Optimum Arrangement to Minimize Total Dose Rate of Iron-Polyethylene Shielding System

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1 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 Technology Division, Ship Research Institute* Received June 1, 1988 Revised September 22, 1988 The shielding experiments using iron-polyethylene slab shields with a 2:2Cf neutron source were carried out to find out an optimum arrangement to minimize the total dose rate ; that is composed of neutron and secondary g-ray dose rates. The total thickness of the iron slabs was fixed at 32 cm, while a variety of thickness and location of a polyethylene slab in the iron slabs were employed as a parameter. The minimum dose point (i. e. optimum shielding arrangement) was observed when the polyethylene slab was located at approximately 20 cm depth from the source side in the arrangements. The ratio of the minimum dose rate obtained for the optimum arrangement to the maximum for the worst arrangement became 1/1.3 for the polyethylene slab of 1-cm-thick, 1/2.0 for 3-cm-thick, 1/2.9 for 6-cm-thick, 1/3.9 for 10-cm-thick and 1/3.6 for 14-cm-thick. The appearance of the optimum arrangement for the total dose rates, and changing profiles of the secondary g-ray as well as the neutron dose rates were reproduced for the typical two cases of the polyethylene slab thickness by the Monte Carlo calculations with the splitting technique. KEYWORDS: californium 252, optimum shielding arrangement, iron polyethylene slabs, thicknss, neutrons, secondary gamma radiation, dose rates, Monte Carlo calculation, MORSE-CG, rem counters, ionization-type survey meters, spent fuels, shipping cask I. INTRODUCTION Finding the optimum materials arrangement for a shield is the most interesting problem in the area of radiation shielding. As taking up in this study, shielding design of a spent fuel shipping cask is required to minimize a shield thickness and weight, and yet the designed cask must satisfy a criterion of neutron and r-ray dose rates around the cask. As described in Ref. (1), many researchers have been challenged to examine the shield optimization, but a general solution has not been found and may not be. Only a few experiments have contributed some insight to the study of optimization. Even in the recent textbook written by Chilton et al.(2), the description of the optimization techniques specifically applicable to shield design is a little, only on the properties for neutrons or r-rays of shielding materials are referred in the textbook. Many experiments have investigated the shielding properties of materials. Verbinski et al.(3) measured and calculated the spectral as well as spatial details of the fast neutron flux (>0.82 MeV) in water shields, and sulfur activation ratios were obtained in water. Schmidt(4) summarized the attenuation properties of concrete and various materials for neutrons of energy <15 MeV. Not only neutron transmission but also secondary r-rays produced in the materials were included in the study. Stoddard & Hootman(5) intensively calculated the neutron and g-ray dose rates in a slab of shielding materials with the source by the one-dimensional ANISN code. They obtained dose attenuation curves of fast neutrons, thermal neutrons, primary r-rays * Shinkawa, Mitaka-shi 181, -9-

2 412 J. Nucl. Sci. Technol., and capture g-rays for each material. In the Winfrith series of shielding benchmark experiments, Carter & Packwood(6)(7) investigated the shielding properties of iron and graphite to fission neutrons. The benchmark or the mock-up experiments on the multi-layers problem for the purpose of the shielding characteristics confirmation or the analysis accuracy evaluation have been reported, however, those studies which contribute to the shielding optimization have been few. Broder et al.(8) calculated the energy spectra in iron-water mixture shields and obtained the relaxation length for a flux of neutrons with an energy E>1.4 MeV in iron-water of varying composition. Miura et al.(9) measured neutron energy spectra and reaction rates in a laminated iron-water shield with the JRR-4 reactor, a swimming-pool-type research reactor. The results obtained were analyzed by the one-dimensional transport codes of ANISN and PALLAS. Santoro et al.