Shielding Design to Obtain Compact
|
|
- Blake Hawkins
- 6 years ago
- Views:
Transcription
1 Journal of NUCLEAR SCIENCE and TECHNOLOGY, 31[6], pp. 510,-520 (June 1994). Shielding Design to Obtain Compact Marine Reactor Akio YAMAJI and Kiyoshi SAKO Tokai Research Establishment, Japan Atomic Energy Research Institute* (Received July 7, 1993), (Revised November 9, 1993) The marine reactors equipped in previously constructed nuclear ships are in need of the secondary shield which is installed outside the containment vessel. Most of the weight and volume of the reactor plants are occupied by this secondary shield. An advanced marine reactor called MRX (Marine Reactor X) has been designed to obtain a more compact and lightweight marine reactor with enhanced safety. The MRX is a new type of marine reactor which is an integral PWR (The steam generator is installed in the pressure vessel.) with adopting a waterfilled containment vessel and a new shielding design method of no installation of the secondary shield. As a result, MRX is considerably lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships. For instance, the plant weight and volume of the containment vessel of MRX are about 50% and 70% of those of the Nuclear Ship MUTSU, in spite of the power of MRX is 2.8 times as large as the MUTSU's reactor. The shielding design calculation was made using the ANISN, DOT3,5, QAD-CGGP2 and ORIGEN codes. The computational accuracy was confirmed by experimental analyses. KEYWORDS: ship propulsion reactors, shields, shielding, design, computer calculations, computer codes, weight, volume, nuclear ships I. INTRODUCTION A marine reactor should be compact and lightweight since it has to be installed in narrow and limited space in a ship, and also for economical viewpoint of a ship. The reactor type of previously constructed nuclear ships SAVANNAH and MUTSU is a conventional pressurized water reactor (PWR), and that of the nuclear ship OTTO HAHN an integral PWR**. These reactors are in need of the primary shield installed around the pressure vessel and the secondary shield outside the containment vessel. On the PWR of nuclear ships SAVANNAH and MUTSU, the secondary shield is necessary during normal reactor operation and in the time of reactor shutdown, since the primary coolant loop is installed outside the primary shield, and also necessary in the event of an accident with a consequential release of radioactive fission products from the core into the containment vessel. On the integral PWR of nuclear ship OTTO HAHN, the main purpose of the secondary shield is to prevent the crew and others from radiation during the time of an accident. The secondary shield of nuclear ships SAVANNAH, OTTO HAHN and MUTSU occupies the major part of the whole shieldo)-(8). Most of the weight of the marine reactors is due to the secondary shield. For instance, in the case of the Nuclear Ship MUTSU (N. S. MUTSU), the weight of the secondary shield is 88% of the whole shield, and the total shield weight of the N. S. MUTSU exceeds 70% of that of the reactor plant. To obtain a more compact and lightweight * Tokai-mura, Ibaraki-ken ** The steam generator of integral PWR is in - stalled in the pressure vessel. There is no large primary coolant pipe outside the pressure vessel
2 Vol. 31, No. 6 (June 1994) 511 marine reactor with enhanced safety, a design study is being performed on an advanced marine reactor called MRX (Marine Reactor X). A view of MRX is shown in Fig. 1. The following features of MRX design can be identified as specific(4)-(6) : Integral Fig. 1 Conceptual view of MRX PWR -El iminates the possibility of a large LOCA, simplifies the engineered safety system and makes a reactor plant compact. In-pressure vessel type control rod drive mechanisms -Eli minate the possibility of "control rod ejection" accident and simplify the engineered safety system. Water-filled containment vessel assively maintains core -P flooding, and is of great advantage to make a compact reactor plant. Passive decay heat removal system implifies the engineered safety -S system. The water in the containment vessel is useful for, (1) attenuation of radiation transmitted through the pressure vessel and/or generated in the containment vessel, (2) prevention of excessive radiation streaming through shield penetration such as an air gap around the pressure vessel, and (3) shielding in the case of an accident with a consequential release of radioactive fission products from the core into the containment vessel. Due to the adoption of the integral PWR and water-filled containment vessel, it becomes possible to keep the exposure dose to the crew or others on board or in the vicinity of the ship within a permissible level without any bulk shield outside the containment vessel during normal reactor operation, in the time of reactor shutdown and in the event of a hypothetical accident, which means the realization of a reactor which is not only small in size but also lightweight. A new concept, to design without secondary shield, was introduced into the shielding design of MRX. Namely, MRX has been structured so as to satisfy the design criteria without any bulk shield outside the containment vessel. As a consequence, the size of the containment vessel should also be designed so as to satisfy the shielding requirement. Figure 2 shows an example of a two-mrx Fig. 2 Two-MRX equipped icebreaker for use as a scientific observation ship 15
3 512 J. Nucl. Sci. Technol., equipped icebreaker for scientific observations. An icebreaker is used as it is the most conceivable nuclear ship to be equipped with advanced marine reactors in the near future. So for the time being, MRX is designed for testing on an icebreaker. This paper is organized as follows : Chap. II shows the design criteria with the setting conditions. In Chap. III the reactor and shield are described. Chap. IV shows the shielding analyses during normal reactor operation, in the time of reactor shutdown and in the event of a hypothetical accident. The computational accuracy is also described in Chap. IV. The conclusion is presented in Chap. V. II. DESIGN CRITERIA The specifications for the radiation levels on board are as follows : (a) The reactor room outside the containment vessel is classified as the controlled area. The maximum permissible design dose rate equivalent is specified as 10 msv/h for 48 working hours per week. Working in long hours is possible in this room with a satisfactory low dose equivalent. (b) In the double bottom under the containment vessel, which is classified as the controlled area, the maximum permissible dose rate equivalent is specified as 50 msv/h for 10 working hours per week. (c) In the engine room, which is classified as the surveillance area, the maximum permissible design dose rate equivalent is specified as 6 msv/h for 50 working hours per week. The criteria (a), (b) and (c) are set based on the Japanese regulations. (d) In the accommodation space, which is classified as the uncontrolled area, the dose rate equivalent has to be lower than ms/h*, resulting in 50 msv/yr for an unlimited stay in this area ; namely, staying for 365 x 24 h/yr. This criterion is based on the guidelines for the safety examination of the Nuclear Safety Commission of Japan. (e) Activation of the bottom shell due to neutrons from the core is restricted to a value under 3.7x 10-2 Bq/g(7) which is equal to the maximum radioactivity of steel used in nonnuclear