Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation

Radionuclides in primary coolant of a fluoride salt-cooled high-temperature reactor during normal operation 1 first-author 1 1 1 1 2 1 1 1 corresp Corresponding author, chenkun@sinap.ac.cn Corresponding author, chenkun@sinap.ac.cn chenkun@sinap.ac.cn Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China University of Chinese Academy of Sciences, Beijing 100049, China University of Chinese Academy of Sciences, Beijing 100049, China The release of fission products from coated particle fuel to primary coolant, as well as the activation of coolant and impurities, were analysed for a fluoride salt-cooled high-temperature reactor (FHR) system, and the activity of radionuclides accumulated in the coolant during normal operation was calculated. The release rate (release fraction per unit time) of fission products was calculated with STACY code, which is modelled mainly based on the Fick’s law, while the activation of coolant and impurities was calculated with SCALE code. The accumulation of radionuclides in the coolant has been calculated with a simplified model, which is generally a time integration considering the generation and decay of radionuclides. The results show that activation products are the dominant gamma source in the primary coolant system during normal operation of the FHR while fission products become the dominant source after shutdown. In operation condition, health-impacts related nuclides such as 3 H, and 14 C originate from the activation of lithium and coolant impurities including carbon, nitrogen, and oxygen. According to the calculated effective cross sections of neutron activation, 6 Li and 14 N are the dominant 3 H production source and 14 C production source, respectively. Considering the high production rate, 3 H and 14 C should be treated before being released to the environment. Source term FHR Primary coolant Fick’s law Diffusion 1 Introduction 1 The concept of FHR was first proposed at the Oak Ridge National Laboratory (ORNL) in 2003 [ 1 1 ]. FHR makes use of triple-isotropic (TRISO) coated particle fuel [ 2 2 ] developed originally for high temperature reactors, and fluoride salt coolant (chemical composition: 2LiF-BeF 2 ). At the Shanghai Institute of Applied Physics (SINAP), Chinese Academy of Sciences, a test pebble-bed FHR is now under development [ 3 3 ]. The test reactor has a designed thermal power of 10 MWth, and it is molten-salt cooled and graphite moderated. General design parameters of the test FHR are listed in Table 1 Table 1 . 1 1 Item 1 1 Specification 1 1 Power 1 1 10 MWth 1 1 Coolant 1 1 2LiF-BeF 2 1 1 Loading of coolant 1 1 2.6 t 1 1 Abundance of 7Li in the coolant 1 1 99.99% 1 1 Moderator & reflector 1 1 Graphite 1 1 Maximum loading of pebbles 1 1 14650 1 1 Fuel kernel temperature 1 1 600 to 700 o C The pebble fuel used in the test FHR makes use of the TRISO particle ( Fig. 1 Fig. 1 ) which consists of a UO 2 fuel kernel within multiple coating layers to prevent the leakage of radioactive materials at temperatures up to 1600 o C [ 4 4 ]. In fuel pebbles, TRISO particles distribute in the graphite matrix inside a 0.5 cm thick graphite shell. The pebble has a diameter of 6 cm and a uranium loading of 7 g with the 235 U enrichment of 17.08%. In the fuel region, pebbles are uniformly (randomly) collocated. The primary coolant system of the test FHR has a cover gas system to remove gaseous and volatile fission products from coolant. In each refueling cycle, the reactor is designed to be operated for 250 full power days (FPDs). Since the design of the test FHR is different from traditional light water reactors or high temperature reactors, the migration behaviour of radionuclides in the reactor system needs to be investigated to understand the insight of the release mechanisms of radionuclides to the reactor subsystems and to the environment. 