(10) performed several experiments with the 14-MeV neutron source. They measured the transmission of 14-MeV neutrons through laminated slabs of fusion reactor shield materials and obtained the neutron and secondary g-ray energy spectra as a function of Type 304 stainless steel and the borated polyethylene slab composition and thickness. They also analyzed the results by the two-dimensional DOT code. Recently, neutron penetration experiments for FBR (fast breeder reactor) radial shield mockups were carried out by using the ORNL Tower Shielding Facility, and computational accuracy of the shielding design was investigated"". However, those work did not investigated the optimum arrangement for those materials. Fuse et al.(12) measured thermal and epithermal (0.67-~0.92-eV) neutron fluxes in laminated iron-water shields in the JRR-4 reactor pool. They concluded that the optimum arrangement was not a homogeneous mixture of iron and water layers, but a heterogeneous arrangement of thick-iron, water and thiniron. However, their work did not inquire into the optimum arrangement for high energy neutrons. Prior to the present work, Ueki et al.(13) have carried out the shielding experiments using iron-polyethylene slab shields with a 252Cf neutron source to find out an optimum arrangement to minimize the neutron dose rate. The total thickness of the iron slabs was fixed at 32 cm, while a variety of thickness and location of a polyethylene slab in the iron slabs was employed as a parameter. The magnitude of the iron slab thickness of 32 cm is that of a typical mild steel of the TN-series dry-type spent fuel shipping casks" which are very often used m Japan. The mild steel is covered with the 10 cm-thick resin shield. The iron-polyethylene shielding system is a simulation of the lateral shield for a new cask. Depending on the location of the polyethylene slab, the measured neutron dose rates changed remarkably in the iron-polyethylene shielding system. The minimum dose point for neutrons (i.e. optimum shielding arrangement) was observed when the polyethylene slab in the system was located near the detector. The appearance of the optimum arrangement were also reproduced by the Monte Carlo calculations with the splitting technique. However, finding out the optimum arrangement for the total dose rate in which secondary g-rays are included is more important for a shielding design. In this paper, we inquire into the optimum shielding arrangement for the total dose rate. Using the same experimental arrangements as those of Ref. OA the neutron and the secondary g-ray dose rates were measured by a neutron dosimeter and an ionization-type survey meter, respectively, as a function of the polyethylene slab location. The Monte Carlo calculations by the MORSE-CG code were carried out to make a reproduction of the optimum arrangement of the materials obtained from the experiment, and also to reveal that the Monte Carlo calculation can be employed as a reliable way to find out the optimum arrangement of materials in a practical shield design. The coupled neutron and ray library NGCP9-70 prepared for the present work was employed in the Monte Carlo g- code MORSE-CG. -10-

3 Vol. 26, No. 4 (Apr. 1989) 413 Chapter II of this paper describes the geometrical arrangement of the neutron shielding experiment with a 252Cf source. The measured dose rates are also presented in Chap.II as a function of the polyethylene slab location in the iron-polyethylene shielding system. The Monte Carlo techniques employed in the calculations are described in Chap. DI. In Chap. IV, the calculated results are compared with the measured values, and also discusses the characteristics of the energy spectrum of the secondary g-rays as well as the contribution ratio of the g-rays originated in the shields to the total dose rate. The conclusions obtained are summarized in Chap. V. II. DETAILS OF EXPERIMENT 1. Shielding Materials Arrangement Iron and polyethylene employed in the experiment are very popular neutron shielding materials. However, these materials have essentially different shielding characteristics ; iron has larger total cross sections for fast neutrons than polyethylene, whereas polyethylene has larger cross sections for intermediate and slow neutrons. Secondary g-ray production property is also different between iron and polyethylene. Accordingly, it can be expected that some arrangement of an ironpolyethylene system has an optimum shielding effect, as observed for the neutron dose rate(13). The iron-polyethylene slab, neutron source and detector location were set up as follows (see Fig. 1) : Fig. 1 Schematic arrangement of source, shields and detector (1) Californium-252 of the total mass 14.4 mg was used as a neutron source, which corresponds to 3.37x107 n/s. It was contained in the collimator of paraffin with a 25d irradiation cone. The neutron energy spectrum of 252Cf source is simulated by Eq. (1) (Ref.