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 Kurfuerstendamm 200, 10719 Berlin, Germany Abstract. The ICRP has published in its 2007 recommendations new radiation weighting factors for neutrons to replace the values of the ICRP Publication 60 of 1990. Whereas the function of the neutron weighting factor does not change for energies above 1 MeV it decreases to about half of the former values for lower energies. In the present paper changes in the effective doses in mixed neutron gamma radiation fields near spent fuel assemblies and vitrified high level radio active waste (HAW) due to the new neutron weighting factors will be presented. These results are based on calculations of the radiation fields and of the effective doses in the vicinity of the mentioned radiation sources under free air conditions as well as outside of German transport and storage casks of CASTOR type loaded with these sources. The calculations have been made by means of the Monte Carlo Code MCNP for high and intermediate burn up of UO 2 -spent fuel, for MOX-spent fuel and for vitrified high active waste versus distance. The generation of neutrons was considered from spontaneous fission and from alpha capture and the gamma generation was considered from prompt emission and from neutron capture reactions. The results show that for practical applications the effective neutron dose decreases in the range of about 5 to 15 %. Taking into account the dominating contribution of the gamma dose compared to the neutron dose the influence of the new radiation weighting factors for neutrons to the total effective dose will be rather low. KEYWORDS: Effective Neutron Dose, New Neutron Weighting Factors, Spent Nuclear Fuel, HAW, Mixed Neutron Gamma Radiation Fields. 1. Introduction In the present paper changes of the dose rates of the effective dose in mixed neutron gamma radiation fields near spent fuel assemblies (FA) of typical German nuclear power reactors and near vitrified high level radio active waste (HAW) flasks from FA reprocessing depending on the distance up to 10 m according to the 2007 ICRP publication No. 103 [1] are investigated. The calculated dose rates consider the new radiation weighting factors for incident neutrons with energies below 1 MeV. Furthermore the dose rates outside of German transport and storage casks of CASTOR type loaded with theses radiation sources have been calculated versus distance. The calculations have been made by means of the Monte Carlo Code MCNP for high and intermediate burn up of UO 2 -spent fuel, for MOX-spent fuel and for vitrified high active waste. The results are e. g. important for the licensing of storage at the near NPP site interim storages as well as for radiation protection provisions during handling and transporting the casks to the interim storages. 2. Neutron Dose Weighting Factors after ICRP 103 2.1 Neutron Interaction with Human Body The change of radiation weighting factors w R for neutrons with energies E n below 1 MeV was based by the ICRP on the following assumptions [1]: At neutron energies below 1 MeV, the effect of the secondary photons produced in the human body is mainly responsible for the recommended decrease of the neutron weighting with decreasing energy. When Relative Biological Effectiveness (RBE) data obtained from investigations with small animals is used as the basis for the evaluation of a w R value applied to human exposure situations, the higher dose contribution from secondary photons in the human body compared to species with smaller bodies has to be taken into account. These photons are mainly produced by the (n, )-capture reactions of degraded neutrons in nuclei throughout the entire body. Their contribution to the total radiation weighted dose of an organ is strongly dependent on the 1
body size and on the position of the organ considered in the body. For external neutrons and whole body exposure, a mean value can be determined as an average over all tissues and organs of the human body. 2.2 Continuous Function of the Neutron Weighting Factor The effective dose, E, is defined as given in ICRP Publication 60 [2] by E w T w R D T,R (1) T R where w T is the tissue weighting factor with w T = 1 and D T,R is the mean absorbed dose due to radiation of type R and averaged over the volume of a specified organ or tissue T. The sum is performed over all organs and tissues of the human body considered in the definition of E. The unit of effective dose is J kg -1 with the special name sievert (Sv). A set of w R -values for various radiations was described in ICRP Publication 60 [2]. The only modifications recommended in 2007 for the calculation of radiation weighted doses are some numerical adjustments to be introduced for the values of w R for neutrons and protons. As in the 1990 Recommendations, radiation weighting factors are determined by the characteristics of the type and energy of the radiation incident on the body or, in the case of sources within the body, emitted by the source. The radiation weighting factors are then applied to the mean tissue dose in any specified part of the human body. The radiation weighting factor for -radiation recommended for general use in radiological protection is 1. For neutrons a continuous curve (C) is recommended shown in Figure 1. Figure 1: Radiation weighting factor, w R, for incident neutrons versus neutron energy E n. (A) step function and (B) continuous function given in ICRP Publication 60 [2], (C) New function calculated on the basis of equations (2) [1] 2
