Development of Nuclear Power Reactor Shielding Using Two Different Types of Heavy Concrete

Transcription

1 The Egyptian Arab Journal of Nuclear Sciences and Applications Society of Nuclear Vol 50, 3, ( ) 2017 Sciences and Applications 9. ISSN Web site: esnsa-eg.com Development of Nuclear Power Reactor ing Using Two Different Types of Heavy A. A. El- Sawy Nuclear & Radiological Regulatory Authority, Cairo, Egypt (ESNSA) Received: 1/6/2014 Accepted: 9/9/2014 ABSTRACT Nuclear power stations and nuclear research reactors use concrete of different types for building biological shields. Different types of aggregates are used to form concrete of different properties to achieve radiation safety. This paper is concerned with the study of the mathematical modeling of nuclear power reactor using Mont Carlo code to calculate the neutron and gamma ray flux and dose rate distribution at different s in biological shield of a nuclear power reactor using two types of heavy weight concrete. These are made from local magnetite ores mixed with steel punching and basalt ore. The has a density = 5.11 g/cm 3 and the has a density = 3.05 g/cm 3. The results of calculation of neutron and gamma ray fluxes and dose rate distribution through these concretes were displayed in the form tabulated data and fluxes and dose rates attenuation curves. The obtained results are compared with their counterpart calculated for ordinary concrete of = 2.3 g/cm 3. According to the obtained data, heavy weight concretes made from steel magnetite can be used for a proper design of biological shields for nuclear power plants and their associated facilities. Keywords: Radiation shielding / Steel- magnetite- concrete / Magnetite- basalt concrete / Mont Carlo Method. INTRODUCTION Protection against nuclear radiation is a very important topic to be investigated in nuclear science. One of the most widely used materials for suppressing nuclear radiation emitted from nuclear reactors is concrete. These types of concrete are basically made of a mixture of Portland cement, fine and coarse aggregates and water. Ordinary concrete is the cheapest, easy ore to prepare in different compositions and easy to form and to use in construction works (1, 2). s made of concrete can be used to attenuate both neutrons and gamma rays. The concrete shielding properties may vary depending on the material components of the concrete. Aggregates are the largest constituent (about 70%-80% of the total weight of normal concrete) (3, 4). Radiations shielding concrete contains heavy and light weight materials and therefore are quite appropriate to attenuate gamma rays and neutrons of different energies. The mix proportion of concrete largely affects the radiation-shielding characteristics (5, 6, 7, 8). is considered as an excellent and versatile material which can be used for constructing biological shields around the core of nuclear power plants and their associated facilities It is a relatively inexpensive material, which may be easily handled and cast into complex shapes. It contains a mixture of many light and heavy elements and therefore has good nuclear properties for the attenuation of photons and neutrons (9, 10). By varying its composition and density the shielding characteristics of concrete may be adapted to a wide range of uses. has been used in the construction of nuclear facilities of two primary properties, its structural strength and its ability to 151