(15)) with a good approximation : N(E)~exp(-0.88)sinh[(2.0E)1/2], (1) where E is the neutron energy in MeV. (2) The configuration of the iron-polyethylene shield between the collimator and the counter is shown in Table 1. The iron slabs were located before and behind the polyethylene slab and its total thickness was fixed at 32 cm. Several thicknesses of the polyethylene slabs were taken as a arameter ; 1-, 3-, 6-, 10-, and 14-cm-thick. p The neutron and the secondary g-ray dose rates were measured as a function of the polyethylene slab location parameter (i.e. the thickness of the front iron slabs) L as shown in Fig. 1. The parameter L's were taken as 0, 5, 10, 15, 20, 25, 30 and 32 cm. Table 1 Configuration of iron-polyethylene slab system and thickness of each zone (3) The neutron dose rates were measured by a Studsvik 2202 D neutron dosimeter at a point of 15 cm behind the slab system in all the configuration. The secondary g-ray dose rates were also measured by an ALOKA ionization-type survey meter at the same point of the neutron dosimeter. The neutron dosimeter, with a polyethylene cylinder, was 212 mm in radius and 232 mm high ; the survey meter, with an ionization detector cylinder of air, was 150 mm radius and 100 mm high. The location of the effective center of the radiation detection is not yet clear, but the -11-

4 414 J. Nucl. Sci. Technol., authors assumed that it was 5 cm inside the cylinder of each detector. 2. Measurements The 252Cf source generates not only neutrons but also g-rays. Stoddard & Hlootman(5), however, reported that the dose rate of the primary g-rays was attenuated by four orders of magnitude by an iron shield of 32-cm-thick. Additionally, the primary g-ray dose rate at 1 m from a 252Cf source of 1 mg is 1.6x102 mrad/h -mg(15). Then, the g-ray dose rate at the detector location in Fig. 1 was estimated approximately 2.3x10-4 mrad/h. Accordingly, the g-rays observed behind the iron-polyethylene shield in the present experiment were recognized as the secondary ones, having originated in the detector side shields of the iron-polyethylene shielding system. Since the conversion factor of the neutron dosimeter depends on neutron energy in substance, the authors compared the neutron dose rates read from the neutron dosimeter indicator with the values obtained using the conversion factor of 3.3 counts/s=1 mrem/h. The value of 3.3 was obtained from a calibration with an Am-Be neutron source. The comparison of neutron dose rates was carried out for Configuration 3 in Table 1(13). The neutron dose rates obtained from the conversion factor of 3.3 were larger than those values obtained from the dosimeter indicator except for the 0-cm slab location (L=0 cm). However, the difference between the two values is as much as 9.4%. The countings were accumulated at least 1,000 counts (i.e. statistical error of 3.2%) to reduce the statistical error in each measurement. Consequently, the authors assumed that an error of 10% was included in the dose rates obtained from the conversion factor of 3.3. The measured g-ray dose rates with the ionization-type detector were also considered to have an error of 5% around 0.1 mr/h, ~50% around 0.01 mr/h, depending on the magnitude of the measured values. As the measured values decreased, the measurement error was increased by cosmic rays and other g-rays from the environment. To measure the room-scattered neutron components, additional thick polyethylene slabs were piled up only in front of the neutron dosimeter to shield neutrons completely from the source and the slabs. In this configuration, background g-rays from the structures around the iron-polyethylene shield were also measured by the survey meter. The roomscattered neutrons were estimated to be 0.2 count/s (0.06 mrem/h) ; the background g-ray dose rate was to be 0.03 mr/h. The values were subtracted from the measured data in each measurement as the background. Due to the subtraction of the room-scattered neutron and g-ray components, it may be accepted that the room structures (floor and room walls) do not need to take into account in the Monte Carlo analysis. In this measurements, the g-ray energies detected by the survey meter were ranged between ~0.01 and ~10 MeV. Accordingly, in order to estimate the total dose rate, the measured g-ray dose rates in a unit of mr/h were directly added to the neutron dose rate in mrem/h with an assumption of mr/h?? mrem/h(16). 