This function (C) is given also numerically in Equation 2. 2.5 + 18.2 exp[ -(ln E n ) 2 /6] for E n 1 MeV w R = (2) 5.0 + 17.0 exp[ -(ln (2E n )) 2 /6] for E n 1 MeV where E n is in MeV. The radiation weighting factor for neutrons w R is applied to the mean absorbed doses in the relevant tissues and organs. The dose is that from both the neutron induced charged particles and the secondary photons induced in the body. Figure 2: Ratio of weighting factors, w R, corresponding to functions (C) and (B) in Figure 1, i.e. (C)/ (B), versus neutron energy E n 3. Situation concerning Spent nuclear Fuel and HAW in Germany In 2000 the German Federal Government and the German nuclear utilities signed the atomic consensus agreement. This document limited reprocessing of German spent nuclear fuel in France and Great Britain to 2005 and storage of corresponding vitrified HAW shipped back at the central interim storage at Gorleben. The spent nuclear fuel of power reactors, i.e. UO 2 - and MOX-fuel assemblies, has to be stored at the site of the NPP in interim storages licensed for maximum 40 years. Up to now 12 interim storages for spent nuclear fuel have been erected and put into operation (see Figure 3). In the German policy of radioactive waste management a final disposal of spent fuel and HAW at a repository in deep geological formations is foreseen. 3
Figure 3: CASTOR type casks containing HAW flasks in a German central interim storage (left), interim storage of STEAG type at a German NPP site (right) 4. Calculation Method 4.1 Source Term, Radiation Transport and Dose Rate The calculations of the source term, i.e. the quantity of neutron and gamma radiation per unit time emitted from the FA or HAW flask, is the basis of the subsequent radiation transport calculations. The GRS has developed different versions of burn up codes to calculate the radio active inventory based on the Code ORIGEN (see e.g. [3, 4]). In the present paper the ORIGEN-X version with extended number of 14 reaction channels, of neutron cross section data and of energy range up to 20 MeV has bee been applied. The generation of the source term from the calculated radio active inventory data was carried out by the GRS code NGSRC [4], which calculated the neutron and gamma spectra of any homogeneous material composition, considering neutrons resulting from spontaneous fission and (, n)-reactions and gammas from prompt emission and (n, )-capture. The radiation transport and attenuation calculations have been performed with the three dimensional Monte-Carlo Code MCNP-4C, taking the cross section data from the ENDF/B-VI library [4] and the dose factors to convert the flux densities into dose rate from ICRP Publication 74 [5]. 4.2 Geometrical Model In the first case the source has been considered free in air (without cask, distance counted from the source surface center normal to the lenght axis) and in the second case the source has been considered to be contained in the cask as shown in Figure 3 (distance counted from the outer cask surface center next to the source). The geometrical model assumes the FA or the HAW flask as a homogeneous source. The inner diameter of a CASTOR cask amounts to 1,48 m, the wall thickness is 0,42 m and the height is 5,86 m. The steel wall contains rods of polyethylene for neutron shielding. The calculations have always been performed for one FA or HAW position (4 flasks in one column) as the source, the other positions being absorbers (shielding material) in the second case (Figure 3). 4
Figure 3: Geometrical Model of Source Position inside the CASTOR Cask for MCNP Calculations, View from Top, CASTOR V/19 with 19 FA of a PWR, (left), CASTOR HAW with 28 HAW flasks (4 flasks in one vertical column), (right) 5. Calculation Results 5.1 Nuclear Fuel and HAW Data The calculations have been performed for three typical spent fuel characteristics of spent fuel of German PWR power reactors (intermediate and high burn up of UO 2 -fuel and MOX-fuel) and for typical HAW from reprocessing, shown in Table 1. Table 1: Characteristic Input Parameters of Spent Fuel and HAW for Source Term Calculations Source Burn-up / Decay time Enrichment Neutrons / s FA Gammas / s FA UO 2 -FA 5 55 GWd/tHM - 5a 4,05 % U 235 8,07E+08 1,02E+16 UO 2 -FA 7 70 GWd/tHM - 3a 5,00 % U 235 1,48E+09 2,32E+16 MOX-FA 55 GWd/tHM - 3a 5,19 % Pu-tot 3,70 % Pu-fiss 5,46E+09 1,14E+16 HAW-Flask 1 column 40GWd/tHM - 7a 2,33E+09 2,99E+16 5
5.2 Source Term The source term calculations resulted in total neutron and total gamma emissions presented in columns 4 and 5 of Table 1. These results show that the gamma generation is about 7 Orders of magnitude higher than the neutron generation for all types of sources. The ratio of secondary to primary gamma generation is shown in Table 2 for the example of FA5. The results demonstrate that the generation of secondary gamma rays can be neglected. This conclusion is also valid for the other sources. Table 2: Ratio of secondary to primary gamma generation Rate of Secondary to Primary Gamma Ray Generation Gamma Source UO 2 -FA 5 (n, )-Capture / Prompt Emission Free Air Conditions 1,02 E-07 Behind CASTOR Wall 1,04 E-02 Analogously, the contributions of neutrons to the source term resulting from spontaneous fission and from (, n)-reactions have been analyzed for all sources. The results given in Table 3 show a small contribution from (, n)-reactions for spent fuel to the source term of less than 2 %, the numbers for the different UO 2 -FA are almost identical. The corresponding portion for HAW is considerably high and almost of the same order as that for spontaneous fission. Table 3: Contributions of neutrons to source term resulting from spontaneous fission and from (, n)-reactions for all sources Neutron Source Percentage of Neutron Generation Rate UO 2 -FA MOX-FA HAW-Flask Spontaneous Fission 98,69 % 99,04 % 55,29 % (, n)-reaction 1,31 % 0,96% 44,71 % The corresponding neutron spectra of HAW are subsequently harder compared to that of FA which has different influence on the calculation of the effective neutron dose for both types of sources on the basis of the new ICRP weighting factor for neutrons. To demonstrate this difference in more detail the corresponding neutron spectra are depicted in Figure 4. 6
