SET2011, 10th International Conference on

ASSESSMENT OF RADIONUCLIDE DISPERSION FROM BUSHEHR NUCLEAR POWER PLANT STACK DURING NORMAL AND ACCIDENT CONDITIONS AND ITS EFFECT ON POPULATION Ahmad Pirouzmand 1, Peyman Dehghani1, Kamal Hadad1, Mohammadreza Nematollahi1, Ebrahim Sharifi2 1

Department of Nuclear Engineering, School of Mechanical Engineering, Shiraz University, Shiraz, Iran 2

Bushehr Nuclear Power Plant, Bushehr, Iran Abstract

Assessment of the total effective dose equivalent (TEDE) around the Bushehr nuclear power plant (BNPP) site during both normal operation and accident conditions which are essential for the safety and environmental analyses is presented in this paper. Release of radioactive material to the environment following a design basic accident is evaluated using HOTSPOT and CAP88 health physics computer codes. Both Codes utilize a Gaussian dispersion air transport plume model to simulate the radionuclide atmospheric dispersion in different atmospheric stability classes (A–F) and various wind speeds and directions. To calculate the TEDE, the surrounding area of the BNPP unit 1 within a radius of 65 km is gridded into 12 concentric rings and 16 sectors, and the distribution of population and agricultural products are calculated for each grid. In normal operation condition, atmospheric dispersion of radioactive material is calculated using CROM code. The meteorological data on atmospheric stability conditions, population, agriculture, wind speed and the frequency distribution of wind direction based on data collected near the reactor site are also analyzed and presented here. The calculated results of this study in normal and accident conditions are compared to experimental and BNPP FSAR data, respectively. Keywords: CAP88, HOTSPOT, BNPP, Radioactive dispersion model.

1. Introduction One of the most crucial safety and health physics concerns in research or power reactors is the radioactive material release around the reactor. Due to the inevitable presence of personnel within the reactor site and the population outside the site, it should be monitored frequently and properly especially for design basic accidents (DBAs) conditions. The effect of radioactive material, which released normally, should also be studied for the possibility of fatal cancer risk [1]. The BNPP-1 has been recently commissioned and is now operational. The BNPP-1 has been generally sited in a relatively low population zone, with the basic objective of limiting the dose received by the members of the public as a whole under normal and accident conditions [2]. The accidental releases of radioactive material from reactors causes the dose equivalent which may be external or internal to the body. The total effective dose equivalent (TEDE) is the sum of the effective dose equivalent (EDE) that caused by the external material such as cloud submersion and ground shine with the committed effective dose equivalent (CEDE) that caused by the internal material as a result of inhalation. The TEDE is the most complete expression of the combined dose from all applicable delivery pathways [3]. The radioactive materials in the form of gas and particulate may be released into the air and transported by the wind to the around environment. Their concentration in air and deposited on ground surface depend on important parameters such as the amount of released radionuclides, the wind speed, the atmospheric stability and other conditions which are evaluated in this study. To prevent the risk of released radionuclides as radiation protection and health physics aims, The TEDE for released radioactive material from reactors stack on the around under normal and accident conditions must be assessed [1]. In this work the area around the BNPP-1 is divided into three zones as described in the FSAR [4]. Then using HOTSPOT and CAP88 health physics codes in LBLOCA condition and CROM code in the normal operation the TEDE is calculated.

2. Materials and methods The HOTSPOT health physics code provides a first-order approximation of the radiation effects associated with the atmospheric release of radioactive materials. The HOTSPOT program has been created to equip emergency response personnel and planners with a fast, field-portable set of software tools for evaluating incidents involving radioactive material. This program is designed for short-range (less than 10 km), and short-term (less than a few hours) predictions [5]. Also CAP88-PC code utilizes a modified Gaussian plume equation to estimate the average dispersion of radionuclides released from up to six emitting sources. In this code the assessments are done for a circular grid of 

Corresponding Author: [email protected]