2 Origins of Radionuclides 1 Radionuclides in test FHR system are mainly from fission reactions and neutron activations of coolant and structural materials. Fission products such as iodine, noble gases, and some metallic fission products could come out of fuel pebbles and go into the primary coolant through diffusion and leaking processes. The activation of coolant itself and impurities could also produce radionuclides such as 3 H, 14 C, 16 N, 19 O, 20 F, etc. in the coolant system. In addition, activation products of corroded-off Hastelloy-N also exist in the primary coolant system (not being analysed in this work, see Ref. [ 6 6 ] for details). 2.1 Fission Products 2 The SiC layer of the TRISO particle provides retention to fission products both in operation and in accident. However, fission products could still penetrate the coating layer by diffusion or release caused by coating failure. Besides, uranium contamination outside the coating during manufacturing is another dominant source of raidonuclides release. The diffusion of radionuclides can be modelled with physical laws [ 7 7 ], namely the conservation of mass and the Fick’s law. The conservation of mass can be expressed with the following formula <math display="block" id="M1"><mrow><mstyle displaystyle="true"><mrow><msub><mo>∭</mo><mi>V</mi></msub></mrow></mstyle><mo stretchy="false">(</mo><mi>p</mi><mo>−</mo><mi>λ</mi><mi>c</mi><mo>−</mo><mfrac><mrow><mo>∂</mo><mi>c</mi></mrow><mrow><mo>∂</mo><mi>t</mi></mrow></mfrac><mo stretchy="false">)</mo><mtext>d</mtext><mi>V</mi><mo>=</mo><mstyle displaystyle="true"><mrow><msub><mo>∮</mo><mi>S</mi></msub></mrow></mstyle><mover accent="true"><mi>j</mi><mo>→</mo></mover><mo>⋅</mo><mtext>d</mtext><mi>S</mi><mn>,</mn></mrow></math> where V is volume, S is surface area, p is volume specific generation rate, λ is decay constant, c is nuclide concentration, and j is transport flux through outer surfaces. The diffusion flux and time-dependent change of concentration can be described with the Fick’s law [ 8 8 ] in spherical coordinates as shown below <math display="block" id="M2"><mrow><mfrac><mrow><mo>∂</mo><mi>c</mi></mrow><mrow><mo>∂</mo><mi>t</mi></mrow></mfrac><mo>=</mo><mi>D</mi><mrow><mo>(</mo><mrow><mfrac><mrow><msup><mo>∂</mo><mn>2</mn></msup><mi>c</mi></mrow><mrow><mo>∂</mo><msup><mi>r</mi><mn>2</mn></msup></mrow></mfrac><mo>+</mo><mfrac><mn>1</mn><mi>r</mi></mfrac><mfrac><mrow><mo>∂</mo><mi>c</mi></mrow><mrow><mo>∂</mo><mi>r</mi></mrow></mfrac></mrow><mo>)</mo></mrow><mo>−</mo><mi>λ</mi><mi>c</mi><mo>+</mo><mi>Q</mi><mn>,</mn></mrow></math> where c is nuclide concentration, D is diffusion coefficient, r is position or radius, λ is decay constant, and Q is fission production rate. At the surface of the fuel pebble, the boundary condition can be described as [ 7 7 ] <math display="block" id="M3"><mrow><msub><mrow><mrow><mrow><mo>−</mo><mi>D</mi><mfrac><mrow><mo>∂</mo><mi>c</mi></mrow><mrow><mo>∂</mo><mi>r</mi></mrow></mfrac></mrow><mo>|</mo></mrow></mrow><mrow><mi>r</mi><mo>=</mo><msub><mi>r</mi><mtext>s</mtext></msub></mrow></msub><mo>=</mo><mi>β</mi><mo