Study on the production characteristics of ¹³¹I and ⁹⁰Sr isotopes in a molten salt reactor

Study on the production characteristics of 131 I and 90 Sr isotopes in a molten salt reactor 1 2 3 first-author https://orcid.org/0000-0003-4323-0757 1 3 corresp Corresponding author. Rui Yan, E-mail: yanrui@sinap.ac.cn; Corresponding author. Rui Yan, E-mail: yanrui@sinap.ac.cn; yanrui@sinap.ac.cn https://orcid.org/0000-0002-5747-2194 1 1 https://orcid.org/0000-0003-4661-8159 1 https://orcid.org/0000-0002-4671-5816 1 https://orcid.org/0000-0002-8082-5250 1 3 corresp Corresponding author. Yang Zou, E-mail: zouyang@sinap.ac.cn Corresponding author. Yang Zou, E-mail: zouyang@sinap.ac.cn zouyang@sinap.ac.cn https://orcid.org/0000-0002-7825-2457 1 2 3 Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800, China School of Physical Science and Technology, ShanghaiTech University, Shanghai 201210, China School of Physical Science and Technology, ShanghaiTech University, Shanghai 201210, China University of Chinese Academy of Sciences, Beijing 100049, China University of Chinese Academy of Sciences, Beijing 100049, China The production of radionuclides 90 Sr and 131 I in molten salt reactors is an attractive option to address the global shortage of radionuclides. This study evaluated the production characteristics of 90 Sr and 131 I in a modular molten salt reactor, such as equilibrium time, yield, and cooling time of isotopic impurities. The fuel burn-up of a small modular molten salt reactor was analyzed by the Triton module of the scale program, and the variation in the fission yields of the two nuclides and their precursors with burn-up time. The yield of 131 I and 131 Te has been increasing during the lifetime. 131 I has an equilibrium time of about 40 days, a saturation activity of about 40300 TBq, and while 131 Te takes 250 min to reach equilibrium, the equilibrium activity was about 38000 TBq. The yields of 90 Sr and 90 Kr decreased gradually, the equilibrium time of 90 Kr was short, and 90 Sr could not reach equilibrium. Based on the experimental data of molten salt reactor experiment (MSRE), the amount of nuclide migration to the tail gas and the corresponding cooling time of the isotope impurities under different extraction methods were estimated. Using the HF-H 2 bubbling method, 3.49×10 5 TBq of 131 I can be extracted from molten salt every year, and after 13 days of cooling, the impurity content meets the medical requirements. Using the electric field method, 1296 TBq of 131 I can be extracted from the off-gas system (its cooling time is 11 days) and 109 TBq of 90 Sr. The yields per unit power for 131 I and 90 Sr is approximately 1350 TBq/MW and 530 TBq/MW, respectively, which shows that molten salt reactors have a high economic value. Molten salt reactor 90 Sr 131 I Nuclide production 1 Introduction 1 Radionuclides are widely used in medical treatment, metallurgy, mineral exploration, industry, and other fields. Among them, 90 Sr, an artificial radionuclide with a high yield in nuclear reactions, has a high application value. The nuclide can emit beta-rays with an energy of 0.546 MeV, and the penetration depth of the ray in soft tissues is 1.38 mm, which allows it to be used for skin scarring. In addition, the long-term stable power brought by 90 Sr’s 28-year half-life makes it suitable for manufacturing nuclear batteries for long-term missions such as maritime navigation marks. Finally, the popularity of 90 Y radiopharmaceuticals makes it a great prospect in the production of 90 Sr- 90 Y generators. According to the forecast of some institutions, the global market demand for 90 Sr is over $100 million per year, and 90 Y Ibritumomab is worth $30 million per year in sales. 