facilities. (f) Long work hours have to be considered in the containment vessel at a reasonable time after the reactor has been shut down. (g) Repair work has to be possible for the ship's bottom with an unlimited stay after the reactor has been shut down. (h) It has to protect the reactor control room and wheel house from radiation during the time of an accident with a consequential release of radioactive fission products from the core into the containment vessel, and limited operation should be possible in the control room and the wheel house during the accident. (i) On the basis of an expected 20- year plant life, a maximum fast-neutron irradiation of 1019 nvt (E>=1 MeV) is used as the fast-neutron exposure limit of the pressure vessel. III DESCRIPTION OF REACTOR AND SHIELD The reactor generates 100 MWt with UO2 fuel enriched 4.3 wt% on the average. The active fuel length and equivalent lateral core diameter are 140 and 149 cm, respectively. The outer diameter of the pressure vessel is 400 cm. Table 1 shows the major specifications of MRX. The shielding system is designed to satisfy the design criteria during normal operation, in the time of reactor shutdown and in the event of a hypothetical accident, with consideration given to minimum weight, size and cost. For use of an integral PWR, it must be considered that the 16N and 17N-activation of secondary water due to the reactions 16O(n, p)16n and 17O(n, p)17n will occur in the steam generator, because of the very small distance existing between the core and steam generator as shown in Fig. 1, and may raise the dose rate equivalent around the secondary loop in the engine and reactor rooms. This activation must be kept sufficiently low to maintain the dose rate equivalents within the design criteria of these rooms. To satisfy this condition, an iron shield is installed between the core and steam generator, since * The value of msv/h is 1/50 of that of the N.S. MUTSU. 16
4 Vol. 31, No. 6 (June 1994) 513 Table 1 Major specifications of MRX only neutrons with an energy above 10 MeV can produce the above mentioned reactions, and iron has an excellent shielding ability against these high energy neutrons. This iron shield also has a role, (1) to decrease the g-ray flux of the diagonally upward direction, and to keep the dose rate equivalents at points D and E (shown in Fig. 1) on the outer-surface of the containment vessel within a permissible level and to maintain a low exposure dose during the work at point G (shown in Fig. 1) in the containment vessel for a reasonable time after the reactor has been shut down, and (2) to reduce the generation of radioactive corrosion products in the secondary water in the steam generator and to maintain the dose rate equivalent around the secondary loop in the engine and reactor rooms within a permissible level. The containment vessel is filled with water in which the cast steel shield is arranged. The outer diameter and height of the containment vessel are 7 and 13.3 m, respectively. The pressure vessel is immersed in water in the containment vessel. Figure 3 shows a schematic section in the horizontal plane at the core's center position. In the time of reactor shutdown, the water level in the containment vessel can be lowered for maintenance work. The thickness and arrangement of the cast steel shield, and the size of the containment vessel were determined based on shielding calculations by varying the arrangements systematically. The containment vessel can stand under high pressure at LOCA which was confirmed by analysis(6). Maintenance/repair work is capable under this structure(6). Fig. 3 Schematic section in horizontal plane of MRX's core center position (A case of a rectangular fuel assembly. Another case shows a hexagonal fuel assembly which is presented in Ref. (6).) There are an air gap and thermal insulator between the pressure vessel and water in the containment vessel. As described later, radiation streaming through this air gap does not reach any significant level, because of the water filled containment vessel and relatively large radius of the pressure vessel, and there are no other shield irregularities through which radiation streaming rises significantly. Around the pressure vessel, a steel shield 5 cm in thickness is placed to allow for space for the air gap and the thermal insulator, and to lower the exposure dose during repair work in the containment vessel after the reactor has been shut down. 17
5 514 J. Nucl. Sci. Technol., The position of the cast steel installed in water of the containment vessel was determined so as to effectively reduce secondary r-rays, to minimize shield weight and keep sufficient working space in the containment vessel. The water filled containment vessel serves also to reduce the exposure dose to the crew and others on board or in the vicinity of the ship in the event of an accident with a consequential release of radioactive fission products from the core into the containment vessel. The problems for the downwards are, (a) scattered radiation from the bottom shell and seawater getting into the rooms outside the containment vessel, (b) generation of induced activity in the structures outside the containment vessel, and (c) exposure during repair work on the ship's bottom after the reactor has been shut down. The space below the reactor, at the bottom of the ship, does not need to be completely shielded, since usually work in this space is not necessary during reactor operation. From this reason, it is possible to design for the downward direction with a relatively small distance between the core and containment vessel. MRX does not provide any bulk shield outside the containment vessel. As a result, MRX is lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships. group constant made by ANISN calculations with this library. The energy group structures used in ANISN and DOT calculations are shown in Table 2. Table 2 Energy group structures used in ANISN and DOT calculations IV. SHIELDING ANALYSIS The shielding design calculations were performed for normal reactor operation, reactor shutdown and a hypothetical accident, using the discrete ordinates codes ANISN(8) and DOT3.5(9), the point kernel code QAD- CGGP2(10) and the fission product generation code ORIGEN(11). The calculations were made with one-dimensional spherical and cylindrical geometries for ANISN, the twodimensional RZ geometry for DOT, and threedimensional combinatorial geometry for QAD. A data library, DLC-23E, was used for the calculations of ANISN. The DOT calculations were performed using the collapsed 18
6 Vol. 31, No. 6 (June 1994) 515 These discrete ordinates codes were used with P3 Legendre polynomial expansion and S16(ANISN calculations) and S48 (DOT calculations) angular quadrature sets. Evaluation of the computational accuracy is provided in this chapter. 1. Shielding Analysis during Normal Operation The conceptual shield structure and size of containment vessel were determined based on the calculations using the ANISN code with one-dimensional spherical geometries. Many ANISN calculations were made concerning the shielding arrangements of the radial, downward, and diagonally upward and downward directions because of importance of these directions as regards shielding design. Figure 1 shows the determined structure with the ANISN computational directions. The ANISN geometrical arrangements are shown in Table 3 along the radial and downward directions for the determined structure with the numbers of computational mesh points in each region, and Fig. 4 shows the calculated results along these directions. The ratios between dose rate equivalents of primary and secondary g-rays are 1/50 and 1/300 at the outer-surface of the pressure vessel and containment vessel in the radial direction. The calculated dose rate equi- Table 3 Shield sequence in radial and downward directions used in AN1SN calculation 19 --