stretchy="false">(</mo><msub><mi>c</mi><mrow><mtext>bl</mtext></mrow></msub><mo>−</mo><msub><mi>c</mi><mrow><mtext>cl</mtext></mrow></msub><mo stretchy="false">)</mo><mn>,</mn></mrow></math> where r s is radius of the pebble, β is mass transfer coefficient from the surface to the coolant, c bl is concentration in the coolant right above the surface of the pebble, an c cl is the mean concentration in the coolant. The diffusion of fission products is a structure dependent process, and effective diffusion coefficients have been applied with the Fick’s law to describe the overall diffusion process [ 7 7 ]. In the Jülich Research Center, Germany, the STACY code [ 9 9 ] was developed to model the diffusion behaviour in coated particles and fuel pebbles and to estimate the release rate and fraction of fission products. The failed fuel particles are treated as bare UO 2 kernel, while uranium contamination is directly in the graphite matrix. Free evaporation assumption ( β →∞) is applied to account for the release rate from fuel surface to the coolant. 2.2 Activation Products 2 The chemical components of coolant salt, 6 Li, 7 Li, and 19 F could be activated by neutrons through reactions including 6 Li(n, α ) 3 H, 7 Li(n, n α ) 3 H, 19 F(n, γ ) 20 F, 19 F(n, α ) 16 N, and 19 F(n, p) 19 O. Activation products of salt include 3 H and short-lived nuclides, such as 16 N, 19 O, and 20 F, which will decay and emit gammas. For example, 20 F emits gammas of 1.64 MeV, 2.2 MeV and 2.45 MeV [ 10 10 ] while 16N emits gammas of 6.13 MeV and 4.12 MeV [ 11 11 ]. 20 F and 16 N have short half lives of about 10 seconds while it is about 30 s for 19 O. During manufacturing of the coolant salt, impurities including carbon, nitrogen, and oxygen could be introduced. The reactions of coolant impurities activation include 13 C(n, γ ) 14 C, 14 N(n, p) 14 C, and 13 C(n, α ) 14 C. The activation of those impurities could produce a certain amount of 14 C, which would exist mostly in the chemical form of oxides, oxycarbides and so on [ 12 12 ]. 3 Accumulation in Primary Coolant 1 During operation of the reactor, fission products continually enter the primary coolant. Among those fission products, noble gases, iodine, and some metallic fission products could come out the SiC shell of fuel particles. Using the German code STACY, the release rate of fission products could be calculated according to the reactor operation conditions. On a conservative basis, an assumption was made that radionuclides reach their maximum activities at the beginning of the refueling cycle and remain unchanged. Thus, radionulcides release from fuels with a constant activity per unit time. The dynamic accumulation process of radionuclides in the primary coolant includes the release of fuel elements, the leak of primary coolant system, and the removal of purification system. A simplified model has been applied to estimate the maximum amount of radionuclide accumulated in the primary coolant system without considering leak and removal mechanism. Based on the release, rate the activities of radionuclides released into the primary coolant can be calculated with the following equation: <math display="block" id="M4"><mrow><msub><mi>A</mi><mn>0</mn></msub><mo stretchy="false">(</mo><mi>t</mi><mo stretchy="false">)</mo><mo>=</mo><mtext>c</mtext><mstyle