131 I is a pure beta radionuclide with a half-life of 8.05 days. It is mainly used in the treatment of thyroid diseases and malignant tumors, as iodine can be enriched in the thyroid. Currently, there are two mature 131 I medical products: Na 131 I injection and Na 131 I capsules. With an annual sale of $400 million, 131 I is one of the three most sought-after radionuclides in the radiation market [ 1 1 ]. Simultaneously, the domestic production of medical radionuclides in China tends to be stagnant. In 2014, the import quantity of 131 I was as high as 20.5 kCi [ 2 2 ], while 90 Sr was mostly imported. The common production methods of radionuclides include the accelerator method, post-treatment product extraction method, reactor irradiation method[ 3 3 ], and reactor production method. Among them, the latter two have higher flexibility. Since 2007, the number of existing irradiation and processing facilities for medical isotopes in the world has been limited, and many of them are aging, with problems such as imminent withdrawal from the supply chain, target conversion, decommissioning and maintenance, and some planned and unplanned reactor shutdowns affect market supply [ 4 4 ]. Therefore, many countries have begun to develop research on the production of medical isotopes based on new reactors and processes. Therefore, research on nuclide production based on a molten salt reactor (MSR) has also attracted people's interest after the aqueous solution reactor [ 5 5 ]. As one of the four generation reactors, the MSR has the advantages of on-line addition of fission fuel and on-line treatment of fission products because of its use of liquid fuel, which can flow in the entire main loop. Thus, the extraction of isotopes from fission products of fluorinated salts in MSRs may be an attractive technique with the feasibility of large-scale on-line mass production of isotopes. Research has indicated that for inert gases and volatile metals produced by fission, an on-line bubbling and purging device can be used to introduce a tail gas treatment system that can then be radionuclide by an electric field or spray. Chuvilin [ 6 6 ] and Kang [ 7 7 ] evaluated the potential of a MSR to produce 99 Mo, and designed the relevant in-reactor and out-reactor extraction devices, which entered the gas phase through the gas–liquid interface, and has the capacity to produce 1.96×10 6 Bq of 99 Mo by-products per 1 MW of electricity produced. Sheu [ 8 8 ] studied the production characteristics of 99 Mo under different core designs. Result shows that compared with HomoType and RingType, the HeteType core model, which offered superior fuel utilization and radiotoxicity minimization, was considered the most promising design. Yu [ 9 9 ] studied the recovery of 131 I from MSRs by extracting solid Te through a solid separator of a by-pass loop system (BPLS). The result shows that in a 2 MW reactor, about 155000 Ci of 131 I could be produced annually. From the above, it can be concluded that many research institutions are producing radionuclides in MSRs. Additionally, there are other nuclides and methods that have potential research value. A considerable amount of I and Sr deposits were found in the tail gas system of MSRE [ 10 10 ], which reveals a new method of extracting 90 Sr and 131 I from a MSR. 131 I exists mainly in the ionic state in the molten salt, and 131 Te, the precursor of 131 I, is a semi-soluble noble metal, half of which is dissolved in salt in the form of tellurium ions such as Te 3- [ 11 11 ] or Te 2- [ 12 12 ]. The other half enters the exhaust gas in the form of gaseous Te [ 10 10 ], and then decays in the trap before being collected. Furthermore, 131 I can be removed from the molten salt by HF-H 2 bubbling [ 13 13 ] and then collected by spray. 90 Sr and its precursor 90 Rb are dissolved in the salt in the form of stable fluoride, while 90 Kr, the precursor of 90 Rb, is an insoluble rare gas, which will overflow at the gas–liquid interface of the molten salt pump or be carried by helium bubbles out of the molten salt into the off-gas treatment system. In this study, a fundamental benchmark of a small modular MSR was established. Then, critical calculation and burn-up analysis of the core physical model will be carried out by SCALE6.1. Later, the reaction chain analysis, fission yield, equilibrium concentration, yield, and cooling time of 90 Sr and 131 I nuclides, and their precursors 90 Kr and 131 Te, in the core will be launched. This provides key input data for the study of target isotope production, migration rules, and extraction methods. 