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

Liang Chen· Rui Yan· Xu-Zhong Kang · Gui-Feng Zhu·Bo Zhou· Liao-Yuan He· Yang Zou· Hong-Jie Xu

Abstract The production of radionuclides 90Sr and 131I in molten salt reactors is an attractive option to address the global shortage of radionuclides. This study evaluated the production characteristics of 90Sr and 131I 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 131I and 131Te has been increasing during the lifetime. 131I has an equilibrium time of about 40 days,a saturation activity of about 40,300 TBq,and while 131Te takes 250 min to reach equilibrium, the equilibrium activity was about 38,000 TBq. The yields of 90Sr and 90Kr decreased gradually, the equilibrium time of 90Kr was short, and 90Sr could not reach equilibrium.Based on the experimental data of molten salt reactor experiment,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-H2 bubbling method,3.49 × 105 TBq of 131I 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 131I can be extracted from the off-gas system (its cooling time is 11 days) and 109 TBq of 90Sr. The yields per unit power for 131I and 90Sr is approximately 1350 TBq/MW and 530 TBq/MW, respectively, which shows that molten salt reactors have a high economic value.

Keywordst Molten salt reactor · 90Sr · 131I · Nuclide production

1 Introduction

Radionuclides are widely used in medical treatment,metallurgy, mineral exploration, industry, and other fields.Among them,90Sr, 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 by90Sr’s 28-year half-life makes it suitable for manufacturing nuclear batteries for long-term missions such as maritime navigation marks. Finally, the popularity of90Y radiopharmaceuticals makes it a great prospect in the production of90Sr-90Y generators. According to the forecast of some institutions,the global market demand for90Sr is over$100 million per year,and90Y Ibritumomab is worth$30 million per year in sales.131I 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 mature131I medical products: Na131I injection and Na131I capsules.With an annual sale of$400 million,131I is one of the three most sought-after radionuclides in the radiation market[1].Simultaneously, the domestic production of medical radionuclides in China tends to be stagnant. In 2014, the import quantity of131I was as high as 20.5 kCi [2], while90Sr was mostly imported. The common production methods of radionuclides include the accelerator method,post-treatment product extraction method, reactor irradiation method [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].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].As one of the Generation IV reactors,the MSR has the advantages of online 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 online 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] and Kang[7]evaluated the potential of a MSR to produce99Mo 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 × 106Bq of99Mo by-products per 1 MW of electricity produced. Sheu [8] studied the production characteristics of99Mo under different core designs.Result shows that compared with HomoType and RingType, the Hete-Type core model, which offered superior fuel utilization and radiotoxicity minimization, was considered the most promising design.Yu [9]studied the recovery of131I 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 155,000 Ci of131I 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], which reveals a new method of extracting90Sr and131I from a MSR.131I exists mainly in the ionic state in the molten salt, and131Te, the precursor of131I, is a semi-soluble noble metal, half of which is dissolved in salt in the form of tellurium ions such as Te3-[11] or Te2-[12]. The other half enters the exhaust gas in the form of gaseous Te [10] and then decays in the trap before being collected. Furthermore,131I can be removed from the molten salt by HF-H2bubbling[13]and then collected by spray.90Sr and its precursor90Rb are dissolved in the salt in the form of stable fluoride, while90Kr,the precursor of90Rb,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 of90Sr and131I nuclides,and their precursors90Kr and131Te,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

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–BeF2–ZrF4–UF4–ThF4(compared with the chloride used in MCFR [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.The main parameters are listed in 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.