displaystyle="true"><mrow><msubsup><mo>∫</mo><mn>0</mn><mi>t</mi></msubsup></mrow></mstyle><msup><mtext>e</mtext><mrow><mo>−</mo><msub><mi>λ</mi><mtext>0</mtext></msub><mo stretchy="false">(</mo><mi>t</mi><mo>−</mo><mi>τ</mi><mo stretchy="false">)</mo></mrow></msup><mtext>d</mtext><mi>τ</mi><mo>=</mo><mfrac><mtext>c</mtext><mrow><msub><mi>λ</mi><mtext>0</mtext></msub></mrow></mfrac><mo stretchy="false">(</mo><mn>1</mn><mo>−</mo><msup><mtext>e</mtext><mrow><mo>−</mo><msub><mi>λ</mi><mtext>0</mtext></msub><mi>t</mi></mrow></msup><mo stretchy="false">)</mo><mn>,</mn></mrow></math> where A 0 ( t ) is activity of parent nuclide in the coolant, c is activity of nuclide released to the coolant per unit time, λ 0 = decay constant of nuclide, and t is release duration. Several nuclides, such as 85m Kr, 133m Xe, and 135m Xe, produce 85 Kr, 133 Xe, and 135 Xe through decay, and the daughter nuclides decay, too. The radioactivity of daughter nuclides can be calculated with the following equation: <math display="block" id="M5"><mrow><mtable columnalign="left"><mtr columnalign="left"><mtd columnalign="left"><mrow><msub><mi>A</mi><mn>1</mn></msub><mo stretchy="false">(</mo><mi>t</mi><mo stretchy="false">)</mo></mrow></mtd><mtd columnalign="left"><mrow><mo>=</mo><mstyle displaystyle="true"><mrow><msubsup><mo>∫</mo><mn>0</mn><mi>t</mi></msubsup></mrow></mstyle><msub><mi>λ</mi><mn>1</mn></msub><msub><mi>f</mi><mn>1</mn></msub><msub><mi>A</mi><mn>0</mn></msub><mo stretchy="false">(</mo><mi>τ</mi><mo stretchy="false">)</mo><msup><mtext>e</mtext><mrow><mo>−</mo><msub><mi>λ</mi><mn>1</mn></msub><mo stretchy="false">(</mo><mi>t</mi><mo>−</mo><mi>τ</mi><mo stretchy="false">)</mo></mrow></msup><mtext>d</mtext><mi>τ</mi></mrow></mtd></mtr><mtr columnalign="left"><mtd columnalign="left"><mrow/></mtd><mtd columnalign="left"><mrow><mo>=</mo><mstyle displaystyle="true"><mrow><msubsup><mo>∫</mo><mn>0</mn><mi>t</mi></msubsup></mrow></mstyle><msub><mi>λ</mi><mn>1</mn></msub><msub><mi>f</mi><mn>1</mn></msub><mfrac><mi>c</mi><mrow><msub><mi>λ</mi><mn>0</mn></msub></mrow></mfrac><mo stretchy="false">(</mo><mn>1</mn><mo>−</mo><msup><mtext>e</mtext><mrow><mo>−</mo><msub><mi>λ</mi><mn>0</mn></msub><mi>τ</mi></mrow></msup><mo stretchy="false">)</mo><msup><mtext>e</mtext><mrow><mo>−</mo><msub><mi>λ</mi><mn>1</mn></msub><mo stretchy="false">(</mo><mi>t</mi><mo>−</mo><mi>τ</mi><mo stretchy="false">)</mo></mrow></msup><mtext>d</mtext><mi>τ</mi></mrow></mtd></mtr><mtr columnalign="left"><mtd columnalign="left"><mrow/></mtd><mtd columnalign="left"><mrow><mo>=</mo><mfrac><mrow><mi>c</mi><msub><mi>f</mi><mn>1</mn></msub></mrow><mrow><msub><mi>λ</mi><mn>0</mn></msub></mrow></mfrac><mo stretchy="false">(</mo><mn>1</mn><mo>−</mo><msup><mtext>e</mtext><mrow><mo>−</mo><msub><mi>λ</mi><mn>1</mn></msub><mi>t</mi></mrow></msup><mo stretchy="false">)</mo><mo>−</mo><mfrac><mrow><msub><mi>λ</mi><mn>1</mn></msub></mrow><mrow><msub><mi>λ</mi><mn>0</mn></msub></mrow></mfrac><mfrac><mrow><mi>c</mi><msub><mi>f</mi><mn>1</mn></msub></mrow><mrow><msub><mi>λ</mi><mn>1</mn></msub><mo>−</mo><msub><mi>λ</mi><mn>0</mn></msub></mrow></mfrac><mo stretchy="false">(</mo><msup><mtext>e</mtext><mrow><mo>−</mo><msub><mi>λ</mi><mn>0</mn></msub><mi>t</mi></mrow></msup><mo>−</mo><msup><mtext>e</mtext><mrow><mo>−</mo><msub><mi>λ</mi><mn>1</mn></msub><mi>t</mi></mrow></msup><mo stretchy="false">)</mo><mn>,</mn></mrow></mtd></mtr></mtable></mrow></math> where A 1 ( t ) is activity of first generation daughter nuclide in the coolant, f 1 is decay fraction of parent nuclide to daughter nuclide, and λ 1 is decay constant of daughter nuclide. 