2 Model and method 1 In this study, a thorium molten salt reactor (TMSR) with a miniaturized and modular design was studied. The entire reactor block, which contains the reactor container, core, and fuel salt, is designed as a common module which is replaced every 10 years by lifting. The thermal power of the reactor is 30 MW. One of the two molten salt loops is a fuel salt loop loaded with LiF-BeF 2 -ZrF 4 -UF 4 -ThF 4 (compared with the chloride used in MCFR [ 14 14 ], the behavior of fission products in fluoride salt has been studied more maturely, hence, fluoride salt is selected here), and the other is a cooling salt loop loaded with FLiBe or FNaBe. From inside to outside, the modular reactor core consists of the core active region, the graphite reflector, and the down-comer. A schematic diagram is shown in Fig. 1 Fig. 1 . The main parameters are listed in Table 1 Table 1 . The active region is composed of 127 hexagonal graphite assemblies with an edge distance of 18 cm and a height of 300 cm, of which 116 are fuel channels with an inner diameter of 6 cm, eight control rod channels, and three molten salt contact irradiation channels. The other four irradiation channels are located in a reflective layer with an equivalent thickness of 35 cm formed by splicing fan-shaped assemblies. The graphite reflective layer can reduce the neutron leakage rate and slow down the fast neutrons to reduce the radiation damage to the alloy material. 1 1 Parameter 1 1 Values 1 1 Thermal power (MW) 1 1 30 1 1 Fuel type 1 1 LiF-BeF 2 -ZrF 4 -UF 4 -ThF 4 1 1 Uranium fuel enrichment 1 1 19.75% 1 1 Lithium-7 enrichment 1 1 99.995% 1 1 Core diameter (cm) 1 1 300 1 1 Core height (cm) 1 1 300 1 1 Reactor inlet, outlet temperature (℃) 1 1 600, 700 1 1 Upper and lower partition height (cm) 1 1 27.5 SCALE [ 15 15 ](Standardized Computer Analyses for Licensing Evaluation), a modular programming system, released in 1980 by Oak Ridge National Laboratory, is widely used in nuclear safety analysis and design, covering reactor criticality safety, reactor physics, radiation shielding, spent fuel and radioactive source term analysis, sensitivity and uncertainty analysis, etc. At present, it is widely used in the optimization design of water reactors and various advanced nuclear energy systems, such as gas-cooled fast reactors and MSRs [ 16 16 - 20 20 16-20 ]. This study is based on SCALE 6.1 updated in 2008. In the calculation, TRITON, a control module, is used for the coupling calculation of transport and fuel consumption. The functional modules involved include KENO-VI of the Monte Carlo transport program, Origen-S of the burn-up calculation module, and the corresponding coupling program. The burn-up equation used for calculating the nuclides concentration is as follows: <math display="block" id="M1"><mrow><mfrac><mrow><mtext>d</mtext><msub><mi>N</mi><mi>i</mi></msub><mo stretchy="false">(</mo><mover accent="true"><mi>r</mi><mo stretchy="true">→</mo></mover><mo>,</mo><mi>t</mi><mo stretchy="false">)</mo></mrow><mrow><mtext>d</mtext><mi>t</mi></mrow></mfrac><mo>=</mo><msub><mi>β</mi><mrow><mi>i</mi><mo>−</mo><mn>1</mn></mrow></msub><msub><mi>N</mi><mrow><mi>i</mi><mo>−</mo><mn>1</mn></mrow></msub><mo stretchy="false">(</mo><mover accent="true"><mi>r</mi><mo stretchy="true">→</mo></mover><mo>,</mo><mi>t</mi><mo stretchy="false">)</mo><mo>−</mo><mo stretchy="false">(</mo><msub><mi>λ</mi><mi>i</mi></msub><mo>+</mo><mstyle displaystyle="true"><msubsup><mo>∑</mo><mrow><mi>g</mi><mo>=</mo><mn>1</mn></mrow><mi>G</mi></msubsup><mrow><msub><mi>σ</mi><mrow><mi>a</mi><mo>,</mo><mi>g</mi><mo>,</mo><mi>i</mi></mrow></msub></mrow></mstyle><mo>+</mo><msub><mi>ϕ</mi><mi>g</mi></msub><mo stretchy="false">(</mo><mover accent="true"><mi>r</mi><mo stretchy="true">→</mo></mover><mo>,</mo><mi>t</mi><mo stretchy="false">)</mo><mo stretchy="false">)</mo><msub><mi>N</mi><mi>i</mi></msub><mo stretchy="false">(</mo><mover accent="true"><mi>r</mi><mo stretchy="true">→</mo></mover><mo>,</mo><mi>t</mi><mo stretchy="false">)</mo><mo>+</mo><msub><mi>F</mi><mi>i</mi></msub><mo>.