7 516 J. Nucl. Sci. Technol., valents are presented in Table 4 at points A, B, C, D and E outside the containment vessel shown in Fig. 1. Fig. 4 ANISN calculation along radial and downward directions of MRX Table 4 ANISN calculated dose rate equivalent on outer-surface of containment vessel and in double bottom (msv/h), and ratio between DOT and ANISN calculated total dose rate equivalents The ANISN calculation in the downward direction shows that the maximum value of thermal neutron flux at the bottom shell is 2 x10-1 cm2/s during full power, from which the activation of the bottom shell is estimated to be 3 x 10-4 Bq/g two weeks after the reactor shutdown after continuous 20 year full power operation. This value is sufficiently small compared with the design criteria of 3.7 x 10-2 Bq/g described in Chap. II. The ANISN calculations in the radial, and diagonally upward and downward directions also show that the activation values in structures outside the containment vessel are under 1/100 of 3.7x10-2 Bq/g two weeks after the reactor shutdown after continuous 20 year full power operation. The DOT calculation was performed with fission neutrons and secondary g-rays to confirm the overall radiation level with this final structure. Figure 5 shows the computational geometry and the contours of the dose rate equivalent due to fission neutrons and secondary g-rays. This figure shows that no significant radiation streaming occurs in the containment vessel and the radiation levels are adequate at all positions around the containment vessel. Table 4 also shows the ratios between the DOT and ANISN values outside the containment vessel. The ANISN and DOT calculated values of the irradiation of fast neutrons are below 8 X 1015 nvt (E>= 1.11 MeV) at the innersurface of the pressure vessel for full power reactor operation for 20 continuous years. This value is sufficiently low compared with the design 20
8 Vol. 31, No. 6 (June 1994) 517 Fig. 5 Contour of dose rate equivalent due to fission neutrons and secondary -rays obtained by DOT RZ calculation (Sv/h at full power) g criteria of 1019 nvt (.E>=1 MeV) for radiation damage of the pressure vessel. This small value is due to the relatively large radius of the pressure vessel. For the calculations in diagonally upward direction, it is important to give consideration to fast neutron flux above 10 MeV, since these neutrons contribute to the generations of 16N g-rays and 17N neutrons in the secondary water in the steam generator. For this reason, another DOT calculation was made for the diagonally upward direction with a neutron energy structure of 14 groups above 10 MeV to obtain P(E) with a fine energy structure above 10 MeV in the steam generator. The calculations showed that a fast neutron flux above 10 MeV is below 106 n/cm2-5 at the undersurface of the steam generator. Using fast neutron fluxes for each energy group in the steam generator and the cross sections of 16N(n, p)16o and 17O(n, p)17n, the source densities due to 16N and 17N-decays were calculated in the high pressure steam, where the densities were assumed to be constant up to the turbine because of the high velocity of the steam. For the cross section of the 16O(n, p)16n reaction, the experimental data obtained by Martin("' was used, and for (n, p)17n reaction, the cross section data 17O in the Reactor Shielding Design Manual.(13). These activities will result in a dose rate equivalent of 2 x 10-5 msv/h at a distance of 30 cm from the steam pipeline, and even a small value at the turbine which are based on the QAD-CGGP2 calculations with the 16N -ray source density and ANISN one-dimen- g sional cylindrical calculations with the 17N neutron source density. At the hotwell, the dose rate equivalent will increase to 2 x 10-4 Sv/h at a distance of 30 cm due to mthe higher source density. The dose rate equivalent due to the corrosion products in the secondary loop is estimated to be sufficiently small in comparison with the design criteria of 10 msv/h at the surface of all components of the secondary system, where the estimation was made based on the calculated neutron flux in the steam generator and dose measurements in actual plants(14). 2. Shielding Analysis in Time of Reactor Shutdown To satisfy the design criteria, repair or maintenance work must be possible outside 21
9 518 J. Nucl. Sci. Technol., the pressure vessel, namely, in the containment vessel, in the double bottom, under the bottom shell, etc., 24 h after the reactor has been shut down. The fission products in the core were calculated using the ORIGEN code at 24 h after continuous reactor operation at full power to the maximum degree of burnup (25,000 MWD/t), and dose rate equivalents were calculated using the ANISN code with same geometrical condition in the calculations of normal reactor operation. The values are sufficiently low, namely (1) 5 x 10-6 msv/h and 1 x 10-6 msv/h at points F and G (shown in Fig. 1) in the containment vessel, (2) 3 x 10-9 Sv/h at point A on the outer-surface of m the containment vessel in the radial direction, (3) 7 x 10-8 msv/h at point C in the double bottom and (4) 3 x 10-9 msv/h at the under-surface of the bottom shell. The corrosion products in the primary coolant and deposited on the equipment of the primary system, and activation of the reactor components near the core are not evaluated precisely at the present designing stage, it seems however from the N. S. MUTSU shielding modification design calculation(14) that the dose rate equivalent due to these source terms does not become large enough to interfere with the work in containment vessel, since there are thick iron between these source positions and the working area in the containment vessel. From the above-mentioned radiation level due to the fission products in the core, and the discussion on the other source terms, it can be said that the maintenance and/or repair work in the containment vessel which are performed with a low level of water in the containment vessel are possible at a low exposure dose, and it is also possible to classify the area of seawater under the bottom shell into the uncontrolled area during the reactor shutdown. 