133 I and 135 I produce Xe-133m and 135m Xe while 133m Xe and 135m Xe produce 133 Xe and 135 Xe by decay. The radioactivity of second-generation daughter nuclides can be calculated with the following equation: http://html.publish.founderss.cn/rc-pub/api/common/picture?pictureId=10334853&type= http://html.publish.founderss.cn/rc-pub/api/common/picture?pictureId=10334849&type= http://html.publish.founderss.cn/rc-pub/api/common/picture?pictureId=10334853&type= where A 2 ( t ) is activity of the second generation daughter nuclide in the coolant, f 2 is decay fraction of first generation daughter nuclide to second generation daughter nuclide, and λ 2 is decay constant of second generation daughter nuclide. 4 Results and Discussions 1 4.1 Results 2 After 250 FPDs, calculated inventory of key radionuclides, that could be potentially released from fuel pebbles, in the fuels are listed in Table 2 Table 2 . The calculations were performed using TRITON and ORIGEN-S modules of SCALE 6.1 package [ 13 13 ]. 1 1 Nuclides 1 1 Activity (Bq) 1 1 Nuclides 1 1 Activity (Bq) 1 1 3 H 1 1 1.29×10 12 1 1 131 I 1 1 9.24×10 15 1 1 83m Kr 1 1 1.62×10 15 1 1 132 I 1 1 1.39×10 16 1 1 85 Kr 1 1 4.12×10 13 1 1 133 I 1 1 2.11×10 16 1 1 85m Kr 1 1 3.77×10 15 1 1 134 I 1 1 2.44×10 16 1 1 87 Kr 1 1 7.68×10 15 1 1 135 I 1 1 1.97×10 16 1 1 88 Kr 1 1 1.03×10 16 1 1 131m Xe 1 1 1.01×10 14 1 1 89 Sr 1 1 1.38×10 16 1 1 133 Xe 1 1 2.02×10 16 1 1 90 Sr 1 1 3.00×10 14 1 1 133m Xe 1 1 6.16×10 14 1 1 134 Cs 1 1 7.20×10 13 1 1 135 Xe 1 1 7.68×10 15 1 1 137 Cs 1 1 3.16×10 14 1 1 135m Xe 1 1 3.98×10 15 1 1 88 Rb 1 1 1.04×10 16 1 1   1 1   TRITON produced problem dependent burn-up library for the test FHR, and ORIGEN-S has been used to calculate inventory and for activation analysis. STACY codes have been used to calculate the release rate ( Table 3 Table 3 ) according to the design parameters of the test FHR. The calculated release rate is a sum of effects including diffusion, coating failure, and uranium contamination. Effective diffusion coefficients of cesium, stronium, silver, and noble gases are taken from Ref. [ 7 7 , 14 14 ]. For those nuclides lacking diffusion coefficients, the data of chemically similar nuclides have been applied. For example, the diffusion coefficients of cesium was used for rubidium. Since the pebble fuel produced in China, which was originally developed for high temperature gas-cooled demonstration reactor, is planned to be used in the test FHR. For the amount of uranium contamination (5.0×10 -6 ) and the particle failure rate (6.0×10 -5 ) refer to the data of HTR-10 and HTR-PM fuel elements (Ref. [ 15 15 , 16 16 ]). 1 1 Nuclide 1 1 Half-live 1 1 Release rate (s-1) 1 1 83m Kr 1 1 1.83 h 1 1 5.28×10 -10 1 1 85 Kr 1 1 10.76 a 1 1 7.45×10 -13 1 1 85m Kr 1 1 4.48 h 1 1 2.17×10 -10 1 1 87 Kr 1 1 76.3 min 1 1 7.58×10 -10 1 1 88 Kr 1 1 2.80 h 1 1 1.30×10 -11 1 1 131m Xe 1 1 12.0 d 1 1 4.66×10 -12 1 1 133 Xe 1 1 5.29 d 1 1 9.26×10 -12 1 1 133m Xe 1 1 2.2 d 1 1 2.02×10 -11 1 1 135 Xe 1 1 9.17 h 1 1 1.07×10 -10 1 1 135m Xe 1 1 15.3 min 1 1 3.75×10 -9 1 1 131 I 1 1 8.04 d 1 1 6.49×10 -12 1 1 132 I 1 1 2.38 h 1 1 4.07×10 -10 1 1 133 I 1 1 20.8 h 1 1 4.85×10 -11 1 1 134 I 1 1 52.0 min 1 1 1.11×10 -9 1 1 135 I 1 1 6.59 h 1 1 1.49×10 -10 1 1 134 Cs 1 1 2.06 a 1 1 2.10×10 -14 1 1 137 Cs 1 1 30.1 a 1 1 2.03×10 -14 1 1 88 Rb 1 1 17.8 min 1 1 5.59×10 -13 1 1 89 Sr 1 1 50.5 d 1 1 1.37×10 -17 1 1 90 Sr 1 1 28.5 a 1 1 1.36×10 -17 1 1 110m Ag 1 1 250.4 d 1 1 1.36×10 -13 Based on the results shown in Table 2 Table 2 and Table 3 Table 3 , the activity of radionuclides released from fuel pebbles to the coolant can be calculated with Eq. (4), Eq. (5), and Eq. (6). After 250 full power days, the amount of fission products are listed in Table 4 Table 4 . 