Fig. 1 (Color online) Horizontal (a) and vertical (b) sections of the core model

Table 1 Design parameters of 30 MW MSR

SCALE [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 gascooled fast reactors and MSRs[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:

Among which:

where λ is the decay constant, γi,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 2a shows the main decay-transmutation chain of131I (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 radionuclide131Sb. The vast majority decay products of131Sb are131Te,and only 6.2%of131Sb produces131mTe in the excited state when it decays, which accounts for half of all131mTe sources. The excited state of131mTe has two decay pathways. Approximately 3/4 of the131mTe is converted into131I through βdecay,and the remaining131mTe returns to the ground state131Te (T1/2= 25 min) by emitting internal conversion electrons. Finally, the131I produced by the two pathways generates stable131Xe through β-decay (T1/2= 8 d). Figure 2b shows the main decay-transmutation chain of90Sr(the path that accounts for less than 5%of the fission yield is omitted in the figure). The analysis of the decaytransmutation chain starts at90Kr due to the short half-life of90Br(T1/2= 1.92 s).When a thermal neutron bombards a235U atom to initiate a fission reaction,90Kr and a small amount of90mRb are produced, of which 87% of90Kr produces90Rb via β-decay, and the remaining90Kr produces90mRb. A small number (＜5%) of90mRb produce90Rb by emitting an internal conversion electron,and most of90mRb produces90Sr by β-decay.90Sr has a long half-life of 28.79 a, the decay product is90Y, which can be used in radiopharmaceuticals, and the final product of the decay chain is a stable nuclide-90Zr.

Fig. 2 Decay-transmutation chain of 131I (a) and 90Sr (b)

3 Results and discussion

3.1 Yield analysis of 90Sr,131I, and their precursors

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 UF4was 7.77%, ThF4was 1.78%, and the corresponding initial keffwas 1.2972. As burn-up proceeds,235U is consumed, and238U and232Th absorb neutrons to produce fissionable nuclides239Pu and233U. 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. 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 of235U in the entire lifetime is extremely high.However,even at the end of its life,235U accounts for more than 90% of the fissionable nuclides, whereas239Pu and233U account for only about 5%.

The thermal neutron yield data of isotopes131I,131Te,and131mTe are shown in Table 2 [21]. The independent yield 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.From235to233U to239Pu, the yield of fission nuclides to the same nuclides increases gradually.Among them,the independent yield of131I accounted for 0% of235U, 0.7% of233U, and 0.8% of239Pu, indicating that131I is basically derived from the decay of the precursor rather than the direct fission of the fissile nuclide,such as235U.Furthermore,the fission yields of131Te accounted for 89.3% of235U, 79.0% of233U, and 82.6% of239Pu of the cumulative yield of131I; therefore,131Te of the two precursors contributed most of the fission yields of131I.

Since the yields of the three fissionable nuclides to the target nuclides are different,the average cumulative yields of131I and131Te over the burn-up time of the three fissionable nuclides are studied in combination with the fission yield of the fissionable nuclides to131I and131Te, as shown in Fig. 4. The yields of131I and131Te increased continuously during the lifetime, which resulted from the increase in the contents of239Pu and233U in the fissile nuclides.The cumulative yield of131Te was approximately 3 and 10 times that of235U, respectively. Even though the yields of131I and131Te were 1.84% and 2.06%, respectively, the yields of131I and131Te reached 1.72 × 10–4TBq/MW/s and 0.282 TBq/MW/s, respectively.

In addition,131Te has a short half-life of 25 min;therefore, it takes less time to reach equilibrium. From Fig. 5a, at around 250 min, the production rate and decay rate of131Te become balanced, and the saturation activity of131Te in molten salt is about 37,405 TBq, while the corresponding activity of131mTe is about 261 TBq. Consequently,if 50%of Te enters the gas phase,the annual131I produced by Te decay in the tail gas can reach 66,310 TBq.