3. Shielding Analysis for Accident For shielding design, it is necessary to evaluate the exposure dose equivalent of the crew and others due to direct g-rays and skyshine r-rays during a hypothetical accident with fission products being released into the containment vessel. The computational conditions are as follows : (1) The accident is assumed to occur immediately after continuous reactor operation at full power to the maximum degree of burn-up (25,000 MWD/t). (2) The ratios between the amount of fission products released into the containment vessel and those existing in the core before the release are assumed to be 100% for noble gas, 5% for iodine, 1% for halogen except for iodine and 0.01% for the other nuclei in the gaseous region in the containment vessel, on the other hand 50% for halogen and 1% for the other nuclei in the liquid region, these values were chosen based on the guidelines for the safety examination of the Nuclear Safety Commission of Japan. The isotope generation and depletion were calculated using the ORIGEN code. Thereafter, the dose rate equivalents on board and in the vicinity were calculated with the QAD- CGGP2 code using the above-mentioned source conditions in the containment vessel. The calculation shows the following which satisfies the Japanese regulations : (1) Installed a shield for the control room if necessary ; the exposure dose of crew can be restricted below 100 msv which is the limit of dose equivalent for an accident in which the crew are subjected to exposure. (2) For instance, if an unhabitation area is set within a distance of 200 m from the reactor, the dose equivalent at the boundary becomes 0.01 Sv which is sufficiently small compared with the dose limit of 0.25 Sv for public during the time of an accident. 4. Computational Accuracy Special attention was paid to evaluating of the accuracy of the calculation. The accuracy of calculations from the core to outside of the containment vessel was estimated by analyzing the primary shield tank experiment of the Nuclear Ship OTTO HAHN(15) by means of the ANISN code with the same conditions used in the design calculations, for order of Sn and Pi, mesh width, energy group structure and geometrical modeling method. A comparison between the experiment and calculation was made for the reac- 22
10 Vol. 31, No. 6 (June 1994) 519 tion rates of fast neutron detectors, epithermal and thermal neutron flux, and 7-ray dose rate equivalent. Figure 6 shows an example of the comparison between the measured and calculated values of 115In(n, n')115min reaction rate. Based on the experimental analysis, three times as large as the ANISN calculated dose rate equivalents were used as evaluated values at points A and C outer-surface of the containment vessel of MRX. The shielding design of MRX was made with considering this accuracy. Namely, the ANISN calculated dose rate equivalents on the outer-surface of the containment vessel are under 1/3 of the design criteria as shown in Table 4. V. CONCLUSION MRX has been designed to obtain a compact and lightweight marine reactor with enhanced safety. This was accomplished by adopting an integral PWR and a water-filled containment vessel. These adoptions enabled the authors to design without any bulk shield outside the containment vessel. A new concept, to design without secondary shield, was introduced into the shielding design of MRX. Namely, MRX has been structured so as to satisfy the design criteria without any bulk shield outside the containment vessel. As a result, MRX is considerably lighter in weight and more compact in size as compared with the reactors equipped in previously constructed nuclear ships SAVANNAH, OTTO HAHN and MUTSU. For instance, the plant weight and volume of the containment vessel of MRX are about 50% and 70% of those of the N.S. MUTSU, in spite of the power of MRX is 2.8 times as large as the MUTSU's reactor and the design criteria of dose rate equivalent in the accommodation space of MRX equipped nuclear ship is small as 1/50 of that of the N.S. MUTSU. The computational accuracy of the shielding design calculations was confirmed by experimental analyses. ACKNOWLEDGMENT Fig. 6 Comparison of 115In(n, n')115min reaction rate between ANISN calculation and experiment in shield tank of N.S. OTTO HAHN The analyses using the QAD code were already made by the authors for the following experiments, (1) shield mock-up tests for the first nuclear ship of Japan(16), (2) T- ray experiment using the LIDO nitrogen 16 facility at Harwell(17)(18) and (3) g-ray experiment for shield effect of ship structure(19). These analyses show an overestimation of the QAD values(20). In the MRX design, QAD calculated values were used without any corrections. The authors would like to thank Mr. H. Kawasaki and Mr. K. Ohuchi of CRC Research Inst. Inc. for assisting with the calculations. --REFERENCES- (1) SMITH, W. R., TURNER, M. A.: BA W , (1959). (2) JAEGER, R. G., et al. (ed.) : "Engineering Compendium on Reactor Shielding", Vol. III, (1970), Springer-Verlag. (3) YAMAJI, A., et el.: Proc. 6th Int. Conf. Radiation Shielding, Tokyo, 617 (1983). (4) SAKO, K., et al.: SMiRT 11 Trans., Vol. SD2, 357 (1991). (5) ISHIZAKA, Y., et al.: ibid., 363. (6) SAKO, K., et al.: Proc. Int. Conf. on Design and Safety of Advanced Nuclear Power Plants. Tokyo, (1992). 23
11 520 J. Nucl. Sci Technol., (7) Radioactivity : Recommendations of the International Commission on Radiological Units and Measurements, ICRU Rep. 10c, NBS Handbook 86, (1963). (8) ENGLE, Jr., W. W.: K-1693, (1967). (9) RHODES, W.A. : ORNL-TM-4280, (1973) ; NEA- CPL (1977). (10) SAKAMOTO, Y., TANAKA, S.: JAERI-M 90110, (1990). (11) BELL, M. J.: ORNL-4628, (1973), and KOYAMA, K., et al.: JAERI-M 6954, (1977). 2) MARTIN, B.C.: Phys. Rev., (1 93(3), 498 (1954). (13) ROCKWELL III, T. (Ed.) : "Reactor Shielding Design Manual", (1956), McGraw-Hill Book. (14) YAMAJI, A., et al.: To be published in JAERI- M. (15) FIEBIG, R., et al.: Atomkernenergie, 18(1), 71 (1971). (16) Japan Nucl. Ship Dev. Agency : JNS-4-1~6 (in Japanese), (1967). (17) BISHOP, G. B., SMITTON, C., PACKWOOD, A. : Ann. Nucl. Energy, 3, 65 (1976). ) SMITTON, C., BISHOP, G.B. (18 : J. Br. Nucl. Energy Soc., 14(1), 89 (1975). (19) YAMAKOSHI, H., UEKI, K., NAKATA, M.: J. Nucl. Sci. Technol., 20(2) 127 (1983). (20) YAMAJI, A., et al.: J. At. Energy Soc. Jpn., (in Japanese), 26(2), 139 (1984). 24
Nuclear Data for Emergency Preparedness of Nuclear Power Plants Evaluation of Radioactivity Inventory in PWR using JENDL 3.3
Nuclear Data for Emergency Preparedness of Nuclear Power Plants Evaluation of Radioactivity Inventory in PWR using JENDL 3.3 Yoshitaka Yoshida, Itsuro Kimura Institute of Nuclear Technology, Institute
More informationEstimation of Radioactivity and Residual Gamma-ray Dose around a Collimator at 3-GeV Proton Synchrotron Ring of J-PARC Facility
Estimation of Radioactivity and Residual Gamma-ray Dose around a Collimator at 3-GeV Proton Synchrotron Ring of J-PARC Facility Y. Nakane 1, H. Nakano 1, T. Abe 2, H. Nakashima 1 1 Center for Proton Accelerator
More informationReactor radiation skyshine calculations with TRIPOLI-4 code for Baikal-1 experiments
DOI: 10.15669/pnst.4.303 Progress in Nuclear Science and Technology Volume 4 (2014) pp. 303-307 ARTICLE Reactor radiation skyshine calculations with code for Baikal-1 experiments Yi-Kang Lee * Commissariat
More informationThree-dimensional RAMA Fluence Methodology Benchmarking. TransWare Enterprises Inc., 5450 Thornwood Dr., Suite M, San Jose, CA
Three-dimensional RAMA Fluence Methodology Benchmarking Steven P. Baker * 1, Robert G. Carter 2, Kenneth E. Watkins 1, Dean B. Jones 1 1 TransWare Enterprises Inc., 5450 Thornwood Dr., Suite M, San Jose,
More informationNeutronics Experiments for ITER at JAERI/FNS
Neutronics Experiments for ITER at JAERI/FNS C. Konno 1), F. Maekawa 1), Y. Kasugai 1), Y. Uno 1), J. Kaneko 1), T. Nishitani 1), M. Wada 2), Y. Ikeda 1), H. Takeuchi 1) 1) Japan Atomic Energy Research
More informationNeutron Dose near Spent Nuclear Fuel and HAW after the 2007 ICRP Recommendations
Neutron Dose near Spent Nuclear Fuel and HAW after the 2007 ICRP Recommendations Gunter Pretzsch Gesellschaft fuer Anlagen- und Reaktorsicherheit (GRS) mbh Radiation and Environmental Protection Division