1 1 Category 1 1 Nuclide 1 1 Activity (Bq) 1 1 Noble Gases 1 1 83m Kr 1 1 8.13×10 9 1 1   1 1 85 Kr 1 1 8.24×10 8 1 1   1 1 85m Kr 1 1 1.91×10 10 1 1   1 1 87 Kr 1 1 3.86×10 10 1 1   1 1 88 Kr 1 1 1.95×10 9 1 1   1 1 131m Xe 1 1 1.37×10 9 1 1   1 1 133 Xe 1 1 2.37×10 11 1 1   1 1 133m Xe 1 1 3.61×10 9 1 1   1 1 135 Xe 1 1 1.60×10 11 1 1   1 1 135m Xe 1 1 3.63×10 10 1 1 Iodine 1 1 131 I 1 1 6.01×10 10 1 1   1 1 132 I 1 1 6.94×10 10 1 1   1 1 133 I 1 1 1.11×10 11 1 1   1 1 134 I 1 1 1.21×10 11 1 1   1 1 135 I 1 1 1.01×10 11 1 1 Metallic fission products 1 1 134 Cs 1 1 2.72×10 7 1 1   1 1 137 Cs 1 1 1.32×10 8 1 1   1 1 88 Rb 1 1 8.95×10 6 1 1   1 1 89 Sr 1 1 1.16×10 6 1 1   1 1 90 Sr 1 1 8.78×10 4 1 1   1 1 110m Ag 1 1 7.59×10 5 Based on the experiences of MSRE, 100% of noble gases, 10% of iodine, and 10% of solid fission products (such as metallic fission products) would be released from the salt [ 17 17 ], and the rest of the nuclides remain in the primary coolant. Noble gases are in the form of elementary gases. Iodine after released from the salt is in the form of iodine ion or I 2 , and metallic fission products are in the form of the fluorides [ 18 18 ]. Activation products of coolant and impurities are shown in Table 5 Table 5 . In the calculation, the contents of nitrogen, carbon, and oxygen impurities were selected as 100 ppm wt according to reactor design requirements. 1 1 Nuclides 1 1 Activity (Bq) 1 1 3 H 1 1 1.73×10 15 1 1 14 C 1 1 2.87×10 10 1 1 16 N 1 1 2.95×10 15 1 1 19 O 1 1 2.44×10 14 1 1 20 F 1 1 1.15×10 16 4.2 Discussions 2 During operation of the reactor, the activity of activation products in the primary coolant system is several orders of magnitude higher than that of fission products ( Table 4 Table 4 and Table 5 Table 5 ). As shown in Fig. 2 Fig. 2 , gammas are mainly from activation products, and those short-lived activation products (16N, 20F) emitting high energy gammas should be considered in shielding design work of the test FHR. After shutdown, short-lived nuclides decay quickly and disappear shortly, while the activity of long-lived nuclides decreases slowly. Among those radionuclides accumulated in the primary coolant system, the release of 3 H and 14 C need to be paid more attention during normal operation as they can be released to the environment and result in exposure to surrounding residents through ground-shine, immersion, and food chain. 3 H production rate of the test FHR is about 0.69 TBq/MWd while it is about 0.05 TBq/MWd for CANDU 6 reactor [ 19 19 ]. 14 C production rate of the test FHR is about 1.15×10 7 Bq/MWd. Normalized to 3000 MWth, both 3 H and 14 C produced in the test FHR exceed the release limits in "Regulations for environmental radiation protection of nuclear power plant" (GB 6249-2011). Therefore, treatment of 3 H and 14 C should be considered. 