2 attenuate and absorb nuclear radiation. The use of concrete in nuclear facilities for containment and shielding of radiation and radioactive materials has made its performance crucial for the safe operation of the facility (11, 12). is by far the most widely used material for reactor shielding due to its cheapness and its satisfactory mechanical properties. It is usually a mixture of hydrogen, light nuclei and nuclei of fairly high atomic number. It is therefore efficient both in absorbing gamma rays and slowing down fast neutrons by elastic and inelastic scattering. The hydrogen contained in the water of hydration of cement is sufficient for the rapid thermalization of the intermediate energy neutrons (13). Increasing the density of the concrete shield by adding a heavy material to it such as iron punching has a major contribution in the suppression of gamma rays. The efficiency of the biological shield is affected by the heat generated from the attenuation of neutrons and gamma rays within the first layers of the shield. A lot of work was done on the calculation of neutron and gamma ray flux and dose rate distribution in different types of shield using different materials and different radiation sources (14). Special aggregates for radiation shielding concrete are either natural mineral aggregates or synthetic aggregates; heavy mineral aggregates are used to attenuate photons. They are mainly heavy mineral ores such as magnetite, barites and hematite. For the attenuation of neutrons, natural hydrous aggregates are used such as bauxite, serpentine and limonite as well as boron additives in the form of calcium borates. Synthetics aggregates are used to produce concrete of high density. Examples for these aggregates are ferrophosphours and metallic iron products such as steel punching and iron shot. Magnetite is one of the most widely used types of heavy aggregates for high density radiation concrete. Iron and steel scrap may be used as a coarse aggregate in conjunction with fine aggregates of either magnetite or limonite. Boron and boron compounds such as borax are often used in concrete to increase the probability of neutron capture without producing secondary capture gamma rays of high energy (15, 16). In the present work, the radiation shielding properties of special concretes made from local magnetite aggregates combined with either steel punching or basalt, i.e. steel- magnetite concrete (S-M) with density of 5.11 g/cm 3 and basalt- magnetite concrete (M-B) with density 3.05 g/cm 3 were investigated and compared with their counterpart ordinary concrete of density 2.3 g/cm 3. The two types of materials have been used as a material of a biological shield in nuclear power reactor by using a mathematical model to calculate the neutron and gamma- ray fluxes and gamma dose rates through different layers of the shield (17). THEORETICAL STUDY The study is based on the modeling of nuclear power reactor with Mont Carlo (MC) method to analyze the neutron and gamma ray fluxes and dose rates distributions over the core, through the pressure vessel and biological shield. The Mont Carlo method is well suited to solve complicated three-dimensional, time- dependent problems. The MC can be used to duplicate theoretically a statistical process (such as the interaction of nuclear particles with materials) and is particularly useful for complex problems that cannot be modeled by computer codes that use deterministic methods (18). The MCNP Code was developed for analyzing the transport of neutrons and gamma rays (hence NP for neutral particles) by the Monte Carlo method. The code deals with the transport of neutrons, gamma rays, and coupled transport, i.e., transport of secondary gamma rays resulting from neutron interactions. The code can also treat the transport of both primary electrons emitted from the source and secondary electrons created from gamma-ray interactions (19). The calculation of the flux, energy distribution and dose of neutrons and gamma rays is important for the investigation of radiation shielding materials and shield design. The nuclear power reactor model contains the radial components including the core, coolant channels, a core blanket of 3cm thickness, a thermal shield of 9.15 cm thickness, a pressure vessel of 22.8 cm thickness and biological shield of 175 cm thick and constructed of concrete and extends vertically to the top of the cylindrical section of the reactor vessel. The source characterizations developed in this work refer to a 152

3 PWR operating at a power level of 1650 MW (th). The same model was used for a biological shield that was constructed from ordinary concrete with density 2.3 g/cm 3 (20). The geometry of the reactor shield model is shown in figure 1. Fig. (1): Plan view of the homogenized reactor design 1: Reactor core (d= 2.45 m) 2: Pressure vessel (22.8 cm thickness) 3: Biological shield (1.75 m thickness) Development of a reactor shield by using the two different types of heavy concrete, made from steel- magnetite and magnetite- basalt was very important in shield design. These two types of concrete aggregates, as the main component of concrete, were analyzed to provide radiation protection. The study aims to develop a reactor radiation shield to reduce the radioactive doses. A comparative study is made between the dose rates when using ordinary concrete ( = 2.3 g/cm 3 ), and these types of heavy weight concrete. The mix proportions by weight; the steel- magnetite concrete was; steel punching, magnetite ore, 7.55 Portland cement and 4.53 water, resulting in a concrete mix of density 5.11 g/cm 3.In case of magnetite- basalt concrete, the mix proportions by weight was 41.6 magnetite ore, basalt ore, Portland cement and 7.45 water, resulting in a concrete mix of density 3.05 g/cm 3. The chemical compositions of steel- magnetite and magnetite basalt concrete under investigation are given in Table (1). (15). Table (1): Chemical Composition of and concretes Magnetite - Basalt Steel Magnetite Material % Material % Fe Fe2 O Fe2 O SiO SiO Al2 O CaO 7.26 CaO MgO 0.31 MgO 2.57 Al2 O Fe O 5.17 Fe O 1.37 Na O 0.54 Na O 0.34 K2O 0.37 SO Na2o 1.42 K2O 0.02 TiO H2 O 4.59 P2O MnO 0.03 SO H2 O