</mo></mrow></math> Among which: <math display="block" id="M2"><mrow><mi>β</mi><mi>i</mi><mo>−</mo><mn>1</mn><mo>=</mo><mrow><mo>{</mo><mrow><mtable><mtr><mtd><mrow><msub><mi>λ</mi><mrow><mi>i</mi><mo>−</mo><mn>1</mn></mrow></msub><mo>,</mo></mrow></mtd></mtr><mtr><mtd><mtable columnalign="left"><mtr><mtd><mtext>              or</mtext></mtd></mtr><mtr><mtd><mstyle displaystyle="true"><msubsup><mo>∑</mo><mrow><mi>g</mi><mo>=</mo><mn>1</mn></mrow><mi>G</mi></msubsup><mrow><msub><mi>σ</mi><mrow><mi>γ</mi><mo>,</mo><mi>g</mi><mo>,</mo><mi>i</mi><mo>−</mo><mn>1</mn></mrow></msub><msub><mi>ϕ</mi><mi>g</mi></msub><mo stretchy="false">(</mo><mover accent="true"><mi>r</mi><mo stretchy="true">→</mo></mover><mo>,</mo><mi>t</mi><mo stretchy="false">)</mo></mrow></mstyle></mtd></mtr></mtable></mtd></mtr></mtable></mrow></mrow><mo>,</mo></mrow></math> <math display="block" id="M3"><mrow><msub><mi>F</mi><mi>i</mi></msub><mo>=</mo><mstyle displaystyle="true"><msubsup><mo>∑</mo><mrow><mi>g</mi><mo>'</mo><mo>=</mo><mn>1</mn></mrow><mi>G</mi></msubsup><mrow><mstyle displaystyle="true"><msub><mo>∑</mo><mrow><mi>i</mi><mo>'</mo></mrow></msub><mrow><msub><mi>γ</mi><mrow><mi>i</mi><mo>,</mo><mi>i</mi><mo>'</mo></mrow></msub><msub><mi>σ</mi><mrow><mi>f</mi><mo>,</mo><mi>g</mi><mo>'</mo><mo>,</mo><mi>i</mi><mo>'</mo></mrow></msub><msub><mi>ϕ</mi><mrow><mi>g</mi><mo>'</mo></mrow></msub><mo stretchy="false">(</mo><mover accent="true"><mi>r</mi><mo stretchy="true">→</mo></mover><mo>,</mo><mi>t</mi><mo stretchy="false">)</mo><mo>,</mo><msub><mi>N</mi><mrow><mi>i</mi><mo>'</mo></mrow></msub><mo stretchy="false">(</mo><mover accent="true"><mi>r</mi><mo stretchy="true">→</mo></mover><mo>,</mo><mi>t</mi><mo stretchy="false">)</mo></mrow></mstyle></mrow></mstyle><mo>,</mo></mrow></math> where λ is the decay constant, γ i,i’ is the yield of nuclide i when nuclear fission reaction occurs on fissile nuclide i’ . The first term on the right of Eq.(1) represents the rate of production of the isotope i- 1 due to neutron absorption or decay, the second term represents the total loss rate of isotope i due to neutron absorption or decay, and the third term represents the rate of production due to fission reactions. Figure 2(a) Figure 2(a) shows the main decay-transmutation chain of 131 I (the path that accounts for less than 5% of the fission yield is omitted in the figure). The fission reaction induced by thermal neutrons produces radionuclide 131 Sb. The vast majority decay products of 131 Sb are 131 Te, and only 6.2% of 131 Sb produces 131m Te in the excited state when it decays, which accounts for half of all 131m Te sources. The excited state of 131m Te has two decay pathways. Approximately 3/4 of the 131m Te is converted into 131 I through β-decay, and the remaining 131m Te returns to the ground state 131 Te ( T 1/2 =25 min) by emitting internal conversion electrons. Finally, the 131 I produced by the two pathways generates stable 131 Xe through β-decay ( T 1/2 =8 d). Figure 2(b) Figure 2(b) shows the main decay-transmutation chain of 90 Sr (the path that accounts for less than 5% of the fission yield is omitted in the figure). The analysis of the decay-transmutation chain starts at 90 Kr due to the short half-life of 90 Br ( T 1/2 = 1.92 s). When a thermal neutron bombards a 235 U atom to initiate a fission reaction, 90 Kr and a small amount of 90m Rb is produced, of which 87% of 90 Kr produces 90 Rb via β-decay, and the remaining 90 Kr produces 90m Rb. A small number ( < 5%) of 90m Rb produces 90 Rb by emitting an internal conversion electron, and most of 90m Rb produces 90 Sr by β-decay. 