3 H originates mainly from the activation of 6 Li and 7 Li [ 20 20 ]. According to the results in Table 6 Table 6 , which were calculated according to the neutron spectrum of the test FHR obtained with Monte Carlo simulation, 3 H production cross section of 6 Li is about 2.0×10 5 times higher than that of 7 Li. Although the amount of 6 Li in the coolant is much lower ( Table 1 Table 1 ), it is still the dominant 3 H production source. 14 C has several sources including the activation of 13 C, 14 N, and 17 O. Since the cross section of 14 N (n,p) 14 C is the highest, it is the dominant source of 14 C. 1 1 Reaction 1 1 Cross section (barns) 1 1 6 Li(n, α ) 3 H 1 1 1.82×10 2 1 1 7 Li(n,n α ) 3 H 1 1 9.10×10 -4 1 1 13 C(n, γ ) 14 C 1 1 3.13×10 -4 1 1 14 N(n,p) 14 C 1 1 3.55×10 -1 1 1 17 O(n, α ) 14 C 1 1 5.39×10 -2 5 Conclusion 1 The amount of radionuclides accumulated in the primary coolant system of a test FHR has been analysed in this work. The release of fission products from fuel elements to the coolant system were considered, as well as the neutron activation of the coolant itself and impurities. A simplified model, which did not take leak and removal mechanism into account, was used to account for the decay and accumulation of released radionuclides. According to the analysis, several conclusions could be made as follows: 1. During operation, short-lived activation products of the coolant are a much stronger radiation source than fission products. 2. 3 H and 14 C originate from the activation of coolant and impurities, respectively. Because of their high specific production rates, 3 H and 14 C should be treated and monitored before being released to the environment in order to comply with the regulation of authorities. 3. 6 Li in the coolant of the test FHR is a dominant 3 H production source. Efforts should be made to reduce the content of 6 Li to avoid high production rate of 3 H. 4. 14 N could be introduced in to the coolant during manufacturing; it is should be reduced as it is the dominant 14 C production source. This work is to give a first glimpse of the radionuclides migration mechanism in a pebble-bed FHR system. Since simplified models and assumptions were applied, further investigation has to be made theoretically and experimentally to gain additional insights. References C.W. Forsberg , P.F. 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Zhang , et al . Comparison among SCALE packages applied to the analysis of trium in an FHR . Nucl Tech , 2015 , 38 : 080601 . (in Chinese) DOI: 10.11889/j.0253-3219.2015.hjs.38.080601 http://doi.org/10.11889/j.0253-3219.2015.hjs.38.080601 Comparison among SCALE packages applied to the analysis of trium in an FHR 10.1007/s41365-017-0186-y This work was supported by the "Strategic Priority Research Program" of the Chinese Academy of Sciences (No. XDA02050100) and the Scientific Research Foundation for the Returned Overseas Chinese Scholars, State Education Ministry (No. Y519011032) 2015-09-03 2015-11-25 2015-12-09 2017-02-17 2017-03-01 2021-04-11 © China Science Publishing & Media Ltd. (Science Press), Shanghai Institute of Applied Physics, the Chinese Academy of Sciences, ChineseNuclear Society and Springer Nature Singapore Pte Ltd. 2017 This is a post-peer-review, pre-copyedit version of an article published in Nuclear Science and Techniques. The final authenticated version is available online at: http://dx.doi.org/10.1007/s41365-017-0186-y http://dx.doi.org/10.1007/s41365-017-0186-y . 2017 Nuclear Science and Techniques 03 28