4 RESULTS AND DISCUSSION The neutrons and gamma rays distribution through the reactor core of the pressurized water reactor, and through the biological shields made of the (S-M) and (M-B) concretes, were modeled with the Mont Carlo Code, where the flux and dose values were calculated. The results of calculation of neutron and gamma ray fluxes and dose rates are presented in the form of tabulated data and attenuation relations for the for fluxes and dose rates. The obtained results are compared with those calculated by the same author in case of ordinary concrete (20). The total neutron and total gamma ray flux distribution for S-M, M- B and ordinary concretes for different thicknesses are presented in the form of tabulated data given in Tables (2 and 3). The calculated total neutrons, total gamma rays and total dose rates at various shielding layers of S-M, M- B and ordinary concrete are given in Tables (4, 5 and 6). Table (2): Total neutron fluxes calculated behind different thicknesses of Steel-Magnetite, and ordinary concrete (n/cm 2.s) (n/cm 2.s) Ordinary (20) (n/cm 2.s) E E E E E E E E E E E E E E E E E E12 Table (3): Total gamma ray fluxes for Steel-Magnetite, and ordinary concretes (#/cm 2.s) (#/cm 2.s) Ordinary (20) (#/cm 2.s) E E E E E E E E E E E E E E E Table (4): Total neutron dose rates for Steel-Magnetite, and ordinary s Ordinary (20) E E E E E E E E E E E E E E E E E E7 154

5 Table (5): Total gamma dose rates for Steel-Magnetite, and ordinary s Ordinary (20) ( E E E E E E E E E E E E E E E E E E6 Table (6): Total dose rates (N+G) for Steel-Magnetite, and ordinary s Ordinary (20) E E E E E E E E E E E E E E E E E E7 The flux and dose rate distributions of total neutrons, total gamma-rays and total neutrons+ total gamma-rays in S-M, M-B and ordinary concretes are shown in Figs. (2-6). The attenuation relations represented in these figures show that the flux and dose rate values decrease with increasing the shield thickness. They also show that the rate of depression in the flux and dose is quite high at the first layer of the biological shield for both S-M and M-B as well for. The presented attenuation relations given in these figures also show that S-M is the best attenuator and is the poorest attenuator. However, for penetration above 150 cm the difference is to some extent quite observable. These phenomena are observed for both total neutrons and total gamma- rays flux and dose rate distributions. Ne Neutron flux (n/cm 2.s) S-M concrete M-B concrete Distance Fig. 2. Attenuation of total neutron fluxes in S-M, M-B and O.s 155

6 Total gamma flux (#/cm 2.s) S-M concrete M-B cincrete Distance (cm Fig. 3 Attenuation of total gamma-ray fluxes in S-M, M-B and O.s Neutron dose rate 1E11 1E10 1E9 1E8 1E7 S-M concrete M-B cincrete Distance Fig. 4. Attenuation of total neutron doses in S-M, M-B and s Total gamma dose rate (ms//h) S-M concrete M-B cincrete 10 5 Distance Fig. 5. Attenuation of total gamma- ray doses in S-M, M-B and s 156