NuRER – 4. International Conference on Nuclear and Renewable Energy Resources, Antalya, TURKEY, 26-29 Oct. 2014

distances and directions for a radius of up to 80 kilometers around the facility. The Gaussian plume model is simple to use, has been widely tested, and has been shown to give reasonable predictions if a proper selection of model parameters is made. The HOTSPOT code has been designed for short-term (less than a few hours) release durations. Users requiring radiological release consequences for release scenarios over a longer time period, e.g., annual windrose data, are directed to such long-term models as CAPP88-PC. Users requiring more sophisticated modeling capabilities, e.g., complex terrain; multi-location real-time wind field data; and etc. [5]. CROM application is a code has been designed to automate the calculation of radionuclide concentrations in different environments and their impact in the nutritional chain, as well as in the human being, allowing to the researcher to center in the obtained results analysis. Herein this code is applied for the effective annual dose assessment in the normal operation. HOTSPOT and CAP88 codes utilize the Gaussian model that has been widely used and verified in the scientific community and is still the basic workhorse for initial atmospheric dispersion calculations. The Gaussian model generally produces results that agree well with experimental data in simple meteorological and terrain conditions [6]. Eq. (1) determines the time-integrated atmospheric concentration of a gas or an aerosol at any point in space using the Gaussian model.



C x, y , z , H



 1  y 2   1  z  H 2   1  z  H  2     x    exp          exp      exp   DF  X  2 Y  Z u  2   Y    2   z    2   z     u  Q

(1)

Where C is the time-integrated atmospheric concentration (Ci-s)/(m3); Q is the source term (Ci); H is effective release height (m); x,y and z are the downwind, crosswind and vertical axis distance (m); σy and σz are the standard deviation of the integrated concentration distribution in the crosswind and vertical directions (m); u is the average wind speed at the effective release height (m/s); L is the inversion layer height (m); DF(x) is the plume depletion factor and  is the radioactive decay constant. The x axis is the downwind axis, extending horizontally with the ground in the average wind direction. The y axis is the crosswind axis, perpendicular to the downwind axis, also extending horizontally. The z axis extends vertically from the ground. If the inversion layer option is in effect, and  z exceeds the inversion height (L), Eq. (2) is used: C  x, y , z , H  

2   1 y     x exp     exp    DF  X   2  Y   2 Y Lu  u  Q



(2)



To avoid the sharp transition between the two above equations, the transition into the inversion layer equation begins when σz equals 70% of L and is complete when σz equals L. Between these two values, the two equations are linearly interpolated. During the analysis of the consequences of primary LOCA with large diameter pipe, radioactive materials released into the environment through the stack. Table 1 provides the calculated results of fission products (FPs) release into the environment in case of accident involving the rupture of the largest primary circuit pipeline (e.g. Din 850 mm) [4]. Meteorologists distinguish general states of the atmospheric surface layer as to how they compare to the adiabatic lapse rate (unstable, neutral, and stable). These categories refer to how a parcel of air reacts when displaced adiabatically in the vertical direction [7, 8, 9]. Table 2 contains criteria for the six stability classes, based on five categories of solar insolation. This scheme is widely used in the meteorology and is accepted for stability class estimation. For an elevated release, the location of the maximum concentration depends on the chosen stability class. For materials with a deposition velocity of zero and a release point at or very near the ground level, the maximum Table 1. Radionuclides release into the environment in case of large break LOCA [4]. Radionuclides Release (Bq) Radionuclides Release (Bq) Molecular iodine Gases I-131 9.56E+11 Kr-85m 6.17E+10 I-132 2.07E+11 Kr-87 1.77E+10 I-133 2.07E+11 Kr-88 9.52E+10 I-134 5.58E+09 Xe-133 3.28E+14 I-135 4.42E+10 Xe-135 2.13E+11 Xe138 2.27E+09 Organic iodine Aerosols I-131 5.00E+11 Cs-134 4.94E+10 I-132 4.78E+09 Cs-137 9.34E+09 I-133 2.31E+10 I-134 8.26E+07 I-135 1.96E+09