More informationSome Comments to JSSTDL-300
Some Comments to -300 Chikara KONNO Center for Proton Accelerator Facilities Japan Atomic Energy Research Institute Tokai-mura Naka-gun Ibaraki-ken 39-95 JAPAN e-mail : konno@cens.tokai.jaeri.go.jp The
More informationChem 481 Lecture Material 4/22/09
Chem 481 Lecture Material 4/22/09 Nuclear Reactors Poisons The neutron population in an operating reactor is controlled by the use of poisons in the form of control rods. A poison is any substance that
More informationIrradiation Behaviors of Nuclear Grade Graphite in Commercial Reactor, (II)
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 22[3J, pp. 225-232 (March 1985). 225 TECHNICAL REPORT Irradiation Behaviors of Nuclear Grade Graphite in Commercial Reactor, (II) Thermal and Physical Properties
More informationARTICLE. Progress in Nuclear Science and Technology Volume 4 (2014) pp Yoshiko Harima a*, Naohiro Kurosawa b and Yukio Sakamoto c
DOI: 10.15669/pnst.4.548 Progress in Nuclear Science and Technology Volume 4 (2014) pp. 548-552 ARTICLE Parameter search of geometric-progression formula for gamma-ray isotropic point source buildup factors
More informationSOME ASPECTS OF COOLANT CHEMISTRY SAFETY REGULATIONS AT RUSSIA S NPP WITH FAST REACTORS
Federal Environmental, Industrial and Nuclear Supervision Service Scientific and Engineering Centre for Nuclear and Radiation Safety Scientific and Engineering Centre for Nuclear and Radiation Safety Member
More informationLesson 14: Reactivity Variations and Control
Lesson 14: Reactivity Variations and Control Reactivity Variations External, Internal Short-term Variations Reactivity Feedbacks Reactivity Coefficients and Safety Medium-term Variations Xe 135 Poisoning
More informationACTIVATION ANALYSIS OF DECOMISSIONING OPERATIONS FOR RESEARCH REACTORS
ACTIVATION ANALYSIS OF DECOMISSIONING OPERATIONS FOR RESEARCH REACTORS Hernán G. Meier, Martín Brizuela, Alexis R. A. Maître and Felipe Albornoz INVAP S.E. Comandante Luis Piedra Buena 4950, 8400 San Carlos
More informationIntegral Benchmark Experiments of the Japanese Evaluated Nuclear Data Library (JENDL)-3.3 for the Fusion Reactor Design
1 Integral Benchmark Experiments of the Japanese Evaluated Nuclear Data Library (JENDL)-3.3 for the Fusion Reactor Design T. Nishitani 1), K. Ochiai 1), F. Maekawa 1), K. Shibata 1), M. Wada 2), I. Murata
More informationBenchmark Tests of Gamma-Ray Production Data in JENDL-3 for Some Important Nuclides
Journal of NUCLEAR SCIENCE and TECFINOLOGY, 27[9], pp. 844~852 (September 1990). TECHNICAL REPORT Benchmark Tests of Gamma-Ray Production Data in JENDL-3 for Some Important Nuclides CAI Shao-huit, Akira
More informationRadiation Shielding of Extraction Absorbers for a Fermilab Photoinjector
Fermilab FERMILAB-TM-2220 August 2003 Radiation Shielding of Extraction Absorbers for a Fermilab Photoinjector I.L. Rakhno Fermilab, P.O. Box 500, Batavia, IL 60510, USA August 12, 2003 Abstract Results
More informationRecommendation on Decay Heat Power in Nuclear Reactorst
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 28[12], pp. 1134~1142 (December 1991). SUMMARY REPORT Recommendation on Decay Heat Power in Nuclear Reactorst Kanji TASAKAt1, Toshio KATOHt1, Junichi KATAKURAt2,
More informationProgress in Nuclear Science and Technology, Volume 6,
DOI: 1.15669/pnst.6 Progress in Nuclear Science and Technology Volume 6 (19) pp. 1-16 ARTICLE A study on calculation method of duct streaming from medical linac rooms Takuma Noto * Kazuaki Kosako and Takashi
More informationAP1000 European 15. Accident Analyses Design Control Document EVALUATION MODELS AND PARAMETERS FOR ANALYSIS OF RADIOLOGICAL CONSEQUENCES OF ACCIDENTS
APPENDIX 15A EVALUATION MODELS AND PARAMETERS FOR ANALYSIS OF RADIOLOGICAL CONSEQUENCES OF ACCIDENTS This appendix contains the parameters and models that form the basis of the radiological consequences
More informationRadioactive effluent releases from Rokkasho Reprocessing Plant (1) - Gaseous effluent -
Radioactive effluent releases from Rokkasho Reprocessing Plant (1) - Gaseous effluent - K.Anzai, S.Keta, M.Kano *, N.Ishihara, T.Moriyama, Y.Okamura K.Ogaki, K.Noda a a Reprocessing Business Division,
More informationTAGS and FP Decay Heat Calculations( ) -Impact on the LOCA Condition Decay Heat-
TAGS and FP Decay Heat Calculations( ) -Impact on the LOCA Condition Decay Heat- Akira HONMA, Tadashi YOSHIDA Musashi Institute of Technology, Tamazutsumi 1-28-1, Setagaya-ku, Tokyo 158-8557, Japan e-mail:
More informationDETERMINATION OF THE SERVICE LIFE FOR THE EXCORE NEUTRON DETECTOR CABLES IN SEABROOK STATION
DETERMINATION OF THE SERVICE LIFE FOR THE EXCORE NEUTRON DETECTOR CABLES IN SEABROOK STATION John R. White and Lee H. Bettenhausen Chemical and Nuclear Engineering Department University of Massachusetts-Lowell,
More informationDetermination of research reactor fuel burnup
Determination of research reactor fuel burnup INTERNATIONAL ATOMIC ENERGY AGENCY January 1992 DETERMINATION OF RESEARCH REACTOR FUEL BURNUP IAEA, VIENNA, 1992 IAEA-TECDOC-633 ISSN 1011-4289 Printed FOREWORD
More informationCritical Experiment Analyses by CHAPLET-3D Code in Two- and Three-Dimensional Core Models
Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 42, No. 1, p. 101 108 (January 2005) TECHNICAL REPORT Critical Experiment Analyses by CHAPLET-3D Code in Two- and Three-Dimensional Core Models Shinya KOSAKA
More informationMA/LLFP Transmutation Experiment Options in the Future Monju Core
MA/LLFP Transmutation Experiment Options in the Future Monju Core Akihiro KITANO 1, Hiroshi NISHI 1*, Junichi ISHIBASHI 1 and Mitsuaki YAMAOKA 2 1 International Cooperation and Technology Development Center,
More informationDetection of Xe135 at Nuclear Reactor of Unit 2, Fukushima Daiichi Nuclear Power Station. November 4, 2011 Tokyo Electric Power Company
Detection of Xe135 at Nuclear Reactor of Unit 2, Fukushima Daiichi Nuclear Power Station November 4, 2011 Tokyo Electric Power Company On November 1, as a sampling result by the gas control system that
More informationCONFORMITY BETWEEN LR0 MOCK UPS AND VVERS NPP PRV ATTENUATION
CONFORMITY BETWEEN LR MOCK UPS AND VVERS NPP PRV ATTENUATION D. Kirilova, K. Ilieva, S. Belousov Institute for Nuclear Research and Nuclear Energy, Bulgaria Email address of main author: desi.kirilova@gmail.com
More informationTRANSMUTATION OF CESIUM-135 WITH FAST REACTORS
TRANSMUTATION OF CESIUM-3 WITH FAST REACTORS Shigeo Ohki and Naoyuki Takaki O-arai Engineering Center Japan Nuclear Cycle Development Institute (JNC) 42, Narita-cho, O-arai-machi, Higashi-Ibaraki-gun,
More informationChemistry 500: Chemistry in Modern Living. Topic 5: The Fires of Nuclear Fission. Atomic Structure, Nuclear Fission and Fusion, and Nuclear.
Chemistry 500: Chemistry in Modern Living 1 Topic 5: The Fires of Nuclear Fission Atomic Structure, Nuclear Fission and Fusion, and Nuclear Weapons Chemistry in Context, 2 nd Edition: Chapter 8, Pages
More informationCALCULATION OF ISOTOPIC COMPOSITION DURING CONTINUOUS IRRADIATION AND SUBSEQUENT DECAY IN BIOLOGICAL SHIELD OF THE TRIGA MARK ΙΙ REACTOR
International Conference Nuclear Energy for New Europe 2002 Kranjska Gora, Slovenia, September 9-12, 2002 www.drustvo-js.si/gora2002 CALCULATION OF ISOTOPIC COMPOSITION DURING CONTINUOUS IRRADIATION AND
More informationNATURAL CONVECTION HEAT TRANSFER CHARACTERISTICS OF KUR FUEL ASSEMBLY DURING LOSS OF COOLANT ACCIDENT