3. Measured Results The neutron and the secondary g-ray dose rates were measured behind the iron-polyethylene shield for five different configurations indicated in Table 1 as a function of the polyethylene slab location. The measured results are summarized for each configuration in Fig. 2(a)~(e), The following remarks can be made from the present measurements. As the characteristics of the neutron dose rate profiles were already explained in detail in Ref. (13), the main descriptions in this section are on the secondary g-rays and the total dose rates. (1) The secondary g-ray dose rate increased with an increase of L. For L<15 cm, the significant g-ray dose rates were not observed in all the arrangements. (2) The contribution of the secondary g-ray dose rate to the total was gradually increased as the polyethylene slab was moved to the back side, i. e. to the detector. The maximum contribution of the secondary dose rates to the total was observed when the polyethylene slabs were located at -12-

5 Vol. 26, No. 4 (Apr. 1989) 415 L=30 or 32 cm, i, e. the iron slab of 30- or 32-cm-thick was set to the source side, in all the configurations. In Confiuration 1 (1-cm-thick polyethylene slab), the maximum dose rate of secondary -rays was 0.12 mrem/h at L=30 g cm, and its contribution to the total dose rate was 1.6%; the rest of 98.4% was from ~ the neutrons. In Configurations the secondary dose rates were 0.23, 0.23, 0.17 and 0.12 mremjh and their contributions were 6.6, ~40.0 and ~54.5% at L=30, ~ 32, 32 and 32 cm, respectively. In Configuration 5 (14-cm-thick polyethylene slab), the maximum contribution of the secondary -rays is almost the same as those g of the leakage neutrons through the iron-polyethylene slabs. (3) In Fig. 2(a)~(e) the minimum total dose rate (i.e. result of the optimum shielding arrangement) was observed when the polyethylene slab was located at L=~20 cm in all the configurations. The secondary -ray dose rates were gradually gincreased as the polyethylene slabs were moved to Fig. 2(a)~(e) Changing of total, neutron and secondary r-ray dose rates as function of polyethylene slab location in Configurations 1~5 for 1-, 3-, 6-, 10- and 14-cm-thick polyethylene slab -13-

6 416 J. Nucl. Sci. Technol., Fig. 2 (d) Configulation 4 (Polyethylene 10 cm) Fig. 2 (e) Configulation 5 (Polyethylene 14 cm) the detector, so that minimum total dose points were exhibited more clearly than those of the neutron dose rates(13). Table 2 compares the location L at which the minimum dose rate is observed for the total dose rates with that of the neutron dose rates. Table 2 Location L at which minimum dose rate was observed in Figs. 2 (a)~(e) (4) As indicated in Fig. 2(a)~(e), the maximum total dose rates that mean the worst arrangement were 9.1, 5.6, 2.8, 1.2 and 0.5 mrem/h in Configurations 1-5, respectively. On the other hand, the minimum total dose rates were 6.8, 2.8, 0.97, 0.31 and 0.14 mrem/h. Using the above values, the ratios of the minimum dose rates to the maximum were estimated to be 1/1.3 for Configuration 1, 1/2.0 for Configuration 2, 1/2.9 for Configuration 3, 1/3.9 for Configuration 4, and 1/3.6 for Configuration 5. The values distributed from 1/1.3 to 1/3.9, which corresponded to the ratios for neutron dose rates from 1/1.3 to 1/5.4. Due to the contribution of the secondary g-rays, the location L of which the minimum total dose rates is observed is moved to the source side as compared with that of the neutron dose rate. So that, the minimum value of total dose rate with the g-rays is larger than that of neutron dose rate without the g-rays in all the configurations of Figs. 2(a)~(e). Consequently, the ratios of the minimum to the maximum of the total dose rate become smaller than those of the neutron dose rate. (5) The configuration of a 32-cm-thick iron and 10-cm-thick polyethylene neutron shield is popular in a dry-type spent-fuel shipping cask, like the TN-12A cask(14). The neu - tron source spectrum of a 252Cf source is - 14-