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Table 2. Meteorological conditions used to define the atmospheric stability categories [5]. Ground wind speed (m/s) Sun high in sky Sun low in sky Night time 6 C D D

Figure 1. Population distribution sectors within 100 km from the BNPP-1 stack concentration is always associated with F stability class. However, if the deposition velocity is greater than zero, the worst-case stability at large downwind distances is not always associated with F stability due to the effects of plume depletion, i.e., the concentration of the plume decreases at a faster rate with increasing stability class (A-F) and increasing deposition velocity [8, 9]. To determine the atmospheric stability classes and construct meteorological input files, Table 2 is used in this study. In the process of updating of Bushehr nuclear power plant (BNPP) site characteristics, a surveillance program identifying places of significant population grouping, such as villages, towns, cities within 100 km from BNPP site is carried out. A radius of 100 km from BNPP covers Bushehr harbor that is the most populated city in the region and is the center of Bushehr Province located at about 18 km from the site, is studied. The region under surveillance is partitioned in zones by concentric circles with the reactor unit-1 at the center point. The circles were divided into 22.5 degree sectors with each sector centered on one of the 16 compass points. The population centers within 100 kilometers of the site, as shown in Fig. 1, based on 2012 census conducted by the statistical center of IRAN, are considered.

3. Results In the present study, total effective dose equivalent received by the public living around the site of BNPP-1 due to an accidental release of radionuclides into the environment involving a hypothetical LBLOCA are investigated. The version 4.0 of CAP88-PC code and the version 3.0 of HOTSPOT code are applied for this purpose. For normal operation dose assessment, the version 6.0.3 of CROM code is used. The calculated results using CROM code for normal operation release through the stack to the environment for both internal and external paths in ENE direction and two age intervals (e.g. 7-12 and 12-17) are presented in Table 3. The results demonstrate that nuclear safety standards in the design of BNPP-1 are included and this plant can operate safety without substantially risk to the people and the environment. Table 3. Results of effective annual dose for external irradiation and dose by annual intake for internal irradiation calculated around the ENE sector of BNPP-1 by CROM code External Internal Irradiation External Internal Irradiation Radionuclide Irradiation (Sv/y) Irradiation (Sv/y) (Sv/y ) (Sv/y ) Receptor: Sector ENE, 2km Age: 7-12 Receptor: Sector ENE, 2km Age: 12-17 7.41E-12 7.49E-11 4.36E-11 3.06E-10 Co-58 3.09E-10 4.27E-10 6.92E-10 2.47E-09 Co-60 2.36E-07 8.13E-09 4.83E-07 1.74E-08 Cs-134 5.04E-07 3.05E-11 7.91E-07 9.81E-11 Cs-137 8.87E-13 4.42E-10 3.58E-12 2.06E-09 Fe-59 8.92E-10 4.61E-08 2.74E-09 3.92E-07 I-131 4.40E-11 3.71E-10 6.20E-11 8.12E-10 Mn-54 5.22E-14 1.53E-12 7.04E-14 9.49E-12 Nb-95 7.36E-11 4.08E-10 9.16E-11 7.36E-10 Zr-95