Fig. 3 Percentage (a) and total (b) change of fissile nuclides

Table 2 Fission yield of 131I and its precursors

Fig. 4 Fission yield of 131I and 131Te varied with burn-up time

The activity changes of131I and its precursors are illustrated in Fig. 6, which shows that131mTe has the lowest activity,of only about 1/10 of that for131Te.In fact,although the yield of131mTe is less than131Te (about 20%of131Te of the total lifetime),the cumulative concentration of131mTe in the reactor is higher than131Te due to the long half-life of131mTe (about 80 times that of131Te). After reaching equilibrium, the concentration of131mTe can reach about six times that of131Te. The activity of131I is the highest because131I is the decay product of131Te and131mTe, and131I has a longer half-life.

From Fig. 5b, the production and disappearance of131I reached equilibrium after running for about 40 days, and the equilibrium value was about 40,300 TBq. Since then,although the yield of nuclides continued to increase, as shown in 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-H2bubbling method is used to extract131I from molten salt, and the recovery rate is 95% [22], the annual131I that can be recovered from molten salt is 3.49 × 105TBq.

Fig. 5 Activity of 131Te (a) and 131I (b) varied with burn-up time

Fig. 6 Yields of 131I and its precursors varied with burn-up time

Table 3 shows the thermal neutron fission yield of90Sr and its precursor [21]. The proportions of the independent90Sr yield of the three fissile nuclides in the cumulative yield were 2.1% (235U), 6.8% (233U), and 5.1% (239Pu). It can be seen that90Sr mainly originates from the decay of the precursors90Rb and90mRb. Figure 7 displays the average cumulative yield of90Kr and90Sr and the ratio of the average cumulative yield of90Kr to90Sr versus time.Figure 7b shows that the average cumulative yield of90Kr in the entire lifetime accounts for more than 80% of the average cumulative yield of90Sr,which indicates that90Kr contributes most of the90Sr yield. Therefore, most of the90Sr in the reactor was produced by the path that90Kr decays to90Rb and90mRb, and ultimately, these two nuclides decay into90Sr. This also suggests that blowing90Kr out of the core can greatly reduce the90Sr content of the salt.The upper curve in Fig. 7a is the curve of the90Kr 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 of235U 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.

Figure 8 shows the yield variation of90Sr and90Kr.The90Kr nuclide, represented by the upper curve, reaches the highest activity(about 68,000 TBq)at the first burn-up step(0.25 d),and then,the activity decreases gradually. This is mainly due to its short half-life (T1/2= 32 s), which is much smaller than the first burn-up step (d) chosen in the calculation. The lower curve represents the activity curve of90Sr. In contrast to the activity change of90Kr, the activity of90Sr increased slowly with the increase in running time,and the activity of90Sr did not reach equilibrium at the end of life; however, the increasing speed was obviously slowed down, and the final activity reached 104TBq. This is because90Sr has a long half-life (T1/2 = 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,90Rb and90mRb were produced and decayed in the reactor, which led to an increase in the90Sr activity. In conventional pressurized water reactors,90Sr is one of the main sources of radioactivity for reprocessing waste. Therefore, removing90Kr on-line by the bubbling method can greatly reduce the workload of post-processing[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],and the90Sr entering the exhaust gas every year is approximately 327 TBq,which can reach 20% of90Sr activity in fuel in one year.

Fig. 8 Yields of 90Sr and 90Kr varied with burn-up time

3.2 Effect of power on target nuclide yield

The yield of131I 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 of235U and the mass of heavy metals corresponding to the yield in both cases are shown in Table 4.With an increase in power from 30 to 60 MW,the concentration of235U in the initial charge increases by approximately 1.03 times, corresponding to the increase in 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, where131I was about 1350 TBq/MW and90Sr about 530 TBq/MW.