7 Vol. 26, No. 4 (Apr. 1989) 417 slightly harder than that of 296U fission neutrons, but the difference is small. Accordingly the present results of the optimum shield arrangement taking into account the secondary g-rays can easily be applied to the design of casks containing irradiated nuclear fuels with a strong neutron source. This configuration has also a merit that the inner polyethylene is noncombustible against a fire than the resin shield covering the TN-series casks. III. MONTE CARLO CALCULATIONS 1. Monte Carlo Techniques The Monte Carlo calculations were carried out by the MORSE-CG codeo(17) for Configurations 3 and 4. Source neutrons were emitted isotropically in a 4p steradians with a statistical weight of 1.0. As illustrated in Fig. 1, only the source neutrons emitted within a cone angle of 25d can enter the shield without any collisions in the collimator. The room structures around the experimental system were neglected, and for the reason described in Chap. II, primary g-rays of the 252Cf source did not take into account in the calculations. To sum up the necessary information on particle histories, the authors must select the estimator from several possibilities of collision density, track length, boundary crossing, nextevent surface crossing (NESX), and point detector type ones. In the calculations, the point detector estimator due to the last collision method was employed at the detector position. The air around the polyethylene-iron shields was treated as vacuum. To reduce the fractional standard deviation (FSD) of the calculated dose rate, the splitting technique was employed intensively in the present calculations. The exponential transform and the Russian roulette kills and survivals were also used in the calculations. The space and energy group dependent statistical weights for neutron splitting were given in each region of the iron-polyethylene shielding system by an empirical formula. The formula was introduced by using the neutron dose rate attenuation curves of the polyethylene and the iron slabs. The attenuation curves were obtained from a pre-experiment. The detailed descriptions of the empirical formula were given in. Ref.(13). The statistical weights for the secondary g-ray splitting were also given in each region of the shielding system ; the weights employed in the calculations were WGLOs in the MORSE-CG code. The probability of generating a g-ray is defined as WTN/WGLO in the MORSE-CG code, where WTN and WGLO are the statistical weight of neutrons and the weights to be assigned to secondary g-rays, respectively. The probabilities were artificially given so as to obtain a good statistics of secondary g-ray dose rate behind the iron-polyethylene shielding system, by considering that the g-rays generated in the detector side region of the iron-polyethylene shielding system were contributed more strongly to the detector response behind the shielding system than the g-rays near the source. In particular, the capture -rays of thermal neutrons had very dominant g contribution at the detector point. Accordingly, the parameter WGLO was taken as 1.0 in the source side region to restrain the production of less contributing secondary g-rays ; on the other hand, to originate enough g-rays in the detector side region, WGLO was 0.035, and in the middle region of the shielding system, it was 0.35~0.1. The NGCP9-70 cross-section library(13) which was produced by using the NJOY(18) and the AMPX(19) codes from the ENDF/B-IV library(20) was employed in the MORSE-CG code. The library contains coupled 50-groups neutron and 20-groups g-ray microscopic cross sections. The maximum Legendre expansion coefficient was P9. The atomic composition of the iron was 8.476x1022 atom/cm3 and was considered in the iron slab. The atomic compositions of hydrogen and carbon in the polyethylene slab were considered to be 7.912x1022 and 3.956x 1022 atom/cm3, respectively. IV. CALCULATED RESULTS AND DISCUSSIONS The Monte Carlo calculations with 50,000 source neutrons were carried out for Configu- -15-