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Table 4 presents the calculated TEDE and cancer risk using CAP88 code. The total effective dose equivalent values are equal to 2.79E-06 Sv, and 7.30E-02 person-Sv for individual and collective, respectively. Also results of the risk calculations show that the total value of fatal cancer risk to an individual is equal to 1.28E-07. The total number of deaths in populations is equal to 1.13E-05 deaths per year. Calculated total effective dose using HOTSPOT code and a comparison with CAP88 results and FSAR data are tabulated in Table 5. Also Fig. 2 presents these data as a bar presentation for further investigation. This figure shows that while HOTSPOT gives more accurate results near the center of BNPP site, the CAP88 presents more precision in the far distances. Fig. 3 illustrates the total effective dose pathways contribution for the individual and the collective doses are calculated by CAP88 code. Table 4. Calculated effective dose equivalent and cancer risk per year for the main radionuclides using CAP88 code Effective Dose Equivalent Cancer Risk Nuclide Selected Individual Collective Population Selected Individual Collective Population (Sv) (person-Sv) Total Lifetime Population Fatal Fatal Cancer Risk Cancer Risk (Deaths) I-131 1.83E-06 5.86E-02 7.77E-08 5.87E-06 Xe-131m 7.42E-10 4.22E-06 3.18E-11 2.35E-09 I-132 1.74E-08 7.01E-05 9.27E-10 4.84E-08 I-133 4.97E-08 3.25E-04 2.24E-09 1.61E-07 Xe-133m 6.84E-11 3.72E-07 3.35E-12 2.36E-10 Xe-133 6.80E-08 5.74E-04 3.33E-09 3.64E-07 I-134 2.36E-10 6.97E-07 1.26E-11 4.80E-10 I-135 6.95E-09 3.44E-05 3.64E-10 2.33E-08 Xe-135m 2.97E-10 1.56E-06 1.59E-11 1.08E-09 Xe-135 1.38E-09 7.81E-06 7.20E-11 5.28E-09 Cs-135 1.35E-18 3.39E-13 1.33E-19 4.49E-16 Kr-85m 5.96E-11 4.06E-07 3.13E-12 2.77E-10 Kr-85 6.76E-19 4.49E-14 2.04E-20 1.76E-17 Kr-87 1.89E-25 4.13E-07 4.96E-12 2.90E-10 Rb-87 2.63E-25 3.87E-19 2.04E-26 1.78E-23 Kr-88 1.28E-09 7.78E-06 7.08E-11 5.57E-09 Rb-88 2.02E-09 2.60E-06 1.01E-11 1.68E-09 Xe-138 9.98E-12 1.95E-08 5.47E-13 1.39E-11 Cs-138 7.26E-12 3.83E-08 3.89E-13 2.66E-11 Cs-134 7.51E-07 1.20E-02 3.23E-11 4.48E-06 Cs-137 5.05E-09 9.36E-04 9.37E-11 2.16E-08 Ba-137m 5.81E-08 4.56E-04 3.14E-09 3.20E-07 TOTAL 2.79E-06 7.30E-02 1.28E-07 1.13E-05 Table 5. Results of total effective dose (Sv) calculation with CAP88, HOTSPOT compared with FSAR data in 8 sectors around the BNPP-1 (distance in km from stack) Distance 0.6

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10

12

15

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Codes CAP88-PC FSAR HOTSPOT CAP88-PC FSAR HOTSPOT CAP88-PC FSAR HOTSPOT CAP88-PC FSAR HOTSPOT CAP88-PC FSAR HOTSPOT CAP88-PC FSAR HOTSPOT

NNW 1.40E-05 2.27E-03 9.85E-03 1.00E-05 7.96E-05 2.44E-04 9.80E-06 2.77E-05 2.49E-05 9.80E-05 2.27E-05 1.07E-05 9.70E-06 1.67E-05 3.12E-06 9.70E-06 7.00E-06 8.48E-09

N 1.50E-05 3.32E-03 1.00E-02 1.00E-05 8.12E-05 2.92E-04 9.80E-06 2.82E-05 9.65E-05 9.80E-06 2.19E-05 6.76E-05 9.80E-06 1.62E-05 4.10E-05 9.70E-06 6.82E-06 4.35E-06

NNE 1.40E-05 2.56E-03 1.03E-02 1.00E-05 8.78E-05 2.96E-04 9.80E-06 3.03E-05 5.24E-05 9.80E-06 2.34E-05 2.80E-05 9.70E-06 1.73E-05 1.13E-05 9.70E-06 7.21E-06 1.60E-07

NE 1.30E-05 2.77E-03 7.92E-03 9.90E-06 9.36E-05 2.14E-04 9.80E-06 3.20E-05 1.68E-05 9.70E-06 2.48E-05 6.50E-06 9.70E-06 1.82E-05 1.62E-06 9.70E-06 7.53E-06 1.99E-09