90 Sr has a long half-life of 28.79 a, the decay product is 90 Y, which can be used in radiopharmaceuticals, and the final product of the decay chain is a stable nuclide- 90 Zr. 3 Results and Discussion 1 3.1 Yield analysis of 90 Sr, 131 I, and their precursors 2 The reactor block was designed to be replaced once every 10 years. The fuel is not processed on-line during the entire life, but rather processed in batches after being removed as a whole. To maintain the backup reactivity required for long-life full-power operation, a higher initial fuel load is used. The initial concentration of UF 4 was 7.77%, ThF 4 was 1.78%, and the corresponding initial k eff was 1.2972. As burn-up proceeds, 235 U is consumed, and 238 U and 232 Th absorb neutrons to produce fissionable nuclides 239 Pu and 233 U. Since there are three fissionable nuclides, and the yield of different fissionable nuclides to the target nuclides may not be the same, the variation in the total amount and percentage of fissionable nuclides in the reactor is discussed to measure the change in the average fission yield, as shown in Fig. 3 Fig. 3 . As the burn-up time increases, the fissile nuclides in the reactor are gradually consumed and the total mass continues to decrease. U is dominant in the initial charge and the other two fissionable nuclides need to absorb neutrons before they can be produced, therefore, the proportion of 235 U in the entire lifetime is extremely high. However, even at the end of its life, 235 U accounts for more than 90% of the fissionable nuclides, whereas 239 Pu and 233 U account for only about 5%. The thermal neutron yield data of isotopes 131 I, 131 Te, and 131m Te are shown in Table 2 Table 2 [ 21 21 ]. The independent y ield is on the left, which is the yield before the β-decay but after the fission reaction. On the right is the sum of the cumulative yield, which is the independent yield and the yield of nuclides produced by β-decay. From 235 U to 233 U to 239 Pu, the yield of fission nuclides to the same nuclides increases gradually. Among them, the independent yield of 131 I accounted for 0% of 235 U, 0.7% of 233 U, and 0.8% of 239 Pu, indicating that 131 I is basically derived from the decay of the precursor rather than the direct fission of the fissile nuclide, such as 235 U. Furthermore, the fission yields of 131 Te accounted for 89.3% of 235 U, 79.0% of 233 U, and 82.6% of 239 Pu of the cumulative yield of 131 I, therefore, 131 Te of the two precursors contributed most of the fission yields of 131 I. 1 1 Nuclide 2 1 131 Te 2 1 131m Te 2 1 131 I 1 1   1 1 Independent yield 1 1 Cumulative yield 1 1 Independent yield 1 1 Cumulative yield 1 1 Independent yield 1 1 Cumulative yield 1 1 235 U 1 1 6.79 × 10 -4 1 1 1.84 × 10 -2 1 1 1.47 × 10 -3 1 1 3.02 × 10 -3 1 1 0 1 1 2.06 × 10 -2 1 1 233 U 1 1 8.44 × 10 -4 1 1 1.92 × 10 -2 1 1 4.95 × 10 -3 1 1 6.28 × 10 -3 1 1 1.70 × 10 -4 1 1 2.43 × 10 -2 1 1 239 Pu 1 1 2.70 × 10 -3 1 1 3.17 × 10 -2 1 1 5.75 × 10 -3 1 1 8.13 × 10 -3 1 1 3.20 × 10 -4 1 1 3.84 × 10 -2 Since the yields of the three fissionable nuclides to the target nuclides is different, the average cumulative yields of 131 I and 131 Te over the burn-up time of the three fissionable nuclides is studied in combination with the fission yield of the fissionable nuclides to 131 I and 131 Te, as shown in Fig. 4 Fig. 4 . The yields of 131 I and 131 Te increased continuously during the lifetime, which resulted from the increase in the contents of 239 Pu and 233 U in the fissile nuclides. The cumulative yield of 131 Te was approximately 3 and 10 times that of 235 U, respectively. Even though the yields of 131 I and 131 Te were 1.84% and 2.06%, respectively, the yields of 131 I and 131 Te reached 1.72 × 10 -4 TBq/MW/s and 0.282 TBq/MW/s, respectively. In addition, 131 Te has a short half-life of 25 min, therefore, it takes less time to reach equilibrium. From Fig. 5(a) Fig. 5(a) , at around 250 min, the production rate and decay rate of 131 Te become balanced, and the saturation activity of 131 Te in molten salt is about 37405 TBq, while the corresponding activity of 131m Te is about 261 TBq. Consequently, if 50% of Te enters the gas phase, the annual 131 I produced by Te decay in the tail gas can reach 66310 TBq. The activity changes of 131 I and its precursors are illustrated in Fig. 6 Fig. 6 , which shows that 131m Te has the lowest activity, of only about 1/10 of that for 131 Te. In fact, although the