NATURAL CONVECTION HEAT TRANSFER CHARACTERISTICS OF KUR FUEL ASSEMBLY DURING LOSS OF COOLANT ACCIDENT Ito D*, and Saito Y Research Reactor Institute Kyoto University 2-1010 Asashiro-nishi, Kumatori, Sennan,
More informationAP1000 European 11. Radioactive Waste Management Design Control Document
CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT 11.1 Source Terms This section addresses the sources of radioactivity that are treated by the liquid and gaseous radwaste systems. Radioactive materials are generated
More informationEVALUATION OF PWR AND BWR CALCULATIONAL BENCHMARKS FROM NUREG/CR-6115 USING THE TRANSFX NUCLEAR ANALYSIS SOFTWARE
ANS MC2015 - Joint International Conference on Mathematics and Computation (M&C), Supercomputing in Nuclear Applications (SNA) and the Monte Carlo (MC) Method Nashville, Tennessee April 19 23, 2015, on
More informationJOYO MK-III Performance Test at Low Power and Its Analysis
PHYSOR 200 -The Physics of Fuel Cycles and Advanced Nuclear Systems: Global Developments Chicago, Illinois, April 25-29, 200, on CD-ROM, American Nuclear Society, Lagrange Park, IL. (200) JOYO MK-III Performance
More informationStudy on Radiation Shielding Performance of Reinforced Concrete Wall (2): Shielding Analysis
20th International Conference on Structural Mechanics in Reactor Technology (SMiRT 20) Espoo, Finland, August 9-14, 2009 SMiRT 20-Division V, Paper 1865 Study on Radiation Shielding Performance of Reinforced
More informationEvaluation of Geometric Progression (GP) Buildup Factors using MCNP Codes (MCNP6.1 and MCNP5-1.60)
EPJ Web of Conferences 106, 03009 (2016) DOI: 10.1051/epjconf/201610603009 C Owned by the authors, published by EDP Sciences, 2016 Evaluation of Geometric Progression (GP) Buildup Factors using MCNP Codes
More informationError Estimation for ADS Nuclear Properties by using Nuclear Data Covariances
Error Estimation for ADS Nuclear Properties by using Nuclear Data Covariances Kasufumi TSUJIMOTO Center for Proton Accelerator Facilities, Japan Atomic Energy Research Institute Tokai-mura, Naka-gun, Ibaraki-ken
More informationExtension of the MCBEND Monte Carlo Code to Perform Adjoint Calculations using Point Energy Data
Extension of the MCBEND Monte Carlo Code to Perform Adjoint Calculations using Point Energy Data Malcolm Grimstone Abstract In radiation transport calculations there are many situations where the adjoint
More informationAccelerator Facility Accident Report
Accelerator Facility Accident Report 31 May 2013 Incorporated Administrative Agency - Japan Atomic Energy Agency Inter-University Research Institute - High Energy Accelerator Research Organization Subject:
More information11. Radioactive Waste Management AP1000 Design Control Document
CHAPTER 11 RADIOACTIVE WASTE MANAGEMENT 11.1 Source Terms This section addresses the sources of radioactivity that are treated by the liquid and gaseous radwaste systems. Radioactive materials are generated
More informationCRITICALITY DETECTION METHOD BASED ON FP GAMMA RADIATION MEASUREMENT
CRITICALITY DETECTION METHOD BASED ON FP GAMMA RADIATION MEASREMENT Yoshitaka Naito, Kazuo Azekura NAIS Co., inc. Muramatsu 416, Tokaimura, Ibaraki-ken, Japan 319-1112 ynaito@nais.ne.jp azekura@nais.ne.jp
More informationComparison of 2 Lead-Bismuth Spallation Neutron Targets
Comparison of 2 Lead-Bismuth Spallation Neutron Targets Keith Woloshun, Curtt Ammerman, Xiaoyi He, Michael James, Ning Li, Valentina Tcharnotskaia, Steve Wender Los Alamos National Laboratory P.O. Box
More informationChapter 2 HEAT CONDUCTION EQUATION
Heat and Mass Transfer: Fundamentals & Applications 5th Edition in SI Units Yunus A. Çengel, Afshin J. Ghajar McGraw-Hill, 2015 Chapter 2 HEAT CONDUCTION EQUATION Mehmet Kanoglu University of Gaziantep
More informationStructural Health Monitoring of Nuclear Power Plants using Inverse Analysis in Measurements
Structural Health Monitoring of Nuclear Power Plants using Inverse Analysis in Measurements Fumio Kojima Organization of Advanced Science and Technology, Kobe University 1-1, Rokkodai, Nada-ku Kobe 657-8501
More informationChapter 2 HEAT CONDUCTION EQUATION
Heat and Mass Transfer: Fundamentals & Applications Fourth Edition Yunus A. Cengel, Afshin J. Ghajar McGraw-Hill, 2011 Chapter 2 HEAT CONDUCTION EQUATION Mehmet Kanoglu University of Gaziantep Copyright
More informationThe Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit
The Dynamical Loading of the WWER440/V213 Reactor Pressure Vessel Internals during LOCA Accident in Hot and Cold Leg of the Primary Circuit ABSTRACT Peter Hermansky, Marian Krajčovič VUJE, Inc. Okružná
More informationOptimum Arrangement to Minimize Total Dose Rate of Iron-Polyethylene Shielding System
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 26[4], pp. 411~421 (April 1989). 411 Optimum Arrangement to Minimize Total Dose Rate of Iron-Polyethylene Shielding System Kohtaro UEKI and Yoshihito NAMITO Nuclear
More informationVerification of Core Monitoring System with Gamma Thermometer
GENES4/ANP2003, Sep. 15-19, 2003, Kyoto, JAPAN Paper 1006 Verification of Core Monitoring System with Gamma Thermometer Hisashi SHIRAGA 1*, Hiromi MARUYAMA 1, Atsushi FUSHIMI 2, Yoshiji KARINO 3 and Hiroyuki
More informationEvaluating the Safety of Digital Instrumentation and Control Systems in Nuclear Power Plants
Evaluating the Safety of Digital Instrumentation and Control Systems in Nuclear Power Plants John Thomas With many thanks to Francisco Lemos for the nuclear expertise provided! System Studied: Generic
More informationOctober 2017 November Employees Contractors
Attachment Exposure Dose Distribution 1. Effective Dose from External Exposure Table 1 shows the distribution of external exposure dose of workers who were involved in radiation work at the Fukushima Daiichi
More informationNuclear Fission and Fusion A. Nuclear Fission. The process of splitting up of the nucleus of a heavy atom into two nuclei more or less of equal fragments when bombarded with neutron simultaneously releasing
More informationMCNP CALCULATION OF NEUTRON SHIELDING FOR RBMK-1500 SPENT NUCLEAR FUEL CONTAINERS SAFETY ASSESMENT
MCNP CALCULATION OF NEUTRON SHIELDING FOR RBMK-15 SPENT NUCLEAR FUEL CONTAINERS SAFETY ASSESMENT R. Plukienė 1), A. Plukis 1), V. Remeikis 1) and D. Ridikas 2) 1) Institute of Physics, Savanorių 231, LT-23
More informationFuel Element Burnup Determination in HEU - LEU Mixed TRIGA Research Reactor Core
Fuel Element Burnup Determination in HEU - LEU Mixed TRIGA Research Reactor Core Tomaž Žagar, Matjaž Ravnik Institute "Jožef Stefan", Jamova 39, Ljubljana, Slovenia Tomaz.Zagar@ijs.si Abstract This paper
More informationNuclear Energy ECEG-4405
Nuclear Energy ECEG-4405 Today s Discussion Technical History and Developments Atom Nuclear Energy concepts and Terms Features Fission Critical Mass Uranium Fission Nuclear Fusion and Fission Fusion Fission
More informationHTR Spherical Super Lattice Model For Equilibrium Fuel Cycle Analysis. Gray S. Chang. September 12-15, 2005
INEEL/CON-05-02655, Revision 3 PREPRINT HTR Spherical Super Lattice Model For Equilibrium Fuel Cycle Analysis Gray S. Chang September 12-15, 2005 Mathematics And Computation, Supercomputing, Reactor Physics
More information22.06 ENGINEERING OF NUCLEAR SYSTEMS OPEN BOOK FINAL EXAM 3 HOURS
22.6 ENGINEERING OF NUCLEAR SYSTEMS OPEN BOOK FINAL EXAM 3 HOURS Short Questions (1% each) a) The specific power in a UO 2 pellet of a certain LWR is q"'=2 W/cm 3. The fuel 235 U enrichment is 4 % by weight.