3.3 Isotope separation

Since fission reactions produce many i sotopes, 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,131I has a half-life of 8.02 days and a beta-ray energy of 970.8 keV,whereas133I has a half-life of only 20.8 h and a corresponding beta-ray energy of 1.757 MeV [21].To accurately calculate activity and dose, medical radionuclides require high levels of isotopes. Generally,131I products require a radionuclide impurity activity ratio of less than 0.1% [24], and the radionuclide impurity of131I is defined as the ratio of the radioactivity of133I or135I to the total radioactivity of the iodine isotope. Since90Sr originates mainly from fully cooled post-treatment waste,and90Sr has few other isotopes,there is little requirement regarding the impurity activity ratio in this report.The major isotopes with long half-lives are89Sr(T1/2= 50.5 d)and91Sr(T1/2= 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 of131I 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 collect131Te, the Te isotopes(131Te,129Te,133Te,132Te,134Te,135Te,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(131Te 10 times half-life is selected to obtain as much131I as possible).129I has a long half-life(1.57 × 107a),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 of131I(T1/2= 8.02 d) is collected in the container. Additionally,132I(T1/2= 2.30 h,activity before cooling = 1.07 TBq),133I(T1/2= 20.8 h,activity before cooling = 4.81 TBq),134I(T1/2 = 52.5 min,activity before cooling = 15.4 TBq),and135I(T1/2= 6.57 h, activity before cooling = 11.5 TBq) were collected. According to the calculations, after 13 days of cooling,the131I activity will be 0.188 TBq,which is 32.5%of the initial value.The activity of133I was only 1.47 × 10–4TBq, the activity of the remaining nuclides was approximately zero,and the133I impurity was less than 0.1%,which meets the requirements for medicinal purposes.

When the HF-H2bubbling method was used to extract131I from molten salt, the initial product contained131I with an activity of 4 × 104TBq,132I with an activity of 6.25 × 104TBq,133I with an activity of 9.37 × 104TBq,134I with an activity of 1.11 × 105TBq, and135I with an activity of 8.88 × 104TBq.After 11 days of cooling of the extracted I,the activity of131I was 1.55 × 104TBq,which was approximately 38.8%of the initial value,the activity of133I was only 14.2TBq,andtheactivity of theremaining nuclides wasalmost zero.Therefore,the impurity of133I is less than 0.1%,which can also meet the requirements of medicine.

4 Conclusion

The radionuclides131I and90Sr 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,131I can be blown out of the molten salt in the form of HI or Te(g),and90Sr can be blown out of salt in the form of Kr.

In this study,the burn-up chains of131I and90Sr nuclides that are suitable for extraction from the reactor are arranged.131I mainly originates from the decay of131Te and131mTe,and90Sr mainly originates from the decay of90Kr,90Rb, and90mRb.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 of235U becomes smaller,while the proportions of233U and239Pu gradually increased;however, the minimum proportion of235U was greater than 90%.Consequently,theyieldchangesofthetargetnuclidesand their precursors were analyzed by combining the yield contribution and the ratio change of fissile nuclides. The results showed that the yields of131I and131Te increased gradually,while the yields of90Sr and90Kr decreased gradually.

Furthermore, we discussed the time and saturation activity required for the nuclide in the reactor to reach equilibrium activity. The equilibrium time of131I is about 40 days,and the saturation activity in salt is approximately 40,300 TBq. The equilibrium time of131Te 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.If131I is extracted from the molten salt by the HF-H2bubbling method, it can be recovered from the molten salt to 3.49 × 105TBq per year and transported into tail gas to 66,310 TBq. Additionally,90Sr can be transported into tail gas to 327 TBq per year by the helium bubbling method,which can reach 20% of90Sr activity in fuel within one year.The results show that the yield of nuclides is doubled when the power is doubled. The yield of131I and90Sr nuclides per unit power is approximately 1350 TBq and 530 TBq, respectively. Finally, the minimum cooling time of the isotope corresponding to the medicinal use is discussed, and the results showed that the impurity of I extracted by the electric field was less than 0.1% after cooling for 13 days, which could meet the medicinal requirements. The remaining131I is now 32.5% of the original. After 11 days of cooling of the extracted I collected by HF-H2bubbling,the remaining131I was 38.8%of the initial, and the impurities were less than 0.1%, which meets the requirements for medicinal purposes.