yield of 131m Te is less than 131 Te (about 20% of 131 Te of the total lifetime), the cumulative concentration of 131m Te in the reactor is higher than 131 Te due to the long half-life of 131m Te (about 80 times that of 131 Te). After reaching equilibrium, the concentration of 131m Te can reach about six times that of 131 Te. The activity of 131 I is the highest because 131 I is the decay product of 131 Te and 131m Te, and 131 I has a longer half-life. From Fig. 5(b) Fig. 5(b) , the production and disappearance of 131 I reached equilibrium after running for about 40 days, and the equilibrium value was about 40300 TBq. Since then, although the yield of nuclides continued to increase, as shown in Fig. 3 Fig. 3 , the yield did not change much for the total mass of fissile nuclides is decreasing. Even at the end of its life, the activity is only 1.06 times the equilibrium activity. If the HF-H 2 bubbling method is used to extract 131 I from molten salt, and the recovery rate is 95%[ 22 22 ], the annual 131 I that can be recovered from molten salt is 3.49 × 10 5 TBq. Table 3 Table 3 [ 21 21 ] shows the thermal neutron fission yield of 90 Sr and its precursor. The proportions of the independent 90 Sr yield of the three fissile nuclides in the cumulative yield were 2.1% ( 235 U), 6.8% ( 233 U), and 5.1% ( 239 Pu). It can be seen that 90 Sr mainly originates from the decay of the precursors 90 Rb and 90m Rb. Figure 7 Figure 7 displays the average cumulative yield of 90 Kr and 90 Sr and the ratio of the average cumulative yield of 90 Kr to 90 Sr versus time. Fig. 7(b) Fig. 7(b) shows that the average cumulative yield of 90 Kr in the entire lifetime accounts for more than 80% of the average cumulative yield of 90 Sr, which indicates that 90 Kr contributes most of the 90 Sr yield. Therefore, most of the 90 Sr in the reactor was produced by the path that 90 Kr decays to 90 Rb and 90m Rb, and ultimately, these two nuclides decay into 90 Sr. This also suggests that blowing 90 Kr out of the core can greatly reduce the 90 Sr content of the salt. The upper curve in Fig. 7(a) Fig. 7(a) is the curve of the 90 Kr fission yield over the burn-up time. As the burn-up increased, the yield gradually decreased. This is mainly due to the decrease in the proportion of 235 U in fissile nuclides, whose contribution to yield is higher than that of the other two nuclides. Even at the end of its life, its yield is about 5.76%, which corresponds to a yield of approximately 53.7 TBq/MW/s. 1 1 Nuclide 2 1 90 Kr 2 1 90 Sr 1 1   1 1 Independent yield 1 1 Cumulative yield 1 1 Independent yield 1 1 Cumulative yield 1 1 235 U 1 1 4.66 × 10 -2 1 1 5.09 × 10 -2 1 1 1.23 × 10 -3 2 1 5.90 × 10 -2 1 1 233 U 1 1 4.47 × 10 -2 1 1 4.64 × 10 -2 1 1 4.46 × 10 -3 2 1 6.61 × 10 -2 1 1 239 Pu 1 1 1.64 × 10 -2 1 1 1.76 × 10 -2 1 1 1.23 × 10 -3 1 1 2.40 × 10 -2 Figure 8 Figure 8 shows the yield variation of 90 Sr and 90 Kr. The 90 Kr nuclide, represented by the upper curve, reaches the highest activity (about 68000 TBq) at the first burn-up step (0.25 d), then the activity decreases gradually. This is mainly due to its short half-life ( T 1/2 = 32 s), which is much smaller than the first burn-up step (d) chosen in the calculation. The lower curve represents the activity curve of 90 Sr. In contrast to the activity change of 90 Kr, the activity of 90 Sr increased slowly with the increase in running time, and the activity of 90 Sr did not reach equilibrium at the end of life, however, the increasing speed was obviously slowed down, and the final activity reached 10 4  TBq. This is because 90 Sr has a long half-life ( T 1/2 = 28.79 a), which is about three times the lifetime of the reactor, hence, the disappearance rate caused by its decay is relatively small. At the same time, 90 Rb and 90m Rb were produced and decayed in the reactor, which led to an increase in the 90 Sr activity. In conventional pressurized water reactors, 90 Sr is one of the main sources of radioactivity for reprocessing waste. Therefore, removing 90 Kr on-line by the bubbling method can greatly reduce the workload of