More informationBenchmark Test of JENDL High Energy File with MCNP
Benchmark Test of JENDL High Energy File with MCNP Masayuki WADA, Fujio MAEKAWA, Chikara KONNO Intense Neutron Source Laboratory, Department of Materials Science Japan Atomic Energy Research Institute,
More informationNuclear Fission. Q for 238 U + n 239 U is 4.??? MeV. E A for 239 U 6.6 MeV MeV neutrons are needed.
Q for 235 U + n 236 U is 6.54478 MeV. Table 13.11 in Krane: Activation energy E A for 236 U 6.2 MeV (Liquid drop + shell) 235 U can be fissioned with zero-energy neutrons. Q for 238 U + n 239 U is 4.???
More informationComparison of assessment of neutron fluence affecting VVER 440 reactor pressure vessel using DORT and TORT codes
Comparison of assessment of neutron fluence affecting VVER 440 reactor pressure vessel using DORT and TORT codes P. Montero Department of Neutronics, Research Center Rez, Cz International Conference on
More informationAvailable online at ScienceDirect. Energy Procedia 71 (2015 )
Available online at www.sciencedirect.com ScienceDirect Energy Procedia 71 (2015 ) 97 105 The Fourth International Symposium on Innovative Nuclear Energy Systems, INES-4 High-Safety Fast Reactor Core Concepts
More informationThe Physics of Nuclear Reactors. Heather King Physics 420
The Physics of Nuclear Reactors Heather King Physics 420 Nuclear Reactions A nuclear reaction is a reaction that involves atomic nuclei, or nuclear particles (protons, neutrons), producing products different
More informationLectures on Applied Reactor Technology and Nuclear Power Safety. Lecture No 6
Lectures on Nuclear Power Safety Lecture No 6 Title: Introduction to Thermal-Hydraulic Analysis of Nuclear Reactor Cores Department of Energy Technology KTH Spring 2005 Slide No 1 Outline of the Lecture
More information(1) SCK CEN, Boeretang 200, B-2400 Mol, Belgium (2) Belgonucléaire, Av. Arianelaan 4, B-1200 Brussels, Belgium
The REBUS Experimental Programme for Burn-up Credit Peter Baeten (1)*, Pierre D'hondt (1), Leo Sannen (1), Daniel Marloye (2), Benoit Lance (2), Alfred Renard (2), Jacques Basselier (2) (1) SCK CEN, Boeretang
More informationTEPCO s Activities on the Investigation into Unsolved Issues in the Fukushima Daiichi NPS Accident
2015 The Tokyo Electric Power Company, INC. All Rights Reserved. 0 TEPCO s Activities on the Investigation into Unsolved Issues in the Fukushima Daiichi NPS Accident IAEA IEM8 Vienna International Centre,
More informationFP release behavior at Unit-2 estimated from CAMS readings on March 14 th and 15 th
Attachment 2-11 FP release behavior at Unit-2 estimated from CAMS readings on March 14 th and 15 th 1. Outline of the incident and the issue to be examined At Unit-2 of the Fukushima Daiichi NPS, the reactor
More informationAdaptation of Pb-Bi Cooled, Metal Fuel Subcritical Reactor for Use with a Tokamak Fusion Neutron Source
Adaptation of Pb-Bi Cooled, Metal Fuel Subcritical Reactor for Use with a Tokamak Fusion Neutron Source E. Hoffman, W. Stacey, G. Kessler, D. Ulevich, J. Mandrekas, A. Mauer, C. Kirby, D. Stopp, J. Noble
More informationYear 11 Physics booklet Topic 1 Atomic structure and radioactivity Name:
Year 11 Physics booklet Topic 1 Atomic structure and radioactivity Name: Atomic structure and radioactivity Give a definition for each of these key words: Atom Isotope Proton Neutron Electron Atomic nucleus
More informationEmployees Contractors
Attachment Exposure Dose Distribution 1. Effective Dose from External Exposure Table 1 shows the distribution of external exposure dose of workers who were involved in radiation work at the Fukushima Daiichi
More informationRadiation Damage Effects in Solids. Los Alamos National Laboratory. Materials Science & Technology Division
Radiation Damage Effects in Solids Kurt Sickafus Los Alamos National Laboratory Materials Science & Technology Division Los Alamos, NM Acknowledgements: Yuri Osetsky, Stuart Maloy, Roger Smith, Scott Lillard,
More informationIntroduction. Problem Summary. Attila Code-to-Code Comparison Subcritical Spent Nuclear Fuel Canister with Primary Neutron and Gamma Sources
Introduction An example spent nuclear fuel canister calculation was performed with Attila, with results compared to the MCNPX Monte Carlo code. The canister contained representative subcritical neutron
More informationTRAINING IN EXTERNAL DOSIMETRY CALCULATIONS WITH COMPUTATIONAL CODES
TRAINING IN EXTERNAL DOSIMETRY CALCULATIONS WITH COMPUTATIONAL CODES S. MORATÓ, A.BERNAL, A. QUEROL, A. ABARCA, C. GÓMEZ-ZARZUELA, R.MIRÓ, G.VERDÚ Institute for Industrial, Radiophysical and Environmental
More informationReview of the primary coolant chemistry at NPP Temelín and its impact on the fuel cladding
Review of the primary coolant chemistry at NPP Temelín and its impact on the fuel cladding M. Mikloš, K. Vonková, J. Kysela Research Centre Řez Ltd, 250 68 Řež, Czech Republic D. Ernst NPP Temelín, Reactor
More informationStudy of irradiation effects in materials with high-neutron-flux fission reactors
Study of irradiation effects in materials with high-neutron-flux fission reactors Tatsuo Shikama Institute for Materials Research, Tohoku University 2-1-1 Katahira, Aobaku, Sendai, 980-8577 Japan, shikama@imr.tohoku.ac.jp
More informationA Practical Method for Evaluating the Neutron Dose
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 3[11], p.473~478 (November, 1966) 473 A Practical Method for Evaluating the Neutron Dose Equivalent Rate* Yoshikazu YOSHIDA**, Hatsumi TATSUTA**, Hiroshi RYUFUKU**,
More informationNEUTRON ACTIVATION OF REACTOR COMPONENTS DURING OPERATION LIFETIME OF A NPP
NEUTRON ACTIVATION OF REACTOR COMPONENTS DURING OPERATION LIFETIME OF A NPP G. Pretzsch, B. Gmal, U. Hesse Gesellschaft für Anlagen- und Reaktorsicherheit (GRS) mbh, Germany Email address of main author:
More informationEXPERIENCE OF TEST OPERATION FOR REMOVAL OF FISSION PRODUCT NUCLIDES IN TRU-LIQUID WASTE AND CONCENTRATED NITRIC ACID USING INORGANIC ION EXCHANGERS
EXPERIENCE OF TEST OPERATION FOR REMOVAL OF FISSION PRODUCT NUCLIDES IN TRU-LIQUID WASTE AND CONCENTRATED NITRIC ACID USING INORGANIC ION EXCHANGERS ABSTRACT H. Tajiri, T. Mimori, K. Miyajima, T. Uchikoshi
More informationEstimation of accidental environmental release based on containment measurements
Estimation of accidental environmental release based on containment measurements Péter Szántó, Sándor Deme, Edit Láng, Istvan Németh, Tamás Pázmándi Hungarian Academy of Sciences Centre for Energy Research,
More informationInvariability of neutron flux in internal and external irradiation sites in MNSRs for both HEU and LEU cores
Indian Journal of Pure & Applied Physics Vol. 49, February 2011, pp. 83-90 Invariability of neutron flux in internal and external irradiation sites in MNSRs for both HEU and LEU cores M Albarhoum Department
More informationIntroduction. Neutron Effects NSEU. Neutron Testing Basics User Requirements Conclusions
Introduction Neutron Effects Displacement Damage NSEU Total Ionizing Dose Neutron Testing Basics User Requirements Conclusions 1 Neutron Effects: Displacement Damage Neutrons lose their energy in semiconducting
More information12 Moderator And Moderator System
12 Moderator And Moderator System 12.1 Introduction Nuclear fuel produces heat by fission. In the fission process, fissile atoms split after absorbing slow neutrons. This releases fast neutrons and generates
More informationA Method For the Burnup Analysis of Power Reactors in Equilibrium Operation Cycles
Journal of NUCLEAR SCIENCE and TECHNOLOGY, 3[5], p.184~188 (May 1966). A Method For the Burnup Analysis of Power Reactors in Equilibrium Operation Cycles Shoichiro NAKAMURA* Received February 7, 1966 This
More informationEmployees Contractors
Attachment Exposure Dose Distribution 1. Effective Dose from External Exposure Table 1 shows the distribution of external exposure dose of workers who were involved in radiation work at the Fukushima Daiichi
More informationResearch Article Calculation of Radioactivity and Dose Rate of Activated Corrosion Products in Water-Cooled Fusion Reactor
Science and Technology of Nuclear Installations Volume 2016, Article ID 6051834, 6 pages http://dx.doi.org/10.1155/2016/6051834 Research Article Calculation of Radioactivity and Dose Rate of Activated
More informationMCNP neutron streaming investigations from the reactor core to regions outside the reactor pressure vessel for a Swiss PWR
DOI: 10.15669/pnst.4.481 Progress in Nuclear Science and Technology Volume 4 (2014) pp. 481-485 ARTICLE MCNP neutron streaming investigations from the reactor core to regions outside the reactor pressure
More informationThe basic structure of an atom is a positively charged nucleus composed of both protons and neutrons surrounded by negatively charged electrons.