7 Total dose rate S-M cconcrete M-B cincrete 10 7 Distance Fig. 6 Attenuation of total doses due to neutrons and gamma-rays in S-M, M-B and s CONCLUSIONS The obtained calculated results for S-M and M-B concretes as well as the results calculated by the same author for ordinary concrete are given and discussed before. However, the following conclusions can be derived, 1. The flux and dose rate values for both neutron and gamma rays decreases with increasing the shield layer thickness. 2. For the neutron and total gamma ray attenuation, the steel- magnetite concrete used in the present work ( = 5.11 g/cm 3 ) is a better attenuator than the basalt- magnetite concrete ( = 3.05 g/cm 3 ) and ordinary concrete ( = 2.3 g/cm 3 ). 3. The steel- magnetite concrete is a more affective biological shield than the magnetite-basalt concrete. It attenuates the neutron and total gamma ray fluxes and dose rates at a higher rate especially at the inner layers of the biological shield. However, at deeper penetrations at a thickness > 200cm, no notable differences were observed in values of fluxes, dose rates calculated for all the three investigated concretes. 4. In addition, calculations have to be performed on biological shields made from heat resistant concrete which can be used to construct the inner layers of biological shield behind the pressure vessel. Further studies should also be performed to study heavy weight concretes suitable to construct biological shield surrounding primary coolant circuit, steam generator and nuclear fuel bonds of nuclear power stations. Acknowledgment The author gratefully acknowledges the help of Dr. M. A- Sarraf composition data for this work. in the experimental REFERENCES (1) Ali Basheer Azeez, Kahtan S. Mohammed, Mohd Mustafa Al Bakri Abdullah, Kamarudin Hussin, Andrei Victor Sandu and Rafiza Abdul Razak, " The Effect of Various Waste Materials Contents on the Attenuation Level of Anti- Radiation ing ", Materials, 6, ,(2013). (2) I. Akkurt, B. Mavi, A. Akkurt, C Basyigit, S. Kilincarslan, H, Yalim, "Study on Z-dependence of partial and total mass attenuation coefficients". J. Quant. Spectrosc. Radiat. Transf., 94, , (2005). (3) M. F. Kallan, " Radiation ing"; Longman Scientific and Technical: New York, NY, USA, (1989). 157

8 (4) M.H.Kharita, "Development of special radiation shielding concretes using natural local materials and evaluation of their shielding characteristics". Prog. Nucl. Energy, 50, 33 36, (2008). (5) Y. Song Lee, M. Ann Daugherty, and J.Donald Broton, "Assessing Aggregates for Radiation ing ", International, May, (2013). (6) Osman Gencel, Witold Brostow, Gonzalo Martinez- Barrera, Mustafa Sabri Goh, "Mechanical Properties of Polyymer Containing Different Amount of Hematite or Colemanite", Polimery Nauka Przemysl, 57, nr4, wrzesnia, (2012). (7) E. A. Bobadilla-Sanchez, G.Martinez-Barrera, W. Brostow, T. Datashvili: Express Polym. Lett., 3, 615, (2009). (8) M. Muthukumar, D. Mohan: Eur. Polymer j., 40, 2167, (2004). (9) Osman Gencel, Witold Brostow, Gengiz Ozel, Vmumin Filiz, " Containing Hematite Use as ing Barriers", ISSN Materials Science (Medziagotyra). Vol. 16, No. 3. (2010). (10) A. R. Bunsell, J. Renard, " Fundamental of Fibre Reinforced Composite Materials", Institute of Physics, Boston- Philadelphia, (2005). (11) A. El- Sayed Abdou, W. A. Kansouh and R. M- Megahed, Investigation of Radiation Attenuation Properties for Baryte, the Japan Society of Applied Physics, Cairo, Egypt, (2002). (12) M. F. Kaplan, Radiation ing, Gohn Wiley & Sons, New York, (1989). (13) A. S. Makarious, I. I. Bashter, A. El- Sayed, M. Samir Abdel Azim, and W. A- Kansouh, On the Utilization of Heavy For Radiation ing, Egypt, (1997) (14) A. B. Chilton, J. K. Shultis, and R.E. Faw, Principles of Radiation ing, Prentice- Hall, Englewood Chiffs, New Jersey, (1984). (15) I. I. Bashter, A. EL- Sayed Abdo and M. Samir Abdel- Azim, Magnetite Ores with Steel or Basalt for Radiation ing, Jon. J. Appl. Phys. Vol. 36, , June, (1997). (16) ing Materials, ed. R.G. Jaeger, "Engineering Compendium on Radiation ing", Vol. II., Springer- Verlag, Berlin, (1975). (17) J. R. Lamarsh, Introduction to Nuclear Engineering, Addision Wesley, (1975). (18) R. A. Forster, R. C. Little, J. F. Briesmeister, and J. S. Hendricks, MCNP Capabilities for Nuclear Well Logging Calculations, IEEE Transactions on Nuclear Science, 37 (3), 1378, June, (1990). (19) J. K. Shultis and R. E. Faw, "An MCNP Primer", Dept. of Mechanical and Nuclear Engineering, Kansas State University, Manhattan, (2011). (20) Amal A. El- Sawy, Development of Nuclear Power Reactor ing and Radiological Safety of Workers, Journal of Environmental Science, (2012). 158