post-processing [ 23 23 ]. If the Sr produced by Kr decay is collected directly from the exhaust gas and the time of nuclide generation to exhaust gas is ignored, the Kr overflow rate is 90%[ 10 10 ], and the 90 Sr entering the exhaust gas every year is approximately 327 TBq, which can reach 20% of 90 Sr activity in fuel in one year. 3.2 Effect of power on target nuclide yield 2 The yield of 131 I at different power levels is calculated because the power of the reactor may be affected by the switching of operating conditions during its lifetime. The initial concentration of 235 U and the mass of heavy metals corresponding to the yield in both cases are shown in Table 4 Table 4 . With an increase in power from 30 MW to 60 MW, the concentration of 235 U in the initial charge increases by approximately 1.03 times, corresponding to the increase of heavy metal mass, while the corresponding yield of nuclide is approximately twice the original. When the yield was normalized to the power, the results showed that the unit power yield was similar for 30 MW and 60 MW, where 131 I was about 1350 TBq/MW and 90 Sr about 530 TBq/MW. 1 1 Power (MW) 1 1 Initial U concentration 1 1 Heavy metal quality (t) 1 1 Burn-up depth GWD/MTIHM 1 1 131 I yield (TBq) 1 1 Unit power yield (TBq/MW) 1 1 90 Sr yield (TBq) 1 1 Unit power yield (TBq/MW) 1 1 30 1 1 7.77% 1 1 1.982 1 1 93.8 1 1 40300 1 1 1343.3 1 1 15982 1 1 532.7 1 1 60 1 1 15.85% 1 1 3.489 1 1 82.3 1 1 81000 1 1 1350.0 1 1 31539 1 1 525.5 3.3 Isotope separation 2 Since fission reactions produce many isotopes, and isotopes are essentially the same in nature, products collected by chemical means are bound to contain multiple isotopes. The half-life and beta-ray energy of different isotopes are generally different, for example, 131 I has a half-life of 8.02 days and a beta-ray energy of 970.8 keV, whereas 133 I has a half-life of only 20.8 h and a corresponding beta-ray energy of 1.757 MeV[ 21 21 ]. To accurately calculate activity and dose, medical radionuclides require high levels of isotopes. Generally, 131 I products require a radionuclide impurity activity ratio of less than 0.1%[ 24 24 ], and the radionuclide impurity of 131 I is defined as the ratio of the radioactivity of 133 I or 135 I to the total radioactivity of the iodine isotope. Since 90 Sr originates mainly from fully cooled post-treatment waste, and 90 Sr has few other isotopes, there is little requirement regarding the impurity activity ratio in this report. The major isotopes with long half-lives are 89 Sr ( T 1/2 = 50.5 d) and 91 Sr ( T 1/2 = 9.6 h). The half-life can be cooled for a long time until the activity of the remaining isotopes is reduced to a negligible degree. The half-life of 131 I is only 8 day, consequently, the economic benefit will be affected if the cooling time is too long, and therefore, the minimum cooling time is discussed. When an electric field is used to collect 131 Te, the Te isotopes ( 131 Te, 129 Te, 133 Te, 132 Te, 134 Te, 135 Te, etc.) enter the electric field in the form of gaseous Te and then decay to produce charged ions. The iodine isotope was extracted once for 250 min and then decayed for 250 min in the container ( 131 Te 10 times half-life is selected to obtain as much 131 I as possible). 129 I has a long half-life (1.57 × 10 7 a) and a low activity can be neglected, therefore, the remaining nuclides are all affected by a certain amount and need to be considered for cooling. Calculations show that about 0.578 TBq of 131 I ( T 1/2 =8.02 d) is collected in the container. Additionally, 132 I ( T 1/2 =2.30 h, activity before cooling =1.07 TBq), 133 I ( T 1/2 = 20.8 h, activity before cooling = 4.81 TBq), 134 I ( T 1/2 =52.5 min, activity before cooling =15.4 TBq), and 135 I ( T 1/2 =6.57 h, activity before cooling =11.5 TBq) were collected. According to the calculations, after 13 days of cooling, the 131 I activity will be 0.188 TBq, which is 32.5% of the initial value. The activity of 133 I was only 1.47×10 -4 TBq, the activity of the remaining nuclides was approximately zero, and the 133 I impurity was