Figure 4: Neutron generation spectra of UO 2 -FA and HAW (in relative units) versus neutron energy 5.3 Dose Rate The dose rate has been calculated separately for neutrons considering the new ICRP weighting factor w R and for gamma radiation. The results of the corresponding ratios are contained in Table 4 for all sources. These ratios behave slightly decreasing from 0 to 3 m distance from the source surface, but remain nearly constant from 3 to 10 m distance with the number indicated in Table 4. Table 4: Ratio of Neutron to Gamma Dose Rate Source Free Air Behind Wall UO2-FA 5 8,3 E-05 0,24 UO2-FA 7 6,7 E-05 0,14 MOX-FA 4,8 E-04 0,95 HAW-Flask 3,2 E-05 0,35 Although for free air conditions for all of the sources the effective neutron dose rate is negligible compared to that for gamma radiation, it has to be considered after moderation behind the wall of the CASTOR cask. Especially in the case of MOX-FA it reaches comparable values as for gammas. 7
The neutron dose rate both for free air conditions and after moderation behind the cask wall is shown in Figure 5 as a function of distance up to 10 m. Generally, it can be stated that the neutron dose rate in free air is about 3 orders of magnitude higher than the corresponding values behind CASTOR wall. In the near range up to 3 m distance the curves for free air drop more rapidly, however above 3 m the shapes of the curves of both sets are very similar. In both cases the highest dose rate is generated by the MOX-FA because of the highest value of neutron emission (see source term in Table 1). Also the order of the other curves follows nearly the according neutron emission values. Figure 5: Neutron dose rate of FA and HAW sources for free air and behind cask wall versus distance 8
In Figure 6 the ratios of rates of the effective neutron dose according to eq. (1), i.e. dose with the new weighting factors w R after ICRP 2007 versus dose with the old w R after ICRP 1991, of FA and HAW sources for free air and behind cask wall as a function of distance are plotted. The variation of the curves with distance is mainly due to the statistical uncertainty of the Monte Carlo calculation code of less than 5 %. Figure 6: Neutron dose rate ratio (ICRP 2007/ICRP1991) of FA and HAW sources for free air (left ordinate) and behind cask wall (right ordinate) versus distance In the case of free air conditions (upper part of Figure 6) generally a small influence of the change of the weighting factor w R can be stated. The decrease for all curves amounts to 4-5 %. The neutron spectra are in fact not moderated in air and thus the change of the weighting factor which is effective in the soft part of the spectra only does not lead to significant changes. The lowest change is observed for HAW due to the harder neutron spectrum compared to spent fuel (see Figure 4). In the case of neutron shielding, i.e. dose rate behind the CASTOR wall (lower part of Figure 6), neutron moderation in the wall takes place. Therefore, a bigger influence of the weighting factor change results. Generally, the decrease amounts to 14-16 %. The most influence as expected from the neutron emission spetra is observed for the HAW source. 9
6. Summary and Conclusion The aim of the paper was to investigate the influence of the new radiation weighting factor for neutrons after ICRP 2007 to the dose rates calculated for spent nuclear fuel and HAW free in air and behind a German transport and storage cask of CASTOR type. Although the influence for free air conditions is rather low the decrease of the neutron dose rates behind the cask wall amounts to 14-16 %. Taking the contribution of the gamma dose rate to the total dose rate also into consideration the neutron dose rate for free air conditions and hence the changes due to the new weighting factor can be neglected. In the case of radiations fields behind the cask wall however the neutron dose rate is of comparable amount as the gamma dose rate and thus the changes due to the new weighting factor have to be taken into account also for the total dose rate. The most significant influence was observed for spent MOXfuel assemblies. REFERENCES [1] ICRP (2008) 2007 Recommendations of the International Commission on Radiological Protection. ICRP Publication 103, Annals of the ICRP 37, 2-4. [2] ICRP (1991) 1990 Recommendations of the International Commission on Radiological Protection. ICRP Publication 60. Annals of the ICRP 21, 1-3. [3] PRETZSCH, G., Effective Dose near Spent Fuel Casks using the 2005 ICRP Recommendations, Australasian Radiation Protection Society Conference, ARPS 31, November 26-29, 2006, Sydney, Australia. [4] PRETZSCH, G., GMAL, B., HESSE, U., HUMMELSHEIM, K., Neutron activation of reactor components during operation lifetime of a NPP, IAEA-CN-155/062, Second International Symposium on Nuclear Power Plant Life Management, October 15-18, 2007, Shanghai, China. [5] ICRP (1997) Conversion Coefficients for Use in Radiological Protection against External Radiation, ICRP Publication 74, Annals of the ICRP 26, 3. 10