ENE 1.50E-05 2.68E-03 9.99E-03 9.90E-06 9.14E-05 2.33E-04 9.80E-06 3.14E-05 2.95E-05 9.70E-06 2.43E-05 1.49E-05 9.70E-06 1.79E-05 5.48E-06 9.70E-06 7.42E-06 4.88E-08

E 1.50E-05 2.21E-03 7.90E-03 9.90E-06 7.81E-05 3.04E-04 9.80E-06 2.72E-05 6.32E-05 9.70E-06 2.11E-05 3.61E-05 9.70E-06 1.57E-05 1.61E-05 9.70E-06 6.64E-06 3.66E-07

ESE 1.90E-05 1.78E-03 8.03E-03 1.00E-05 6.51E-05 2.68E-04 9.90E-06 2.30E-05 3.43E-05 9.80E-06 1.79E-05 1.61E-05 9.80E-06 1.34E-05 5.37E-06 9.70E-06 5.78E-06 2.86E-08

SE 2.30E-05 1.75E-03 9.53E-03 1.10E-05 6.40E-05 2.94E-04 1.00E-05 2.26E-05 5.08E-05 9.90E-06 1.77E-05 2.69E-05 9.90E-06 1.32E-05 1.07E-05 9.70E-06 5.71E-06 1.40E-07

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Figure 2. Comparison of the calculated TEDE values in this study with FSAR data at 0.6-30 km from the stack

Figure 3. The total effective dose pathways contribution for the individual and the collective doses calculated by CAP88 code.

4. Conclusions The assessment of the total effective dose equivalent around the Bushehr nuclear power plant site during both normal operation and accident conditions which are essential for the safety and environmental analyses was studied in this paper. Release of radioactive material to the environment following a LBLOCA was evaluated using HOTSPOT and CAP88 codes. In normal operation condition, atmospheric dispersion of radioactive material was calculated using CROM code. All required information related to the stack characteristics, weather conditions of the reactor site and its vicinity, and population distribution were considered. This study shows that while HOTSPOT gives more accurate results near the center of BNPP site, the CAP88 presents more precision in the far distances. The occupational and public dose limits are 50 and 1 (mSv/year) respectively [10]. The computed dose values are lower than these limits for all calculations. Therefore, the results of this research prove that BNPP-1 has been designed with enough safety considerations, so that no over-exposure is expected even in case of the worst accident conditions from the viewpoint of radionuclides release to the environment (i.e. LBLOCA). Finally, the computed results are in good agreement with the BNPP FSAR data despite the fact that different computational codes and different sets of input data and assumptions have been applied in the two studies.

Acknowledgements The authors would like to thank Mr. B. Molaei and Mr. M. Sepehri for collaboration and sharing of the meteorological information.

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[3] Anvari, A., and Safarzadeh, L, “Assessment of the total effective dose equivalent for accidental release from the Tehran Research Reactor” Annals of Nuclear Energy, 50, 251–255 (2012). [4] Atomic Energy Organization of Iran, “Final Safety Analysis Report for BNPP (FSAR)” (2007). [5] Homan, S. G., and Aluzzi, F., “HotSpot Health Physics Codes User’s Guide” (2013). [6] International Atomic Energy Agency, “Atmospheric dispersion in nuclear power plant siting”, IAEA Safety Series, 17–28 (1980). [7] International Atomic Energy Agency, “Generic models and parameters for assessing the environmental transfer of radionuclides from routine releases: exposure of critical groups”, IAEA Safety Series, 57, 17 (1982). [8] International Commission on Radiological Protection, “Individual monitoring for internal exposure of workers”, ICRP Publication 78, Pergamon Press (1997). [9] International Commission on Radiological Protection, “Age-dependent doses to members of the public from intake of radionuclides: Part 4. Inhalation dose coefficients”, ICRP Publication 71, Ann. ICRP, 25, (1995). [10] IAEA, “Safety of Nuclear Power Plants: Design”, Safety Standards Series, IAEA Publication, Vienna (2000).

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