8 418 J. Nucl. Sci. Technol., Fig. 3(a), (b) Comparison of total, neutron and secondary g-ray dose rates between measured and Monte Carlo calculated values in Configurations 3 and 4 for 6- and 10-cmthick polyethylene slab, with total iron thickness of 32 cm rations 3 and 4, i.e. 6- and 10-cm-thick polyethylene slabs, respectively. The comparison of total, neutron, and secondary g-ray dose rates between the measured and the calculated values is shown in Fig. 3(a) and (b). Furthermore, the secondary g-ray energy spectra in Configuration 4 were calculated for the location of L=25, 30 and 32 cm. The Monte Carlo calculated energy spectra for each location L were summarized in Fig Total, Neutron and Secondary -ray Dose Rates g All the FSDs of the Monte Carlo results shown in Fig. 3(a), (b) were within 0.05 for the neutron dose rates and within 0.10 for the secondary g-rays. An agreement between the measured and the calculated neutron dose rates was excellent within 10%, for both Configurations 3 and 4, except at the location of L =32 cm for Configuration 4 in Fig. 3(b). At the location, the calculated neutron dose rate was overestimated by ~25% as compared with the measured value. The calculated secondary g-ray dose rates agreed fairly well with the measured values within 20 to 50% for Configuration 3 and Fig. 4 Comparison of Monte Carlo calculated secondary g-ray energy spectra between different polyethylene slab locations of L=25, 30 and 32 cm in Configuration 4-16-

9 Vol. 26, No. 4 (Apr. 1989) 419 within 10 to 30% for Configuration 4. The secondary g-ray dose rates were in order of the magnitude of 0.1 mrem/h at the most for all the arrangements, so that the g-ray dose rates were measured with not so small error within several ten percent by the ionizationtype survey meter. Taking into account the measured errors, it can be said that the calculated dose rates of secondary g-rays agreed fairly well with the measured values. However, some discrepancy emerged between the measured and the calculated dose rates may be due to the uncertainties in the g-ray production cross sections in the ENDF/B-IV(20). Accordingly, it might be concluded that the Monte Carlo calculations can produce fairly good results for the secondary g-ray dose rate in the shielding system as like the neutron dose rate(13) 2. Secondary g-ray Energy Spectra The calculated g-ray spectra are characterized by the remarkable three peaks at 7.5, 8.0, 2.0~2.5 and 0.1~0.5 MeV. The FSDs of the main three peaks were within 10% for the three locations of L=25, 30 and 32 cm. The peak of 0.1~0.5 MeV is due to the Compton scattered g-rays. The peaks of 7.5,8.0 and 2.0~2.5 MeV can be attributed to thermal neutron capture by iron and hydrogen atoms in the iron slab and in polyethylene slab respectively by referring to Table 3 in which capture cross sections and photon yields of thermal neutron capture are summarized for several materials concerning this study. In additions, the secondary g-rays due to the inelastic scattering of neutrons by the iron atoms give a subdominant contribution to the spectrum. Table 3 Cross section and photon yields for thermal neutron capture in various elements(2) The contribution ratio of the remarkable -ray peaks revealed in Fig. 4 to g the g-ray dose rate is summarized in Table 4. Their contribution is gradually increased as L of the polyethylene slab is moved to the detector side, as shown that the total contribution ratio Table 4 Contribution ratio of remarkable g-ray peaks to total g-ray dose rate given in Fig. 4, calculated for Configuration 4 with 10-cm-thick polyethylene slab of the three peaks is 42.7% at L=25 cm, 54.5% at L=30 cm and 67.8% at L=32 cm. Paying our attention to the g-rays of 2.2 MeV in Configuration 4, its contribution is found to be dominant from 12.1 to 36.1% as the location of the polyethylene slab is moved from L=25 cm to the detector side of L=32 cm. On the contrary, the contribution of the 7-rays of 7.5~8.0 MeV at L=30 cm of 24.4% is larger than that of 19.8% at L=32 cm. The secondary g-rays existing in the ironpolyethylene shielding system are mainly produced by the interactions of slow and thermal neutrons with hydrogen and iron atoms. Because those neutrons disappear rapidly as leaving the polyethylene slab, the production of secondary g-rays become less in the distant area from the polyethylene slab. Accordingly, the secondary g-rays originated - 17-