less than 0.1%, which meets the requirements for medicinal purposes. When the HF-H 2 bubbling method was used to extract 131 I from molten salt, the initial product contained 131 I with an activity of 4 × 10 4 TBq, 132 I with an activity of 6.25 × 10 4 TBq, 133 I with an activity of 9.37 × 10 4 TBq, 134 I with an activity of 1.11 × 10 5 TBq, and 135 I with an activity of 8.88 × 10 4 TBq. After 11 days of cooling of the extracted I, the activity of 131 I was 1.55 × 10 4 TBq, which was approximately 38.8% of the initial value, the activity of 133 I was only 14.2 TBq, and the activity of the remaining nuclides was almost zero. Therefore, the impurity of 133 I is less than 0.1%, which can also meet the requirements of medicine. 4 Conclusion 1 The radionuclides 131 I and 90 Sr are both widely used in industry and medicine. Traditional production methods, such as target irradiation or radioactive waste extraction, are costly and waste-intensive, and face the problems of reactor-type aging and unscheduled shutdown with uncertain yields. If the production of these two nuclides is considered in a MSR, 131 I can be blown out of the molten salt in the form of HI or Te (g), and 90 Sr can be blown out of salt in the form of Kr. In this study, the burn-up chains of 131 I and 90 Sr nuclides that are suitable for extraction from the reactor are arranged. 131 I mainly originates from the decay of 131 Te and 131m Te, and 90 Sr mainly originates from the decay of 90 Kr, 90 Rb, and 90m Rb. Subsequently, we analyzed the yield and the proportion of fissionable nuclides in the lifetime of a 30 MW MSR. The results show that the total amount of fissionable nuclides continues to decrease, and the proportion of 235 U becomes smaller, while the proportions of 233 U and 239 Pu gradually increased, however, the minimum proportion of 235 U was greater than 90%. Consequently, the yield changes of the target nuclides and their precursors were analyzed by combining the yield contribution and the ratio change of fissile nuclides. The results showed that the yields of 131 I and 131 Te increased gradually, while the yields of 90 Sr and 90 Kr decreased gradually. Furthermore, we discussed the time and saturation activity required for the nuclide in the reactor to reach equilibrium activity. The equilibrium time of 131 I is about 40 days, and the saturation activity in salt is approximately 40300 TBq. The equilibrium time of 131 Te with a slightly shorter half-life is approximately 250 min, and the equilibrium activity is 37405 TBq. Subsequently, the number of nuclides that migrate into the exhaust gas is estimated based on the migration probability in the literature. If 131 I is extracted from the molten salt by the HF-H 2 bubbling method, it can be recovered from the molten salt to 3.49 × 10 5  TBq per year and transported into tail gas to 66310 TBq. Additionally, 90 Sr can be transported into tail gas to 327 TBq per year by the helium bubbling method, which can reach 20% of 90 Sr activity in fuel within one year. The results show that the yield of nuclides is doubled when the power is doubled. 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Material preparation, data collection and analysis were performed by Liang Chen, Rui Yan, Xu-Zhong Kang, Gui-Feng Zhu, Bo Zhou, Liao-Yuan He, Yang Zou, Hong-Jie Xu. The first draft of the manuscript was written by Liang Chen and all authors commented on previous versions of the manuscript. All authors read and approved the final manuscript. 10.1007/s41365-021-00867-1 This work was supported by the Strategic Priority Research Program of the Chinese Academy of Sciences (No. XDA02010000), Thorium Uranium Fuel Cycle Characteristics and Key Problem Research Project (No. QYZDY-SSW-JSC016), and Shanghai Natural Science Foundation (No.19ZR1468000). 2020-10-12 2021-01-13 2021-01-16 2021-03-22 2021-03-01 2021-03-31 © 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. 2021 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/s41365-021-00867-1 http://dx.doi.org/10.1007/s41365-021-00867-1 . 2021 Nuclear Science and Techniques 03 32