4.4 Atomic structure Ionising radiation is hazardous but can be very useful. Although radioactivity was discovered over a century ago, it took many nuclear physicists several decades to understand the
More informationO-arai Engineering Center Power Reactor and Nuclear Fuel Development Corporation 4002 Narita, O-arai-machi, Ibaraki-ken JAPAN ABSTRACT
Characteristics of TRU Transmutation in an LMFBR M. Yamaoka, M. Ishikawa, and T. Wakabayashi O-arai Engineering Center Power Reactor and Nuclear Fuel Development Corporation 4002 Narita, O-arai-machi,
More informationRequests on Nuclear Data in the Backend Field through PIE Analysis
Requests on Nuclear Data in the Backend Field through PIE Analysis Yoshihira Ando 1), Yasushi Ohkawachi 2) 1) TOSHIBA Corporation Power System & Services Company Power & Industrial Systems Research & Development
More informationNaka-Gun, Ibaraki, , Japan
Examination of Atmospheric Dispersion Model s Performance - Comparison with the Monitoring Data under the Normal Operation of the Tokai Reprocessing Plant - M. Takeyasu 1, M. Nakano 1, N. Miyagawa 1, M.
More informationNuclear Chemistry. Chapter 24
Nuclear Chemistry Chapter 24 Radioactivity Radioisotopes are isotopes that have an unstable nucleus. They emit radiation to attain more stable atomic configurations in a process called radioactive decay.
More informationSpatially Dependent Self-Shielding Method with Temperature Distribution for the Two-Dimensional
Journal of Nuclear Science and Technology ISSN: 0022-3131 (Print) 1881-1248 (Online) Journal homepage: http://www.tandfonline.com/loi/tnst20 Spatially Dependent Self-Shielding Method with Temperature Distribution
More informationDEVELOPMENT OF HIGH RESOLUTION X-RAY CT TECHNIQUE FOR IRRADIATED FUEL ASSEMBLY
More Info at Open Access Database www.ndt.net/?id=18598 DEVELOPMENT OF HIGH RESOLUTION X-RAY CT TECHNIQUE FOR IRRADIATED FUEL ASSEMBLY A. Ishimi, K. Katsuyama, H. Kodaka, H. Furuya Japan Atomic Energy
More information4.4.1 Atoms and isotopes The structure of an atom Mass number, atomic number and isotopes. Content
4.4 Atomic structure Ionising radiation is hazardous but can be very useful. Although radioactivity was discovered over a century ago, it took many nuclear physicists several decades to understand the
More informationResearch Article Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal at Zero Power Benchmark Cases with NODAL3 Code
Hindawi Science and Technology of Nuclear Installations Volume 2017, Article ID 5151890, 8 pages https://doi.org/10.1155/2017/5151890 Research Article Analysis of NEA-NSC PWR Uncontrolled Control Rod Withdrawal
More informationA Monte Carlo Simulation for Estimating of the Flux in a Novel Neutron Activation System using 252 Cf Source
IOSR Journal of Applied Physics (IOSR-JAP) e-issn: 2278-4861.Volume 7, Issue 3 Ver. II (May. - Jun. 2015), PP 80-85 www.iosrjournals.org A Monte Carlo Simulation for Estimating of the Flux in a Novel Neutron
More informationNuclear Chemistry Unit
Nuclear Chemistry Unit January 28th HW Due Thurs. 1/30 Read pages 284 291 Define: Radioactivity Nuclear Radiation Alpha Particle Beta Particle Gamma Ray Half-Life Answer: -Questions 1-3 -Write the symbols
More informationTransmutation of Minor Actinides in a Spherical
1 Transmutation of Minor Actinides in a Spherical Torus Tokamak Fusion Reactor Feng Kaiming Zhang Guoshu Fusion energy will be a long-term energy source. Great efforts have been devoted to fusion research
More informationNEUTRON AND GAMMA FLUENCE AND RADIATION DAMAGE PARAMETERS OF EX-CORE COMPONENTS OF RUSSIAN AND GERMAN LIGHT WATER REACTORS
NEUTRON AND GAMMA FLUENCE AND RADIATION DAMAGE PARAMETERS OF EX-CORE COMPONENTS OF RUSSIAN AND GERMAN LIGHT WATER REACTORS Bertram Boehmer, Joerg Konheiser, Klaus Noack, Anatolij Rogov, Gennady Borodkin
More informationANALYTICAL EVALUATION OF THE UNCERTAINTY OF ON-LINE AXIAL POWER DISTRIBUTION MEASUREMENT WITH THE FOUR-SECTION EX-CORE DETECTOR
ANALYTICAL EVALUATION OF THE UNCERTAINTY OF ON-LINE AXIAL POWER DISTRIBUTION MEASUREMENT WITH THE FOUR-SECTION EX-CORE DETECTOR Jumpei Matsumoto, Kazuya Seki, Yasuo Komano Mitsubishi Heavy Industries,
More informationEvaluating the Core Damage Frequency of a TRIGA Research Reactor Using Risk Assessment Tool Software
Evaluating the Core Damage Frequency of a TRIGA Research Reactor Using Risk Assessment Tool Software M. Nematollahi and Sh. Kamyab Abstract After all preventive and mitigative measures considered in the
More informationJournal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 41, No. 7, p (July 2004)
Journal of NUCLEAR SCIENCE and TECHNOLOGY, Vol. 41, No. 7, p. 765 770 (July 2004) TECHNICAL REPORT Experimental and Operational Verification of the HTR-10 Once-Through Steam Generator (SG) Heat-transfer
More information