10 420 J. Nucl. Sci, Technol., in the major part of the iron slab did, not make so much significant contribution to the dose rate as the g-rays produced around the polyethylene slab did. V. CONCLUSION The shielding experiments using ironpolyethylene slab shields with, a 252Cf neutron source were carried out to find out an optimum arrangement to minimize the total dose rate of transmitted neutrons and secondary g-rays. The Monte Carlo calculations were carried out for the thickness to reproduce the measured profiles. The following remarks are obtained in the study (1) The minimum total dose point was observed where the polyethylene slab was located at L=~20 cm depth in all the ironpolyethylene configurations. The location of L=~20 cm was less depending on the polyethylene slab thickness in the shielding system ; meanwhile, the location was gradually moved toward the detector side for the neutron dose rate(13) as the slab thickness increased. (2) The total dose rates were reduced significantly by the shielding optimization. For instance, the optimum arrangement could make the total dose rate to ~1/4 in Configuration 4, as compared with the worst arrangement. For this result, it can be concluded that the optimization of a shielding system is important not only from a economical view point but also from reduction of radiation level in an environment. (3) The appearance of the optimum arrangement in the configurations for the total dose rates, and changing profiles of the secondary g-ray as well as the neutron dose rates were reproduced in the typical cases by the Monte Carlo calculations with the splitting technique. Consequently, it can be said that the optimum arrangement for a general shielding system is obtained with good accuracy by the Monte Carlo calculation. (4) Because the most secondary g-rays were originated in the polyethylene slab and in the iron slab near the polyethylene, and the polyethylene had a little shielding ability to g-rays, the maximum g-ray dose rate was observed in the case of the arrangement where the polyethylene slab was located just in front of the detector. As the thickness of the polyethylene slab increased, the contribution of the secondary g-rays to the total dose rate became more significant ; the maximum contribution was 7% at the most for the polyethylene slab of- less than 3-cm-thick, and approximately the same magnitude as the neutrons for the 14-cm-thick. (5) In the present stages, it is difficult to predict or to determine the optimum arrangement for dose rate in the general shielding configuration. However, we will extent this study to the goal of the shielding optimization through an extensive use of the Monte Carlo calculations as well as further experiments with various materials and arrangements. ACKNOWLEDGMENT We would like to thank Mr. Masatoshi Tsuji and Mr. Mikinori Ono of Mitsui Engineering & Shipbuilding Co., Ltd. for their helpful works on the experiment. REFERENCES (1) SCHAEFFER, N.M., Ed.: "Reactor Shielding for Nuclear Engineers", (1973), USAEC. (2) CHILTON, A.D., SHULTIS, J.K., FAW, S.E.: "Principles of Radiation Shielding", (1984), Prentice-Hall. (3) VERBINSKI, V. V., et al.: Nucl. Sci. Eng., 27, 283 (1967). (4) SCHMIDT, F. A.R. : The attenuation properties of concrete for shielding of neutrons of energy less than 15 MeV, ORNL-RSIC-26, (1970). (5) STODDARD, D. H., HOOTMAN, H. E.: 252Cf shielding guide, DP-1246, (1971). (6) CARTER, M. D., PACKWOOD, A.: The Winfrith benchmark experiment in iron, NEACRP-U-73, (1976). (7) idem: Private communication, (1984). (5) BRODER, D. L., et al.: At. Energ., 38, 40 (1975). -18-

11 Vol. 26, No. 4 (Apr. 1989) 421 (9) MIURA, T., et al.: J. Nucl. Sci. Technol., 16 (8), 563 (1979). (10) SANTORO, R. T et al.: Nucl. Sci. Eng., 78, 259 (1981), (p) OHTANI, N., et al.: Benchmark experiment and analysis of neutron penetration through FBR radial shield mockups, presented at 7th Int. Conf on Radiation Shielding, Technical Session 5.2, Sep. 1988, Bournemouth. (12) FUSE, T., et al.: Nucl. Struct. Eng., 13, 390 (1970). (13) UEKI, K., NAMITO, Y.: Nucl. Sci. Eng., 96, 30 (1987). (14) UEKI, K., et al.: Nucl. Technol., 74, 164 (1986). (15) PROFIO, A. E. : "Experimental Reactor Physics", (1976), John Wiley & Sons, New York. (16) Radioisotope Assoc. of Japan, Ed. : "Isotope Handbook", (in Japanese), (1984), Maruzen MMETT, M. B.: The MORSE Monte (17)E Carlo radiation transport code system, ORNL-4972, (1975). (18) MACFARLANE, R. E., et al.: The NJOY nuclear data processing system, LA-9303-M (ENDF- 324), (1982). (19) GREENE, N. M., et al.: A modular code system for generation coupled multi-group neutrongamma libraries from ENDF/B, ORNL/TM- 3706, (1976) ; (revised 1978). (20) GARBER, D. Ed. : ENDF/B summary documentation, BNL-NCS (ENDF-201), 2nd ed. (ENDF/B-IV), (1975). -19-