of gamma-ray and neutron radiations. Hydrogenous hydride and borohydride metals, oxide dispersion-strengthened (ODS) steels and superconductors are potential candidates for fusion reactors and IV generation reactors. The hydride and borohydride metals and ODS steels are being used for neutron and gamma-ray shielding. Also common materials such as soil, fly-ash, marble, plaster of paris, gypsum, brick, cement and lime-stone are being found to be shielding materials. Different types of shielding materials are required for a gamma spectrometer which are furnished in detail by Martin [19] and Gilmore [20]. The effective shielding materials are very important in fission and fusion nuclear reactors to minimise the exposure. Ambient radiation measurement is carried out by using various type radiation detectors. These detectors are proportional counter, ionization chamber and Geiger Muller counter. The active volume of these detectors is filled with mixture of gases to detect radiation. The detail of the gaseous mixture based detector is furnished by Knoll [21]. Also the detail of various types of detectors used in gamma-ray spectroscopy is reported by Gilmore [20]. Personal radiation dose is measured using different types of dosimeters such as optically stimulated luminescent (OSL), thermoluminescent dosimeter (TLD), gel dosimeter, solid state nuclear track detector (SSNTD) etc. Overview on some personal dosimeters is given by Cember [18] and Martin [19]. Gamma-ray interaction parameters of these dosimeters are required for proper selection of dosimetric material. The radiation interaction parameters and photon buildup factors for personal dosimeters are essential for dosimetry applications. The personal dosimeter for gamma-ray dose measurement should be radiological equivalent to human body organs to represent radiation measuring quantities. Therefore, the studies on biological materials such as human body organs and their tissue substitutes are also very important. In the present thesis we present the biological half-life of tritium; interaction parameters of photon and fast neutrons with shielding materials and photon interaction with dosimetric, biological and tissue materials. We have investigated the photon buildup factors in shielding, dosimetric, and biological and tissue substitute materials. This study is expected to bring out the outcome which would be utilised for purpose of radiological protection, dosimetry and shielding design applications in the nuclear reactors. We have calculated the mass attenuation coefficients, linear attenuation coefficients, half-value layer, tenth-value layer, effective atomic numbers, effective electron densities, airkermas and buildup factors for the shielding and dosimetric material, their dependencies upon the photon energy, chemical compositions and other parameters. We have calculated the macroscopic effective removal cross-section of the materials for fast neutron for assessment of neutron shielding properties. We have also studied the comparison of effective atomic numbers of various materials using different methods and computer programs. In chapter 1, we have discussed about different types of reactors, radiation sources and requirements of shielding, radiation measurement and dosimetry. In chapter 2, we have given brief literature survey. In chapter 3, we have described the interaction of radiation with matter. In chapter 4, we have described theory, analysis and calculation methods for biological half-life of tritium, computation of gamma-ray interaction parameters, buildup factors, G-P fitting method, effective atomic numbers and effective electron densities, removal cross-section for fast neutron and various computational methods. In chapter 5, we 6

7 have discussed the estimation of biological of tritium and it s distribution. In chapter 6, we have discussed various types of radiation shielding materials and their effectiveness. In chapter 7, we have discussed gamma-ray interaction and buildup factors for dosimetric materials and radiation detection medium. In chapter 8, we have discussed gamma-ray interaction with biological and tissue substitute materials and comparison of various methods and programs for computation of effective atomic numbers. In chapter 9, we have concluded our research work. 7